ML20212P811
| ML20212P811 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 06/30/1986 |
| From: | Leonard J LONG ISLAND LIGHTING CO. |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| SNRC-1275, NUDOCS 8609030294 | |
| Download: ML20212P811 (57) | |
Text
{{#Wiki_filter:- SDiIANNUAL RADIOACTIVE EETLUENT RELEASE REPORT lst and 2nd Quarters of 1986 Facility: Shoreham Nuclear Power Station, Unit 1 Licensee: Iong Island Lighting Cmpany, Inc. TABLE OF OCNENTS Section Page Introduction 1 A. Supplemental Information 1 B. Gaseous Effluents 4 C. Liquid Effluents 8 D. Solid Waste 12 E. Not Required F. Not Required G. PCP and ODOi Revisions, RD!P Non Capliances 14 and Major Changes to Radioactive Waste Treatment Systems Attachnent 1 Solid Waste Process Control Program - Shoreham . Nuclear Power Station Unit 1 - Revision 2 Attachment 2 Offsite Dose Calculation Manual - Shoreham Nuclear Power Station Unit 1 - Revision 8 (revised pages only)
- 1. Section E, Radiological Impact on Man, and Section F, Meteorological Data are not required by Technical Specification to be included in this report.
These sections will appear in the next report, due 60 days after January 1,1987, and will enempass all of 1986. 8609030294 860630 PDR ADOCK 0S000322 R PDR {,k
Introduction 'Ihe Shoreham Nuclear Power Station received a low power license to allow testing to 5% power on July 3, 1985. Testing was successfully ccxrpleted and a source outage began on October 8, 1985. The plant has r mained shutdown through the first half of 1986. 'Ihis Settiannual Effluent Pelease Report, submitted in accordance with Technical Specification 6.9.1.7, covers the periods frm January 1, 1986 through March 31, 1986, and April 1, 1986 through June 30, 1986. A. SUPPID1ENIAL INEDRMATION
- 1. Regulatory Idmits Shoreham's effluent regulatory limits are defined in Facility Operating License NPF-36, Shoreham Nuclear Power Station, Appendix A, Technical Specifications.
a) Limits for gaseous effluents and noble gases are covered by Technical Specifications 3.11.2.1 and 3.11.2.2. b&c) Iodines and particulates with half-lifes greater than 8 days in gaseous effluents are addressed in Technical Specifications 3.11.2.3. d) Liquid effluent limits are described in Technical Specifications 3.11.1.1 and 3.11.1.2 e) In addition, the following radionuclides for liquid effluents had typical niininum detectable activities of: Cr-51 4.0 E-07 PCi/ml Zr-95 9.0 E-08 pCi/ml Nb-95 4.5 E-O8 pCi/ml Tc-99m 3.9 E-08 pCi/ml Ba-140 1.6 E-07 pCi/ml La-140 5.6 E-08 pCi/ml
'Ihe following radior.uclides in gaseous effluents had typical minirrum detectable activities of:
Kr-85 1.3 E-06 pCi/cc Kr-85m 4.0 E-09 pCi/cc Xe-135m 2.0 E-08 pCi/cc I-133 5.O E-13 pCi/cc I-135 1.5 E-09 pCi/cc Ba-140 1.5 E-13 p Ci/cc La-140 7.6 E-14 pCi/cc i
- 2. Maxinum Permissible Concentrations a-d) Maxinum permissible concentrations are those specified in 10 CFR 20, Appendix B, Table II, Column 2. If an isotope is listed with values for SOLUBLE and INSOLUBLE states, the rore conservative value is utilized. For gasecus_ cffluents MPCs were not used. Direct calculations of dose were utilized to satisfy the dose rate limitations of Technical Specification 3.11.2.1.
- 3. Average Enenpf No isotopes above mininum detectable activities were measured in gaseous effluents. 'Iherefore, there is no reportable average energy for this time period.
