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{{#Wiki_filter: | {{#Wiki_filter:STATE OF RHODE ISLAND AND PROVIDENCE PLANTATIONS Rhode Island Atomic Energy Commission 16 Reactor Road Narragansett, Rl 02882-1165 Telephone#401-874-2600 July 19, 2017 Docket No.50-193 Document Control Desk u.s.Nuclear Regulatory Commission (NRC)11555 Rockville Pike Rockville, Maryland 20852 Attn: Mr.Patrick Boyle, Project Manager | ||
#401-874-2600 | |||
== | ==Dear Mr.Boyle:== | ||
This letter and the enclosures constitute the annual report required by the RINSC Technical Specifications (Section 6.7.1).Enclosure 1 provides reactor operating statistics. provides information pertaining to unscheduled reactor shutdowns or scrams.Enclosure 3 discusses maintenance operations performed during the reporting period.Enclosure 4 describes changes to the facility carried out under the conditions of Section 50.59 of Chapter 10 of the Code of Federal Regulations. | |||
Lastly, Enclosure 5 summarizes the radiological controls information. | |||
Lastly,Enclosure | If there are any questions regarding this information, please call me at 401-874-9442. | ||
Sincerely, Paul W Martin Jr Reactor Supervisor Enclosures (5)Copy to: Mr.Craig Bassett, USNRC Dr.John J.Breen, Chairman, NRSC Dr.Clinton Chichester, Chairman, RIAEC Dr.Nancy Breen, RIAEC Mr.Howard Chun, RIAEC Dr.Yana K.Reshetnyak, RIAEC Dr.Nitin Padture, RIAEC Reactor Operating Statistics Technical Specification Section 6.7.1.1 Month Year Operating MWHof Hours Operation July 2016 27.52 30.33 August 2016 32.47 31.50 September 2016 26.28 24.36 October 2016 14.40 16.57 November 2016 36.00 41.02 December 2016 30.45 37.36 January 2017 13.35 8.42 February 2017 52.95 54.38 March 2017 77.27 86.81 April 2017 123.42 206.18 May 2017 161.93 269.45 June 2017 154.38 261.78 Total FY2017 750.42 1068.15 MWH or 44.51 MWD Total Energy Output Since Initial Criticality 66,785.33 MWH or 2782.72 MWD ENCLOSURE 2 UNSCHEDULED SHUTDOWNS OR SCRAMS Technical Specification Section 6.7.1.2 Page 1 of 1 The following is a list of the unscheduled shutdowns or scrams that occurred during the 2016-2017 reporting period: Date Run Logbook Page Cause Description No.07/01016 9179 62 75 Watchdog Scrammed twice during start-up due to computer Scram communication error, had to reboot com device on the reactor bridge 07/20116 9183 62 80 Instrumentation Scrammed twice due to range isolator output for WR channell being irratic.Isolator was replaced.09/19116 9201 62 102 Operator Operator error during the performance of rod worth testing with trainee.09/22116 9202 62 104 Power surge Momentary loss of power to scram magnets attributed to AC power pertibations. | |||
Sincerely, | 11/03116 9212 62 115 Operator Auto scram while raising reactor power, range selection error.02/07117 9238 62 148 Instrumentation Short period Scram prior to reaching criticality due to signal noise.02/18/17 9239 62 149 Operator Auto scram while changing reactor operating mode for training.02/14117 9241 62 152 Equipment Manually scrammed due to exhaust blower motor failure failure, confmement ventilation inoperable to support reactor operation. | ||
02/16117 9244 62 155 Instrumentation Scram while raising blade 1, no indications, no known cause.02/21117 9245 63 1 Instrumentation Scram while raising blade 3, prior to reaching criticality due to signal noise.03/02117 9252 63 8 Instrumentation Scram while raising blade 1, prior to reaching criticality due to signal noise.03/07117 9255 63 11 Instrumentation Scram prior to reaching criticality due to signal noise.03/08117 9256 63 12 Instrumentation Short period Scram twice just after reaching criticality due to signal noise.03116117 9260 63 17 Instrumentation Operator manually scrammed reactor due to erratic indication on shim safety blade#2 05/04117 9287 63 46 Operator Operator inadvertently initiated a manual rundown.05117117 9292 63 52 Instrumentation Instrument rack 4 UPS pertibation caused loss of flow auto scram 05118117 9293 63 54 Instrumentation Instrument rack 4 UPS failure caused loss of flow auto scram 06129117 9314 63 77 Instrumentation Scram while raising blade 1, prior to reaching criticality due to signal noise. | |||
09/ | ENCLOSURE 3 MAINTENANCE OPERATIONS Technical Specification 6.7.1.3 requires a listing of the major maintenance operations performed in the 2016-2017 reporting period including their impact upon the safe operation of the reactor and the reasons for the corrective maintenance. | ||
11/ | The new stack radiation monitor was installed. | ||
02/ | The new monitor is a significant upgrade and has the latest technology and far more capabilities than the old monitor.The current plan is that it will be placed in service as the primary stack monitor as soon as it has been made capable of indicating in the control room and the old monitor will continue to be maintained as a backup for as long as it continues to remain functional. | ||
02/ | The Inlet Temperature Scram was implemented to comply with the new RINSe Technical Specifications that were approved as part of the re-licensing effort. | ||
ENCLOSURE | ENCLOSURE 4 FACILITY CHANGES-10CFR50.59 REVIEW Technical Specification 6.7.1.4 requires that we provide a listing and description of any 10 CFR 50.59 evaluations conducted during the 2016-2017 reporting period.A 10 CFR 50.59 evaluation was performed on the following facility changes:*New stack radiation monitor*Inlet Temperature Scram implemented, Outlet Temperature Scram eliminated to comply with the new TS requirements. | ||
*Area Radiation Monitor upgrade.Note: This screened out as no review required, however, at the last NRC inspection visit, it was strongly recommended that we do a review.After completing the review, a conference call was held to try to understand why we were asked to complete a review on something that clearly screened out as no review required.After coming to a better understanding of the configuration and operation of our facility, it was agreed that a review was not necessary or required.All three of these 50.59 reviews are attached. | |||
ENCLOSURE 5 RADIOLOGICAL CONTROLS 1.Environmental Surveys outside the Facility-Technical Specification 6.7.1.6 Quarterly TLD1 badges are deployed outside the reactor building in three separate locations. | |||