- 4. Measurments and Approximations of Total Radioactivity a-d) Sanples were collected in the manner and with the frequency prescribed in Technical Specifications Surveillance Requirements 4.11.1.1.1 and 4.11.2.1.2. Sanples were analyzed in accordance with Technical Specifications Tables 4.11.1.1.1-1 and 4.11.2.1.2-1 regarding both type of analysis and level of sensitivity. Most sanples were analyzed by ganra spectroscopy with a Ge(Li) detector. A liquid scintillation counter was used to analyze for H3, Fe-55 and Sr-89,90. Sanples analyzed for iron and strontium underwent a ch mical separation prior to counting. Approved sanple collection and analysis procedures were followed.
Analytical results are examined to ensure that the mininum sensitivity levels required by Technical Specifications lower limits of detection have been met. Any identifiable peaks above background are quantified.
'Ihe nethods above were used for batch releases. These methods cmbined with gross activity measurements on process streams and total flow for these streams were used for continuous discharges.
No estimate of per nt total error is provided in Table 1A because all values for gaseous effluents were determined to be less than required lower limits of detection (LIDS). Counting LIDS reflect a two-sigma level of confidence. For liquid measurements, the overall error is estimated frm the tank sanpling error and frts counting error at values close to mini ==n detectable activities to be approximately 50%.
- 5. Batch Releases ,
a) Liquid 1st quarter 2nd Quarter
- 1. Number of batches 66 40
- 2. Total Time (minutes) 9990 5830
- 3. Maximan Time (minutes) 237 170
- 4. Average Time (minutes) 151.4 145.8
- 5. Miniman Time (minutes) 1 39
- 6. Average Flow (gpu) 6.60 E+04 2.11E+05 (Dilution) b) Gaseous - None
- 6. Abnonnal Releases a) Liquid - None b) Gaseous - None I
SEMIANNUAL RADIQACTIVE EFHUDff RELEASE REPORP B - GASEQUS EETLUENTS 1st and 2nd Quarters of 1986 All saples of gaseous effluents were analyzed and determined to be at or below mininum detectable activities (MDA) for all radionuclides listed in Shoreham's Technical Specifications. 'Ihese NDAs were below the lower limits of detection required in Technical Specification Table 4.11.2.1.2-1. In addition, no other radionuclides were identified. Therefore, no entries were made in Tables 1A,1B or IC.
TABLE 1 A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1986 GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES Umt Quener Quener Est Totai 2 E rror. % l A. Fision & activation smes
- 1. Total release Ci . E . E E
- 2. Ave: age release rate for penod pCi/sec . E E
- 3. Percent of Technical specification hnut % . E . E-_
B. Iodines
- 1. Total iodine.131 Ci . E E E
- 2. Average release rate for period pCi/sec . E . E
- 3. Percent of technical specification lirrut % . E . E C. Particulates
- 1. Particulates with halflives >B days Ci . E . E E
- 2. Average release rate for penod uCi/see . E . E
- 3. Percent of technical specification hnut % . E E
- 4. Gross alpha radioactmty C . E . E D. Tritiurn
- 1. Total release Ci E E_._ E
- 2. Average release rate for pened pC /sec . E E_
- 3. Percent of technical specification 1:mit % l . E .