The quarterly doses in units of mrem are shown in the table below.LOCATION 3 rd QTR 4 th QTR 1 st QTR 2 nd QTR2016201620172017 Northeast Wall 8 4 a 16 Demineralizer Door 53 59 75 390 Heat Exchanger Door 26 39 53 171 The general public does not frequent these locations and therefore occupancy factors may be used to approximate annual dose.The allowable annual external dose must be below 100 mrem per year.Assuming that the maximum time that a member of the general public would be present in one of these locations is 10 minutes per day, an occupancy factor of 0.025 can be used to obtain the annual dose that would be received by a member of the general public, in any of these areas.The annual dose at the Northeast Wall, Demineralizer and Heat Exchanger Doors is dependent on the operations schedule of the reactor.Ignoring the fact that the dose rate is not present 24 hours per day, and applying the occupancy factor of 0.025 2 , the annual dose that would be receive by an individual in the demineralizer room would be 14.43 mrem.The dose received at the Heat Exchanger Door would be 7.23 mrem.The annual dose received at the Northeast wall would be 0.7 mrem.The variations from quarter to quarter and from previous reports are also due in part to movements of items within the reactor building during the fiscal year and varying use of the different irradiation facilities. | |||
ENCLOSURE | 2.Annual Exposures Exceeding 500 mrem for facility members, 100 mrem for non-staff members or 10 mrem for members of the general public-Technical Specification 6.7.1.7 There were no personnel exposures greater than Technical Specification 6.7.1.7.requirements. | ||
* | lThermoluminescent Dosimeter; Mirion Technology reads the dosimeters at minimum of 1 mrem.2 Occupancy factor was changed from 0.01 to 0.025.0.01 was derived using occupancy factor and use factor (it was assumed that the reactor is not running 8 hours every day).We will use 0.025, from NCRP 147 for Outdoors, unattended parking lots, attics, stairways, unattended elevators, janitor's closets.0.025 is more conservative number than 0.01. | ||
3.Radioactive Effluents-Technical Specification 6.7.1.5 A.Individual gaseous effluent concentrations for each reactor operation are recorded on the Monthly Information Sheets (Form NSC-78).The concentration of radioactive materials in the effluent released from the facility exhaust stacks shall not exceed 1 E+05 times concentrations specified in 1 OCFR20, Appendix B, Table II, when averaged overtimeperiods permitted by 1 OCFR20.3 Gamma spectroscopy of stack gas samples has shown that the principal gaseous effluent is Argon-41.The maximum concentration for this principle contaminant permitted under Technical Specifications is 1 E-8 IJCi/cc x1 E+5=1 E-3 IJCi/cc.Average concentrations released during the year were less than 0.0683 of the limit.The total Argon-41 release during the reporting period was 149.54 curies.The calculated effective dose equivalent for this release is 3 mrem/year (COMPLY Code).The Comply Code report is attached.B.Liquid effluent concentrations released to the environment are documented on the Sewer Discharge Report (NSC-09).Each release was approved prior to discharge with its pH being within the acceptable range and with the sum of the fractions of the respective radioisotopes per month being below the discharge limit of 1.For the reporting period, the total volume of discharge was 1.84E7 ml.The isotopes and their relative activities discharged are given below.Radioisotope Total Activity Discharged (microcuries) | |||
H-3 1720.1 C-14 60.15 Pb-214 2.86 Bi-214 5.58 K-40 34.56 3 Technical Specifications, Section 3.7.2.1 Attachment B 10 CFR 50.59 Review NSC-51 Date: Title: 1/27/17 Approved By: Stack Monitor Upgrade Michael J.Davis As an attachment to this form, provide a written description of the proposed modification, the purpose for making the modification, and a justification for the answer to each of the following questions: | |||
ENCLOSURE | 1.Does the change require a change to the Technical Specifications of the R-95 License?YN A review of the RINSe Technical Specifications indicates that: A.Section 3.7.1.1 Whenever the following operations are in progress:*The reactor is operating. | ||
*Irradiated fuel handling is in progress.*Experiment handling is in progress for an experiment that has a significant fission product, or gaseous effluent activation product inventory. | |||
2. | *Any work on the core or control rods that could cause a reactivity change of more than 0.60%.l\k/k is in progress.*Any experiment movement that could cause a reactivity change of more than 0.60%.l\k/k is in progress.3.7.1.1.1 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement gaseous and particulate effluent (Table 3.2, Required Radiation Monitors, lines 1.1 and 1.2)shall be operating. | ||
lThermoluminescent Dosimeter; | 3.7.1.1.2 If the detector described in specification 3.7.1.1.1 fails during operation, within one hour, place in service a suitable alternative air monitor or begin an hourly grab sample analysis (grab sample analysis applies to particulate only)in lieu of having a functioning monitor.B.Section 3.7.1.2 3.7.1.2.1 The stack gaseous monitor shall alarm when radiation levels of the stack gas are 2.5 times normal levels, or greater. | ||
3.Radioactive Effluents | 3.7.1.2.2 The stack particulate monitor shall alarm when radiation levels of the stack particulates are 2 times normal levels, or greater.C.Section 3.7 Table 3-2 Required Radiation Monitors 3.2.1 Required Radiation Monitors Line Description Maximum Minimum Function Operating#Set Point Required Mode 1.1 Confinement 2.5 times 1 Indication and As per Building Exhaust normal alarm both 3.7.1.1.1 Stack Gaseous locally and in control room 1.2 Confinement 2 times 1 Indication and As per Building Exhaust normal alarm both 3.7.1.1.1 Stack Particulate locally and in control room The new stack monitor is capable of fulfilling all of the monitoring and alarm capabilities required by the Technical Specifications. | ||
-Technical Specification 6.7.1. | Consequently, this change does not require a change to the Technical Specifications of the R-95 License.2.Could the change result in more than a minimal increase in the frequency of occurrence of an accident that has been previously evaluated in the SAR?YN This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation. | ||
Consequently, this change will not increase the frequency of occurrence of an accident that has been previously evaluated in the SAR.3.Could the change result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation. | |||
B. | Consequently, this change will not increase the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.4.Could the change result in more than a minimal increase in the consequences of an accident that is evaluated in the SAR?YN This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation. | ||
Consequently, this change will not increase the consequences of an accident that is evaluated in the SAR.5.Could the change result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?Y N This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation. | |||
H- | Consequently, this change will not result in an increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.6.Could the change result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR?y N This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation. | ||
1. | Consequently, this change will not result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR.7.Could the change result in creating the possibility for a malfunction of a structure, system, or component that is important to safety, with a different result than predicted in the SAR?Y N This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation. | ||
* | Consequently, this change will not result in creating the possibility of a malfunction of a structure, system, or component that is important to safety, with a different result than predicted in the SAR.8.Could the change result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded?y N This change involves replacing a radiation monitoring instrument which is not part of any fission product barrier.Consequently, this change will not result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded.9.Could the change result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility?Y N This change involves replacing a radiation monitoring instrument which was not used in the SAR as a method of evaluation for establishing the design bases for the facility.Consequently, this change will not result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility.If the answer to any of these questions is yes, then an NRC approved license amendment pursuant to 10 CFR 50.90 shall be obtained before the modification is implemented. | ||
*Irradiated | Attachment B 10 CFR 50.59 Review NSC-51 Date: Title: 3/1/17 Approved By: Area Radiation Monitor Upgrade Michael J.Davis As an attachment to this form, provide a written description of the proposed modification, the purpose for making the modification, and a justification for the answer to each of the following questions: | ||
*Experiment | 1.Does the change require a change to the Technical Specifications of the R-95 License?YN A review of the RINSC Technical Specifications indicates that: A.Section 3.7.1.1.3 This specification requires that a minimum of one gamma sensitive radiation monitor that is capable of warning personnel of high radiation levels shall be on the main floor of the Confinement Building and over the pool when the reactor is operating. | ||
* | The new monitors are both gamma and neutron sensitive, and will be maintained on the main floor and at the pool top.B.Section3.7.1.2.3 This specification requires that the area radiation monitors shall alarm when radiation levels are 2 times normal levels, or greater.The new monitors have alarming capability, and will be set to alarm when radiation levels are 2 times normal levels, or greater.C.Sections 4.7.1.1.1 and 4.7.1.1.2 This specification requires that the area radiation monitors on the main floor of the Confinement Building and over the pool are verified to be operable each day prior to the reactor being started up from the shutdown condition, and after the channel has been repaired.This schedule will be maintained. | ||
* | D.Sections 4.7.1.2.1 and 4.7.1.2.2 This specification requires that the area radiation monitors on the main floor of the Confinement Building and over the pool shall be channel calibrated and channel tested annually.Thisschedule will be maintained. | ||
3.7.1.1.1 | Consequently, this change does not require a change to the Technical Specifications of the R-95 License. | ||
2.Could the change result in more than a minimal increase in the frequency of occurrence of an accident that has been previously evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation. | |||
3.7.1.1.2 | Consequently, this change will not increase the frequency of occurrence of an accident that has been previously evaluated in the SAR.3.Could the change result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation. | ||
3.7.1.2.2 | Consequently, this change will not increase the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.4.Could the change result in more than a minimal increase in the consequences of an accident that is evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation. | ||
# | Consequently, this change will not increase the consequences of an accident that is evaluated in the SAR.5.Could the change result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation. | ||
Consequently, | Consequently, this change will not result in an increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.6.Could the change result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation. | ||
Consequently, | Consequently, this change will not result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR. | ||
Consequently, | 7.Could the change result in creating the possibility for a malfunction of a structure, system, or component that is important to safety, with a different result than predicted in the SAR?Y N This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation. | ||
Consequently, | Consequently, this change will not result in creating the possibility of a malfunction of a structure, system, or component that is important to safety, with a different result than predicted in the SAR.8.Could the change result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded?Y N This change involves replacing radiation monitoring instruments which are not part of any fission product barrier.Consequently, this change will not result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded.9.Could the change result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility?Y N This change involves replacing radiation monitoring instruments which are not used in the SAR as a method of evaluation for establishing the design bases for the facility.Consequently, this change will not result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility.If the answer to any of these questions is yes, then an NRC approved license amendment pursuant to 10 CFR 50.90 shall be obtained before the modification is implemented. | ||
Consequently, | Attachment B 10 CFR 50.59 Review NSC-51 Date:___------'-1.:.....:/1'-='0.:.....:/1:....:....7 Ap proved By: Michael J.Davis Title: 170110 Coolant Temperature Scram Set Point Change As an attachment to this form, provide a written description of the proposed modification, the purpose for making the modification, and a justification for the answer to each of the following questions: | ||