E_
TABLE 18 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1986 GASEOUS EFFLUENTS-ELEVATED RELEASF CONTINUOUS MODE BATCH MODE Nuclasm Rehend Unit Owner l Quene' 2 Quenerl Quane' 2
- i. r.si. -
krypton 45 Ci . E . E . E . E kryptone$m Ci . E E . E . E krypton 87 Ci E E E E krypton 48 Ci E E E E menon.133 Ci . E E E E xenon.135 Ci E . E E . E xenon.135m Ci E E E E xenon.138 C . E E E E Others (specify) Ci . E E E E Ci . E . E E E Ci E E E E urudentified Ci E E . E E Total for penod Ci E . E E . E
- 2. lodirvs iodir.e 131 Ci . E E . E E iodine 133 C E . E E E iodine.135 Ci E E E E Total for period Ci . E . E E . E
- 3. Particulates strentium49 Ci E E E E strontium 90 Ci . E E E E asium 134 Ci E E . E E cesium 137 C E E E E banum lanthanum 140 Ci E E . E E Others (specify) Ci E . E E . E Ci . E E E E Ci . E E E E urudentified f Ci . E E . E . E t
I
TABLE 1C EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1986 GASEOUS EFFLUENTS-GROUND-LEVEL RELEASES CONTINUOUS MODE BATCH MODE Nueim nei wo urns overser , ou .se' 2 Q **" I 0 **' 2
- 1. Fision gases krypton SS Ci E
. E E E krypton 65m Ci E E E E krypton 87 Ci E E E E krypton Sb C E E E E xenon.133 Ci E E E E xenon.135 C E E E E xenon.135m Ci E E E E xenon.138 Ci E E E E Others (specify) Ci E E E E
Ci E E E E I Ci E E E E unidentified C E E E E Total for penod C E E E E
- 2. Iodines iodme.131 iodine.133 Ci Ci E E E El E E E E iodme 135 Ci E E E E Total for period Ci E E E E ,
- 3. Particulates strontium.89 Ci E E E . E strontium 90 C E E E E cesium.134 Ci i cenum.137 E E . E I
' Cs . E E E . E banum. lanthanum 140 Ci E E E E Others (specify l Ci E
. E . E El Ci E E E E Ci E E E E unidentified Cs . E , E . E E l
l l
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SEMIANNUAL RADIOACTTE EFFLUENT RELEASE REPORT C - LIQUID EFFLUENTS lst and 2nd Quarters of 1986 Tables 2A and 2E contain information on the liquid batch releases for the first and second quarters of 1986. During the first quarter there was no circulating water for approximately six weeks. During this period dilution was by service water only. No credit was taken for this dilution. There were no radionuclides above NDA in continuous liquid releases,therefore, liquid discharge data for the continuous releases are not included in the table. Isotopes that were at or below the MDA's are not includal in the Tables. While calculating the liquid batch dilution flows for this report it was discovered that there was an error in the em1mlated batch dilution flows reported in the Smiannual Radioactive Effluent Release Report for the third and fourth quarters of 1985. An amended Table 2A for the period is attached. The amended dilution flows are a factor of 52 to 86 smaller than previously reported. Shoreham's radioactive liquid effluents for the previous reporting period were so small that this error did not have a significant effect on the percent of applicable limit reported. For the same reason, dose rates due to the liquid batch releases were not recalculated. l l I t 1 I 1 l l l , . - - - - .. _,-,__ _ _ _ _ _
TABLE 2A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1986 LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES Umt Qwarier Qwerter Est Totas l I 2 Ersor.ta A. Fision and activation products
- 1. Total release (not includmg intium, gases. alpha) Ci 2.iag-04 4.74pos s.ooE+o8
- 2. Average diluted concentration during period pCi'm1 a .7s E-li 1.02 E-il.
- 3. Percent of appbcable limit 9 4.sE os i .i3 E-os B. Trkium
- 1. Total release Ci . E . E . E I
- 2. Average dduted concentration, during period pCi'm! . E . E
- 3. Percent of appbcable limit 9 . E . E C. Dissolved and entrained sases
- 1. Total release Ci . E . E , E l
- 2. Average dduted concentration during period uCi'mi . E . E
- 3. Percent of appbcable limit 9 . E . E D. Gross alpha radioactidty l 1. Total release l Ci l . E l .
E l . t l E. Yo!ume of waste released (prior to dilution) liters 4.42 Ems 2.62Ea i.coE+oi F. Volume of dilution water used during ptriod liters 2.4s E+o9 4.ssEros 2.ooE+ci e
. TABLE 28 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1986 LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE Nuclides Released Unit Quaner Cuaner / Quaneri Quener 2 strontium.89 C: k. E . E[ . E . E strontium.90 Ci \. E . Ef . E . E cesium.134 Ci \. E . E . E E cesium.137 Ci \ E . E . E . E iodme.131 C: .\ E .