Consequently, | 1.Does the change require a change to the Technical Specifications of the R-95 License?YN This modification involves implementing a change in the facility Technical Specifications so that cooling system temperature scram set points are consistent with new specifications. | ||
Consequently, | The new settings are: Indicator Function Required Actual Set Point Set Point Pool Temperature Scram 127 F 126 F Inlet Temperature Alarm None 115 F Inlet Temperature Scram 122 F 119 F Outlet Temperature Alarm None 126 F 2.Could the change result in more than a minimal increase in the frequency of occurrence of an accident that has been previously evaluated in the SAR?YN This modification has no impact on reactor operation, or anything that could cause an accident.It is strictly a slight change in the cooling system temperature alarm and scram set points.3.Could the change result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN This modification has no impact on the ability of the temperature system displays, and their associated relays from functioning. | ||
4.Could the change result in more than a minimal increase in the consequences of an accident that is evaluated in the SAR?YN The new Safety Analysis shows that these changes in temperature set points will have no significant increase in the consequences of an accident that has been evaluated in the SAR. | |||
9. | 5.Could the change result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN The new Safety Analysis shows that these changes in temperature set points will have no significant increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.6.Could the change result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR?YN This modification is simply a set point change and has no impact on any type of accident other than those involving overheating the fuel cladding, which have been evaluated in the SAR.7.Could the change result in creating the possibility for a malfunction of a structure, system, or component that is important to safety, with a different resultthanpredicted in the SAR?YN This modification has no impact on structure, system, or component malfunctions. | ||
8.Could the change result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded?y N This modification has no impact on design basis limits for fission product barriers as described in the SAR.9.Could the change result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility?YN This modification cannot result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility.If the answer to any of these questions is yes, then an NRC approved license amendment pursuant to 10 CFR 50.90 shall be obtained before the modification is implemented. | |||
Consequently, | 2017 Ar-41 release COMPLY: Vl.7.40 CFR Part 61 National Emission Standards for Hazardous Air Pollutants REPORT ON COMPLIANCE WITH 7/12/2017 9:40 THE CLEAN AIR ACT LIMITS FOR RADIONUCLIDE EMISSIONS FROM THE COMPLY CODE-Vl.7.Prepared by: RI Atomic Energy Commission RI Nuclear Science Center 16 Reactor Road)Narragansett) | ||
RI 02882 Sangho Nam 401-874-2600 Prepared for: u.S.Environmental Protection Agency Office of Radiation and Indoor Air Washington) | |||
Attachment | DC 20460 Page 1 2017 Ar-41 release COMPLY: V1.7.2017 Ar-41 Realse Report SCREENING LEVEL 4 DATA ENTERED: Release Rate Nuclide (curies/YEAR) | ||
1. | AR-41 1.495E+02 Release height 35 meters.Building height 18 meters.7/12/2017 9:40 The source and receptor are not on the same building.Building width 18 meters.Building length 20 meters.STACK DISTANCES) | ||
FILE: stack data 100.dat Distance DIR (meters)---------N 100.0 NNE 100.0 NE 100.0 ENE 100.0 E 100.0 Page 2 2017 Ar-41 release ESE 100.0 SE 100.0 SSE 100.0 S 100.0 SSW 100.0 SW 100.0 WSW 100.0 W 100.0 WNW 100.0 NW 100.0 NNW 100.0 COMPLY: V1.7.WINDROSE DATA)FILE: 2016 wind rose data 2m.dat Source of wind rose data: 2016 Wind rose data Dates of coverage: 1954-1994 Wind rose location: Narragansett) | |||
RI Distance to facility: 155 m Percent calm: 0.05 Wind Speed FROM Frequency (meters/s) | |||
D. | ------------------- | ||
Consequently, | N 0.062 2.00 NNE 0.058 2.00 NE 0.044 2.00 ENE 0.013 2.00 E 0.012 2.00 ESE 0.013 2.00 SE 0.058 2.00 SSE 0.049 2.00 S 0.058 2.00 SSW 0.084 2.00 SW 0.105 2.00 WSW 0.064 2.00 W 0.068 2.00 WNW 0.095 2.00 NW 0.104 2.00 NNW 0.068 2.00 Distance from the SOURCE to the FARM producing VEGETABLES is 100 meters.Page 3 7/12/2017 9:40 2017 Ar-41 release Distance from the SOURCE to the FARM producing MILK is 100 meters.Distance from the SOURCE to the FARM producing MEAT is 100 meters.NOTES: The receptor exposed to the highest concentration is located 100.meters from the source in the NE sector.He gets his VEGETABLES from a farm located 100.meters from the source in the NE sector.He gets his MEAT from a farm located 100.meters from the source in the NE sector.He gets his MILK from a farm located 100.meters from the source in the NE sector.Input parameters outside the"normal" range: COMPLY: Vl.7.Windrose wind frequency is unusually LOW.RESULTS: 7/12/2017 9:40 Effective dose equivalent: | ||
2. | ***Comply at level 4.This facility is in COMPLIANCE. | ||
Consequently, | 3.0 mrem/yr.It mayor may not be EXEMPT from reporting to the EPA.You may contact your regional EPA office for more information. | ||
Consequently, | |||
Consequently, | |||
Consequently, | |||
Consequently, | |||
7. | |||
Consequently, | |||
9. | |||
Consequently, | |||
Attachment | |||
1. | |||
4. | |||
5. | |||
8. | |||
AR- | |||
FILE: | |||
1954-1994 | |||
Narragansett) | |||
--------- | |||
---------- | |||
*** | |||
3. | |||
********** | ********** | ||
END OF COMPLIANCE REPORT********** | |||
Page 4}} | |||
Revision as of 19:52, 7 July 2018
| ML17201A005 | |
| Person / Time | |
|---|---|
| Site: | Rhode Island Atomic Energy Commission |
| Issue date: | 07/19/2017 |
| From: | Martin P W State of RI, Atomic Energy Comm, Nuclear Science Ctr |
| To: | Patrick Boyle Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML17201A005 (20) | |
Text
STATE OF RHODE ISLAND AND PROVIDENCE PLANTATIONS Rhode Island Atomic Energy Commission 16 Reactor Road Narragansett, Rl 02882-1165 Telephone#401-874-2600 July 19, 2017 Docket No.50-193 Document Control Desk u.s.Nuclear Regulatory Commission (NRC)11555 Rockville Pike Rockville, Maryland 20852 Attn: Mr.Patrick Boyle, Project Manager
Dear Mr.Boyle:
This letter and the enclosures constitute the annual report required by the RINSC Technical Specifications (Section 6.7.1).Enclosure 1 provides reactor operating statistics. provides information pertaining to unscheduled reactor shutdowns or scrams.Enclosure 3 discusses maintenance operations performed during the reporting period.Enclosure 4 describes changes to the facility carried out under the conditions of Section 50.59 of Chapter 10 of the Code of Federal Regulations.
Lastly, Enclosure 5 summarizes the radiological controls information.
If there are any questions regarding this information, please call me at 401-874-9442.
Sincerely, Paul W Martin Jr Reactor Supervisor Enclosures (5)Copy to: Mr.Craig Bassett, USNRC Dr.John J.Breen, Chairman, NRSC Dr.Clinton Chichester, Chairman, RIAEC Dr.Nancy Breen, RIAEC Mr.Howard Chun, RIAEC Dr.Yana K.Reshetnyak, RIAEC Dr.Nitin Padture, RIAEC Reactor Operating Statistics Technical Specification Section 6.7.1.1 Month Year Operating MWHof Hours Operation July 2016 27.52 30.33 August 2016 32.47 31.50 September 2016 26.28 24.36 October 2016 14.40 16.57 November 2016 36.00 41.02 December 2016 30.45 37.36 January 2017 13.35 8.42 February 2017 52.95 54.38 March 2017 77.27 86.81 April 2017 123.42 206.18 May 2017 161.93 269.45 June 2017 154.38 261.78 Total FY2017 750.42 1068.15 MWH or 44.51 MWD Total Energy Output Since Initial Criticality 66,785.33 MWH or 2782.72 MWD ENCLOSURE 2 UNSCHEDULED SHUTDOWNS OR SCRAMS Technical Specification Section 6.7.1.2 Page 1 of 1 The following is a list of the unscheduled shutdowns or scrams that occurred during the 2016-2017 reporting period: Date Run Logbook Page Cause Description No.07/01016 9179 62 75 Watchdog Scrammed twice during start-up due to computer Scram communication error, had to reboot com device on the reactor bridge 07/20116 9183 62 80 Instrumentation Scrammed twice due to range isolator output for WR channell being irratic.Isolator was replaced.09/19116 9201 62 102 Operator Operator error during the performance of rod worth testing with trainee.09/22116 9202 62 104 Power surge Momentary loss of power to scram magnets attributed to AC power pertibations.
11/03116 9212 62 115 Operator Auto scram while raising reactor power, range selection error.02/07117 9238 62 148 Instrumentation Short period Scram prior to reaching criticality due to signal noise.02/18/17 9239 62 149 Operator Auto scram while changing reactor operating mode for training.02/14117 9241 62 152 Equipment Manually scrammed due to exhaust blower motor failure failure, confmement ventilation inoperable to support reactor operation.
02/16117 9244 62 155 Instrumentation Scram while raising blade 1, no indications, no known cause.02/21117 9245 63 1 Instrumentation Scram while raising blade 3, prior to reaching criticality due to signal noise.03/02117 9252 63 8 Instrumentation Scram while raising blade 1, prior to reaching criticality due to signal noise.03/07117 9255 63 11 Instrumentation Scram prior to reaching criticality due to signal noise.03/08117 9256 63 12 Instrumentation Short period Scram twice just after reaching criticality due to signal noise.03116117 9260 63 17 Instrumentation Operator manually scrammed reactor due to erratic indication on shim safety blade#2 05/04117 9287 63 46 Operator Operator inadvertently initiated a manual rundown.05117117 9292 63 52 Instrumentation Instrument rack 4 UPS pertibation caused loss of flow auto scram 05118117 9293 63 54 Instrumentation Instrument rack 4 UPS failure caused loss of flow auto scram 06129117 9314 63 77 Instrumentation Scram while raising blade 1, prior to reaching criticality due to signal noise.
ENCLOSURE 3 MAINTENANCE OPERATIONS Technical Specification 6.7.1.3 requires a listing of the major maintenance operations performed in the 2016-2017 reporting period including their impact upon the safe operation of the reactor and the reasons for the corrective maintenance.
The new stack radiation monitor was installed.
The new monitor is a significant upgrade and has the latest technology and far more capabilities than the old monitor.The current plan is that it will be placed in service as the primary stack monitor as soon as it has been made capable of indicating in the control room and the old monitor will continue to be maintained as a backup for as long as it continues to remain functional.