/E . E . E cobalt.58 Ci .\E . E i.ss E-o4i 4.74 E-s' cobalt 60 -
Ci . \E ./ E . E I E iron 59 . C . \E / E . E i . E zinc.65 Ci . E I. E . E . E manganese 54 . Ci . R /. E . E E, I chromium.51 C E E E\ /. El zirconium. niobium.95 Ci . E\ . E 2.29 E-os . E molybdenum.99 Ci . EF . E . E . E i technetium.99m Ci . E/ \. E . E . E barium. lanthanum.140 Ci . E/ \. E . E . E cerium.141 C . y \. E . E . E - I \ Other (specify) Ci . /E \ E . E . E Ci . /E .\ E . E . E Ci ./E .\E . E . E Ci ./ E .\E . E . E Ci ./ E . \E . E . E unidentified C / E . Vi . E . E f \ Total for period (above) Ci I
/. E .
E\
\
- 2. tee-04 4.74E-s xenen.133 Ci / . E . E\ . E . E '
xenon.135 Ci I . E . E\ . E . E i I
-- .------.,.,.---.-,-,-,n.,,~_.,,,.,_-,-.n.,, - - - . , , , . . , - , - , , , , .m,. ,,, ,,,--r-,-
.,,.g.a.. ----,wm,,, w -g,,.m .w,,, +,--
AMENDED TABLE 2A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1985 LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES Unit Guarter Owener Est. Total ' 3 4 Error. % A. Fision and activation products ___
!. Total release (not includmg trit;um, gases, alpha) Ci 2.14 E-3 7.71 E-3 5.00 E+1
- 2. Average diluted concentration durint period NCi/ml 7,o4 E-it 4.7sE-n
- 3. Percent of applicable limit 'A s.s7 E-osi s.es E os B. Tritium
- 1. Total release Ci . E. . E . E l
- 2. Average diluted concentration durina period pCi/ml . E . E
- 3. Percent of appbcable limit Fs . E . E C. Dissolved and entrained sases
- 1. Total release Ci . E . E . E l
- 2. Average diluted concentration during period NCi/ml . E . E
- 3. Percent of appbcable limit Fr . E . E l
D. Gross alpha radioacthiry l l 1. Total release l Ci j. E l . E l . E l E. Volume of waste misased (prior to dilution) liters 1.19 E+9 7.70 E+6 5.00 E o F. Volume of dilution water used during period liters 2.sz E+io i.ss E+o 100E+1 i .
; SDEAl@lUAL RADIOACTIVE EFFWEtff RELEASE REPORT D - SOLID WAS1E lst and 2nd Quarters of 1986 Table 3 provides information on ahir==nts of solid waste for the first and second quarters of 1986. '1here were no irradiated fuel ahiyments.
Abbreviations used in Table 3 are as follows: SR - Spent Resin
-FM -
Filter media 1 D - Dewatered. T -120 - High Integrity Container (HIC) H7f-120 - Floor Drain Media in a HIC NA - Not Applicable - '1his section of the waste type code normally indicates the type of solidification agent used. If one was used it would have indicated m for cement. All high integrity containers are qualified to Depart 2nent of Transportation specification 7A Type A containers and were filled with . low specific activity-material. Exterior volune is 158 cubic feet and'the usable interior volane is 140 cubic feet. ; The Solid Waste shipnent stamary (Table 3) for the period July 1 to December 31,
- 1985 was based on calculations that did not include the n value for Cs137,' and thus did not list certain non-ganna emitting isotopes which are scaled fran
.Cs137. '1he calculations were redone using the actual n value for Cal 37 and the results were canpared with the reported results. '1here was no change in the total curies for Waste Type SR-D-NA T-120 and there was a'4.3% increase in the total curies for Waste Type FM-IHR ETff-120.
! 'Ihe Solid Waste shipnent stamary included in this Report is based on
- j. calculations using actual MDA values for measured isotopes.
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'mBLE 3 !