The Inlet Temperature Scram was implemented to comply with the new RINSe Technical Specifications that were approved as part of the re-licensing effort.
ENCLOSURE 4 FACILITY CHANGES-10CFR50.59 REVIEW Technical Specification 6.7.1.4 requires that we provide a listing and description of any 10 CFR 50.59 evaluations conducted during the 2016-2017 reporting period.A 10 CFR 50.59 evaluation was performed on the following facility changes:*New stack radiation monitor*Inlet Temperature Scram implemented, Outlet Temperature Scram eliminated to comply with the new TS requirements.
- Area Radiation Monitor upgrade.Note: This screened out as no review required, however, at the last NRC inspection visit, it was strongly recommended that we do a review.After completing the review, a conference call was held to try to understand why we were asked to complete a review on something that clearly screened out as no review required.After coming to a better understanding of the configuration and operation of our facility, it was agreed that a review was not necessary or required.All three of these 50.59 reviews are attached.
ENCLOSURE 5 RADIOLOGICAL CONTROLS 1.Environmental Surveys outside the Facility-Technical Specification 6.7.1.6 Quarterly TLD1 badges are deployed outside the reactor building in three separate locations.
The quarterly doses in units of mrem are shown in the table below.LOCATION 3 rd QTR 4 th QTR 1 st QTR 2 nd QTR2016201620172017 Northeast Wall 8 4 a 16 Demineralizer Door 53 59 75 390 Heat Exchanger Door 26 39 53 171 The general public does not frequent these locations and therefore occupancy factors may be used to approximate annual dose.The allowable annual external dose must be below 100 mrem per year.Assuming that the maximum time that a member of the general public would be present in one of these locations is 10 minutes per day, an occupancy factor of 0.025 can be used to obtain the annual dose that would be received by a member of the general public, in any of these areas.The annual dose at the Northeast Wall, Demineralizer and Heat Exchanger Doors is dependent on the operations schedule of the reactor.Ignoring the fact that the dose rate is not present 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, and applying the occupancy factor of 0.025 2 , the annual dose that would be receive by an individual in the demineralizer room would be 14.43 mrem.The dose received at the Heat Exchanger Door would be 7.23 mrem.The annual dose received at the Northeast wall would be 0.7 mrem.The variations from quarter to quarter and from previous reports are also due in part to movements of items within the reactor building during the fiscal year and varying use of the different irradiation facilities.
2.Annual Exposures Exceeding 500 mrem for facility members, 100 mrem for non-staff members or 10 mrem for members of the general public-Technical Specification 6.7.1.7 There were no personnel exposures greater than Technical Specification 6.7.1.7.requirements.
lThermoluminescent Dosimeter; Mirion Technology reads the dosimeters at minimum of 1 mrem.2 Occupancy factor was changed from 0.01 to 0.025.0.01 was derived using occupancy factor and use factor (it was assumed that the reactor is not running 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> every day).We will use 0.025, from NCRP 147 for Outdoors, unattended parking lots, attics, stairways, unattended elevators, janitor's closets.0.025 is more conservative number than 0.01.
3.Radioactive Effluents-Technical Specification 6.7.1.5 A.Individual gaseous effluent concentrations for each reactor operation are recorded on the Monthly Information Sheets (Form NSC-78).The concentration of radioactive materials in the effluent released from the facility exhaust stacks shall not exceed 1 E+05 times concentrations specified in 1 OCFR20, Appendix B, Table II, when averaged overtimeperiods permitted by 1 OCFR20.3 Gamma spectroscopy of stack gas samples has shown that the principal gaseous effluent is Argon-41.The maximum concentration for this principle contaminant permitted under Technical Specifications is 1 E-8 IJCi/cc x1 E+5=1 E-3 IJCi/cc.Average concentrations released during the year were less than 0.0683 of the limit.The total Argon-41 release during the reporting period was 149.54 curies.The calculated effective dose equivalent for this release is 3 mrem/year (COMPLY Code).The Comply Code report is attached.B.Liquid effluent concentrations released to the environment are documented on the Sewer Discharge Report (NSC-09).Each release was approved prior to discharge with its pH being within the acceptable range and with the sum of the fractions of the respective radioisotopes per month being below the discharge limit of 1.For the reporting period, the total volume of discharge was 1.84E7 ml.The isotopes and their relative activities discharged are given below.Radioisotope Total Activity Discharged (microcuries)
H-3 1720.1 C-14 60.15 Pb-214 2.86 Bi-214 5.58 K-40 34.56 3 Technical Specifications, Section 3.7.2.1 Attachment B 10 CFR 50.59 Review NSC-51 Date: Title: 1/27/17 Approved By: Stack Monitor Upgrade Michael J.Davis As an attachment to this form, provide a written description of the proposed modification, the purpose for making the modification, and a justification for the answer to each of the following questions:
1.Does the change require a change to the Technical Specifications of the R-95 License?YN A review of the RINSe Technical Specifications indicates that: A.Section 3.7.1.1 Whenever the following operations are in progress:*The reactor is operating.
- Irradiated fuel handling is in progress.*Experiment handling is in progress for an experiment that has a significant fission product, or gaseous effluent activation product inventory.
- Any work on the core or control rods that could cause a reactivity change of more than 0.60%.l\k/k is in progress.*Any experiment movement that could cause a reactivity change of more than 0.60%.l\k/k is in progress.3.7.1.1.1 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement gaseous and particulate effluent (Table 3.2, Required Radiation Monitors, lines 1.1 and 1.2)shall be operating.