*** REGUIA'IORY GUIDE 1.21 REPORT **** !
SOLID WASTE SHIPPED OFFSITE FOR DISPOSAL
** DURING PERIOD FROM 1/1/86 '10 6/30/86 **
NASTE TYPE CUBIC ETERS CURIES % ERROR (CI) SR-D-NA T-120 8.9 1.3E-01 i 25% EM-D-NA EUr-120 26.8 5.2E-03 25%
** ESTIMATES OF MAJOR NUCLIDES BY NASTE TYPE **
NASTE TYPE NUCLIDE ABUNDANCE CURIES SR-IHiA T-120 Cr-51 16.% 2.1E-02 Mn-54 24.% 3.1E-02 Fe-55 18.% 2.4E-02 00-58 11.% 1.4E-02 00-60 22.% 2.9E-02 Ni-59 0.% 2.1E-05 Ni-63 1.% 1.1E-03 Nb-94 0.% 4.3E-07 4 H-3 0.% 2.5E-06 C-14 0.% 1.7E-07 Sr-90 0.% 8.5E-07 Tc-99 0.% 1.0E-07 I-129 0.1 4.9E-09 Cs-137 0.% 4.9E-05 Pu-241 0.% 1.2E-06 EM-D-NA EUr-120 Cr-51 21.% 1.1E-03 Mn-54 18.% 9.5E-04 Fe-55 24.% 1.2E-03 Co-58 23.% 1.2E-03 00-60 4.% 2.2E-04 Ni-63 1.% 2.7E-05 H-3 0.% 1.6E-05 C-14 3.% 1.6E-04 Sr-90 0.% 8.7E-07 Tc-99 0.% 4.3E-06 I-129 0.% 5.9E-06 Cs-137 0.% 1.0E-05
** SOLID NASTE DISPOSITION SINMARY **
NUMBER OF SHIPMENTS MODE OF TRANSPORTATION DESTINATION 8 Truck Barnwell 0 Truck Richland ! O Truck Beatty l
i SD1IANNIEL RADIOACTIVE EETLU3NT RELEASE REPORT G - PCP & 0D01 REVISICNS, REMP NON OCNFLIANCES AND MNTOR OIANGES '1D RADIQACTIVE WASTE TREA'IMENT SYSTEMS According to Technical Specifications 6.9.1.7, 6.13.2a and 6.14.2a, the Semiannual Rarlinactive Effluent Release Report shall include any changes to the Process Control Program (PCP) and to the Offsite Dose Calculation Manual (ODCM) made during the reporting period. 'Ihe following changes and the affected pages are hereby subnitted in ev=line with these requirements. t-A. 1. Revision 2 of the PCP was issued to:
- a. include reference to Plant procedures for High Integrity Container i (HIC) dewatering, waste conpaction, and use of a' ocnputerized tre.cking system;
- b. delineate organizational responsibilities and clarify program
, requirements,
- c. require review and approval of contractor procedures used at SNPS.
- 2. 'Ihe changes to the PCP will not reduce the overall conformance of the solidified waste product ty existing PCP requirements; they are
> mostly of an administrative nature and reflect inprovements over the previous Program. ! 3. 'Ihe affected pages of the PCP and the document cover, indicating Review of Operations ccumittee (ROC) approval, follow. 3 l l 1
B.
- 1. Revision 8 of the GXM incorporates changes that specify a method for release of liquid effluents by use of service water when there is no circulating water flow. Both liquid effluent setpoints and dose calculations are affected. Unlike circulating water, dilution by service water may not lead to increased temperature and, therefore, buoyancy of water released to Iong Island Sound, so a dilution factor of 1.0 is conservatively assuned.
- 2. When dilution is by service water dom calculations and setpoint determinations will be conservative since no dilution is assumed. When circulating water flow is available a dilution factor of 8.85 is used.
- 3. h affected pages of the CDCM and the signature page, indicating POC approval, follow.
Action Statement c of Technict1 Specification 3.12.1 requires that if milk sanpling is discontinued fran a location, a new location be identified, and the . cause of the unavailability of the sanple and the new location be included in this report along with the OD N revision reflecting the change. C. Milk samples at Technical Specification indicator location Ia1 became unavailable in May,1986 due to (a) the owners travel plans, at the primary location 6B1, and (b) the owners refusal to participate in the program, at the secondary location, llc 1. 'Ihe latest milking animal survey, conducted fran March through May 1986, showed no replacanent milking animals for the indicator Iceation. In lieu of milk sanpling, fresh leafy vegetables are being sanpled at time of harvest at three locations as specified in the Technical Specification and the ODN. Monthly sanple runs are being conducted to assure sample collection at time of harvest. Milk sanpling is continuing at Technical Specification control location Ia2. No change will be made to the ODN since the samples at location 6B1 will be available again after approximately one year, and the vegetable sanpling locations are already specified in the CD m. Technical Specification 6.15 states that the Seniannual Radioactive Effluent Release Report shall include major changes to radioactive waste treat 2nent systans. D. There were no major changes to radioactive waste trez.tment systans durirq this reporting period 4 i
Attachment 1 ( SOLID WASTE PROCESS CONTROL PROGRAM FOR SHOREHAM NUCLEAR POWER STATION - UNIT 1 REVISION 2 LONG ISLAND LIGHTING COMPANY Docket No. 50-322 June /7, 1986 PREPARED BY NUCLEAR ENGINEERING DEPARTMENI 4 Approved by: 'N _. Date: 0 /[d
,C Chairman Approved by ' a Date: 4!24 Plant Managdr/ v O
1 TABLE O! CONTENTS l ( PAGE f 1.C PURPOSE , . . . . . . . . . ... . . .. ... ..... 1 2.0 WASTE SOURCES . . . . . . . . . . . . ....... .. . 1 2.1 Evaporator Bottoms 2.2 Floor Drain Filter 2.3 Radwaste Filter 2.4 Spent Resin Tank 2.5 Filter Cartridges and Miscellaneous Wastes 2.6 Trash Compactor 3.0 In-House Solidification System . . ... . . ... . . . . 4 4.0 In-House Dewatering . . . . . . . . . . . . . . . . . . . 5 5.0 Mobile Solidification and Dewatering Services . . . . . . 5 6.0 Solidification Process Control Parameter Determination . 5 7.0 Solidification and Dewatering Process Control . . . . . . 6 7.1 Sampling and Analysis 7.2 ' Conditioning i 1 7.3 Batch Test Solidification 7.4 Waste Classification 7.5 Container Control 7.6 Decontamination
'77 Changes to the Process Control Program 7.8 Records and Inventory Control 8.0 Responsibilities . . . . .. . . .... ... .. . .. 9 9.0 References . . . . . . . . . .. . .. . ..... . .. 10 9.1 LILC0 Operating Procedures 9.2 Mobile Services Contractor Documents 9.3 General References Appendix A Solidification Record Sheet . . .. . ..... 14 - 17 Appendix B Dewatering Record Sheet . . . . .. . .. ... 18 - 19 FIGURE 1 Solid Radwaste System . . . . ... .... .. . . 20 9
Page 1 of 20 ( 1.0 PURPOSE The Shoreham Nuclear Power Station (SNPS) Process Control Program (PCP) describes the administrative and process controls which provide reasonable assurance of a consistent quality radioactive waste product which is acceptable for shipment and burial. Implementation of this PCP will:
. Provide assurance that waste types produced at SNPS will be classified satisfactorily in accordance with the requirements of 10CFR61.
. Provide assurance that the requirements of 10CFR61 and specific disposal site criteria for Class A unstable vaste to be solidified are met by the use of a mobile solidification system supplied by a qualified contractor. When additional sample solidification data becomes available, this PCP will be modified to demonstrate the qualification of the in-house solidification system that may be used, in addition to a mobile solidification system supplied by a qualified contractor,for processing of Class A unstable waste.
. Provide assurance that the waste form stability requirements of 10CFR61 '
for Class B and C wastes are met. This will be accomplished through the use of a mobile solidification system supplied by a qualified contractor or use of approved High Integrity Containers (HICs). Until such time as the contractor's Topical Report has been approved by the NRC, s, qualification will be based en the contractor's past record of producing i acceptable BWR waste packages for waste stream:s similar to those produced at SNPS. The contractor's Process Control Programs are referer.ced in Section 9.2 of this document. SNPS management shall ensure that the contractor's waste processing operationc are performed in accordance with procedures.