3.7.1.1.2 If the detector described in specification 3.7.1.1.1 fails during operation, within one hour, place in service a suitable alternative air monitor or begin an hourly grab sample analysis (grab sample analysis applies to particulate only)in lieu of having a functioning monitor.B.Section 3.7.1.2 3.7.1.2.1 The stack gaseous monitor shall alarm when radiation levels of the stack gas are 2.5 times normal levels, or greater.
3.7.1.2.2 The stack particulate monitor shall alarm when radiation levels of the stack particulates are 2 times normal levels, or greater.C.Section 3.7 Table 3-2 Required Radiation Monitors 3.2.1 Required Radiation Monitors Line Description Maximum Minimum Function Operating#Set Point Required Mode 1.1 Confinement 2.5 times 1 Indication and As per Building Exhaust normal alarm both 3.7.1.1.1 Stack Gaseous locally and in control room 1.2 Confinement 2 times 1 Indication and As per Building Exhaust normal alarm both 3.7.1.1.1 Stack Particulate locally and in control room The new stack monitor is capable of fulfilling all of the monitoring and alarm capabilities required by the Technical Specifications.
Consequently, this change does not require a change to the Technical Specifications of the R-95 License.2.Could the change result in more than a minimal increase in the frequency of occurrence of an accident that has been previously evaluated in the SAR?YN This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation.
Consequently, this change will not increase the frequency of occurrence of an accident that has been previously evaluated in the SAR.3.Could the change result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation.
Consequently, this change will not increase the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.4.Could the change result in more than a minimal increase in the consequences of an accident that is evaluated in the SAR?YN This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation.
Consequently, this change will not increase the consequences of an accident that is evaluated in the SAR.5.Could the change result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?Y N This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation.
Consequently, this change will not result in an increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.6.Could the change result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR?y N This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation.
Consequently, this change will not result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR.7.Could the change result in creating the possibility for a malfunction of a structure, system, or component that is important to safety, with a different result than predicted in the SAR?Y N This change involves replacing a radiation monitoring instrument which is not associated with any of the reactor safety systems and does not have any effect on reactor operation.
Consequently, this change will not result in creating the possibility of a malfunction of a structure, system, or component that is important to safety, with a different result than predicted in the SAR.8.Could the change result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded?y N This change involves replacing a radiation monitoring instrument which is not part of any fission product barrier.Consequently, this change will not result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded.9.Could the change result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility?Y N This change involves replacing a radiation monitoring instrument which was not used in the SAR as a method of evaluation for establishing the design bases for the facility.Consequently, this change will not result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility.If the answer to any of these questions is yes, then an NRC approved license amendment pursuant to 10 CFR 50.90 shall be obtained before the modification is implemented.
Attachment B 10 CFR 50.59 Review NSC-51 Date: Title: 3/1/17 Approved By: Area Radiation Monitor Upgrade Michael J.Davis As an attachment to this form, provide a written description of the proposed modification, the purpose for making the modification, and a justification for the answer to each of the following questions:
1.Does the change require a change to the Technical Specifications of the R-95 License?YN A review of the RINSC Technical Specifications indicates that: A.Section 3.7.1.1.3 This specification requires that a minimum of one gamma sensitive radiation monitor that is capable of warning personnel of high radiation levels shall be on the main floor of the Confinement Building and over the pool when the reactor is operating.
The new monitors are both gamma and neutron sensitive, and will be maintained on the main floor and at the pool top.B.Section3.7.1.2.3 This specification requires that the area radiation monitors shall alarm when radiation levels are 2 times normal levels, or greater.The new monitors have alarming capability, and will be set to alarm when radiation levels are 2 times normal levels, or greater.C.Sections 4.7.1.1.1 and 4.7.1.1.2 This specification requires that the area radiation monitors on the main floor of the Confinement Building and over the pool are verified to be operable each day prior to the reactor being started up from the shutdown condition, and after the channel has been repaired.This schedule will be maintained.
D.Sections 4.7.1.2.1 and 4.7.1.2.2 This specification requires that the area radiation monitors on the main floor of the Confinement Building and over the pool shall be channel calibrated and channel tested annually.Thisschedule will be maintained.
Consequently, this change does not require a change to the Technical Specifications of the R-95 License.
2.Could the change result in more than a minimal increase in the frequency of occurrence of an accident that has been previously evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation.
Consequently, this change will not increase the frequency of occurrence of an accident that has been previously evaluated in the SAR.3.Could the change result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation.
Consequently, this change will not increase the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.4.Could the change result in more than a minimal increase in the consequences of an accident that is evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation.
Consequently, this change will not increase the consequences of an accident that is evaluated in the SAR.5.Could the change result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation.
Consequently, this change will not result in an increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.6.Could the change result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR?YN This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation.
Consequently, this change will not result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR.
7.Could the change result in creating the possibility for a malfunction of a structure, system, or component that is important to safety, with a different result than predicted in the SAR?Y N This change involves replacing radiation monitoring instruments which are not associated with any of the reactor safety systems and does not have any affect on reactor operation.
Consequently, this change will not result in creating the possibility of a malfunction of a structure, system, or component that is important to safety, with a different result than predicted in the SAR.8.Could the change result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded?Y N This change involves replacing radiation monitoring instruments which are not part of any fission product barrier.Consequently, this change will not result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded.9.Could the change result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility?Y N This change involves replacing radiation monitoring instruments which are not used in the SAR as a method of evaluation for establishing the design bases for the facility.Consequently, this change will not result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility.If the answer to any of these questions is yes, then an NRC approved license amendment pursuant to 10 CFR 50.90 shall be obtained before the modification is implemented.
Attachment B 10 CFR 50.59 Review NSC-51 Date:___------'-1.:.....:/1'-='0.:.....:/1:....:....7 Ap proved By: Michael J.Davis Title: 170110 Coolant Temperature Scram Set Point Change As an attachment to this form, provide a written description of the proposed modification, the purpose for making the modification, and a justification for the answer to each of the following questions:
1.Does the change require a change to the Technical Specifications of the R-95 License?YN This modification involves implementing a change in the facility Technical Specifications so that cooling system temperature scram set points are consistent with new specifications.
The new settings are: Indicator Function Required Actual Set Point Set Point Pool Temperature Scram 127 F 126 F Inlet Temperature Alarm None 115 F Inlet Temperature Scram 122 F 119 F Outlet Temperature Alarm None 126 F 2.Could the change result in more than a minimal increase in the frequency of occurrence of an accident that has been previously evaluated in the SAR?YN This modification has no impact on reactor operation, or anything that could cause an accident.It is strictly a slight change in the cooling system temperature alarm and scram set points.3.Could the change result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN This modification has no impact on the ability of the temperature system displays, and their associated relays from functioning.
4.Could the change result in more than a minimal increase in the consequences of an accident that is evaluated in the SAR?YN The new Safety Analysis shows that these changes in temperature set points will have no significant increase in the consequences of an accident that has been evaluated in the SAR.
5.Could the change result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR?YN The new Safety Analysis shows that these changes in temperature set points will have no significant increase in the consequences of a malfunction of a structure, system, or component that is important to safety and evaluated in the SAR.6.Could the change result in creating the possibility of an accident of a different type than the accidents that have been evaluated in the SAR?YN This modification is simply a set point change and has no impact on any type of accident other than those involving overheating the fuel cladding, which have been evaluated in the SAR.7.Could the change result in creating the possibility for a malfunction of a structure, system, or component that is important to safety, with a different resultthanpredicted in the SAR?YN This modification has no impact on structure, system, or component malfunctions.
8.Could the change result in a design basis limit for a fission product barrier as described in the SAR being altered or exceeded?y N This modification has no impact on design basis limits for fission product barriers as described in the SAR.9.Could the change result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility?YN This modification cannot result in a departure from a method of evaluation used in the SAR for establishing the design bases for the facility.If the answer to any of these questions is yes, then an NRC approved license amendment pursuant to 10 CFR 50.90 shall be obtained before the modification is implemented.
2017 Ar-41 release COMPLY: Vl.7.40 CFR Part 61 National Emission Standards for Hazardous Air Pollutants REPORT ON COMPLIANCE WITH 7/12/2017 9:40 THE CLEAN AIR ACT LIMITS FOR RADIONUCLIDE EMISSIONS FROM THE COMPLY CODE-Vl.7.Prepared by: RI Atomic Energy Commission RI Nuclear Science Center 16 Reactor Road)Narragansett)
RI 02882 Sangho Nam 401-874-2600 Prepared for: u.S.Environmental Protection Agency Office of Radiation and Indoor Air Washington)
DC 20460 Page 1 2017 Ar-41 release COMPLY: V1.7.2017 Ar-41 Realse Report SCREENING LEVEL 4 DATA ENTERED: Release Rate Nuclide (curies/YEAR)
AR-41 1.495E+02 Release height 35 meters.Building height 18 meters.7/12/2017 9:40 The source and receptor are not on the same building.Building width 18 meters.Building length 20 meters.STACK DISTANCES)
FILE: stack data 100.dat Distance DIR (meters)---------N 100.0 NNE 100.0 NE 100.0 ENE 100.0 E 100.0 Page 2 2017 Ar-41 release ESE 100.0 SE 100.0 SSE 100.0 S 100.0 SSW 100.0 SW 100.0 WSW 100.0 W 100.0 WNW 100.0 NW 100.0 NNW 100.0 COMPLY: V1.7.WINDROSE DATA)FILE: 2016 wind rose data 2m.dat Source of wind rose data: 2016 Wind rose data Dates of coverage: 1954-1994 Wind rose location: Narragansett)
RI Distance to facility: 155 m Percent calm: 0.05 Wind Speed FROM Frequency (meters/s)
N 0.062 2.00 NNE 0.058 2.00 NE 0.044 2.00 ENE 0.013 2.00 E 0.012 2.00 ESE 0.013 2.00 SE 0.058 2.00 SSE 0.049 2.00 S 0.058 2.00 SSW 0.084 2.00 SW 0.105 2.00 WSW 0.064 2.00 W 0.068 2.00 WNW 0.095 2.00 NW 0.104 2.00 NNW 0.068 2.00 Distance from the SOURCE to the FARM producing VEGETABLES is 100 meters.Page 3 7/12/2017 9:40 2017 Ar-41 release Distance from the SOURCE to the FARM producing MILK is 100 meters.Distance from the SOURCE to the FARM producing MEAT is 100 meters.NOTES: The receptor exposed to the highest concentration is located 100.meters from the source in the NE sector.He gets his VEGETABLES from a farm located 100.meters from the source in the NE sector.He gets his MEAT from a farm located 100.meters from the source in the NE sector.He gets his MILK from a farm located 100.meters from the source in the NE sector.Input parameters outside the"normal" range: COMPLY: Vl.7.Windrose wind frequency is unusually LOW.RESULTS: 7/12/2017 9:40 Effective dose equivalent:
- Comply at level 4.This facility is in COMPLIANCE.
3.0 mrem/yr.It mayor may not be EXEMPT from reporting to the EPA.You may contact your regional EPA office for more information.
END OF COMPLIANCE REPORT**********
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