. Provide assurance that dewatered Class A, B, or C waste products meet l the applicable burial site criteria for free standing water when the in-house or contractor's dewatering equipment and procedures are used.
l . Provide assurance that the processing and packaging of solid radioactive wastes meet the requirements of federal and state regulations and disposal site criteria.
. Ensure that the quality assurance requirements delineated in 10 CFR 71.101, 71.103 and 71.105 are met for both in-house and mobile
' contractor processing. l 2.0 WASTE SOURCES 2.1 EVAPORATOR BOTTOMS ! 2.1.1 Watte Evaporntor i e i
Page 2 of 20 ( The design output of the Waste Evaporator is an 18 weight percent concentrate of dissolved and suspended solids. 2.1.2 Regenerant Evaporator The Regenerant Evaporator receives liquid chemical wastes produced by the acid / caustic regeneration of the condensate demineralizer resins. The waste is collected, neutralized, and sampled in the Regenerant Liquid and Evaporator Feed Tanks, and then pumped to the Regenerant Evaporator for concentration to a 25 weight percent mixture of sodium sulfate and other dissolved and { suspended solids. , 2.1.3 Bottoms Transfer Each evaporator is a forced circulation design with a reboiler providing process heat and an overhead entrainment separator and rectifying column which minimizes liquid droplets in the vapor. When the desired specific gravity has been reached, the concentrated evaporator bottoms are cooled and sent to the Evaporator Bottoms Tank or directly to the Evaporator Bottoes Metering Pump for solidification according to the contractor's procedure F458-P002, " Operating Procedures for Mobile In-Container Solidification of Sodium Sulfate Slurries." l In order to provide maximum flexibility, each evaporator can be used as a back-up for the other. 2.2 FLOOR DRAIN FILTER The Floor Drain Filter processes ef fluents from the floor and laundry drains. The filter is a horizontal traveling screen, precoat type, designed for air drying and air-aided discharge of the cake (without backflushing) into a shipping container for further dewatering. The waste may contain filter redia such as diatomaceous earth or a powdered resin / fiber blend type material. Class A, B and C waste which is dewatered using in-house equipment is processed according to SP R3.710.02, " Dewatering of Spent Radwaste i Media." Waste which is dewatered by the contractor's mobile equipment will be processed according to references 9.2.6 or 9.2.8.
. 2.3 RADWASTE FILTER The Radwaste Filter is used to process the following combined liquid radwaste streams:
. Low conductivity equipment drains i
. _ . . - , . . , _ , __,,_,,,.y.,m.-___-_m..-,_ --. . _ , . . . . _ . _ _ _ _._.._-.,,_,..-......-____,,,m
Page 3 of 20 ( . Low conductivity wastes from the condensare demineralizer regeneration systems
. Ultrasonic resin cleaner backwash
. Decanted liquid from the Phase Separator and the Spent Resin Tank
. Blowdown from the reactor water cleanup and residual heat rer. oval systems
. Blowdown from the Fuel Pool Cooling and Clean-up System The radwaste filter units are each composed of stacked horizontal filter discs assembled on an axially located hollow shaft. After draining the filter vessel and air-drying the filter cake, the filter assembly is spun to remove the filter cake from the filter discs and discharged directly into a waste shipping container for devatering and disposal.
The waste resulting from the filters may contain diatomaceoue earth, Ecodex or similar powdered resin / fiber blend material. If Class A, B or C waste is being dewatered using in-house equipment, it is processed according to SP R3.710.02, " Dewatering of Spent Radwaste Media." All classes of waste may also be dewatered by the contractor using procedures referenced in 9.2.6 or 9.2.8. 2.4 SPENT RESIN TANK The Spent Resin Tank accepts resin / sludge slurries (via the Phase Separator Tanks) from the reactor water cleanup (RWCU) filter demineralizers, the sludge from the backwash storage tank of the Fuel Pool Cooling and Clean-up etched-disc type filter, in addition to spent bead resin from the condensate demineralizers, the fuel pool demineralizers, and the radvaste demineralizers. The resin is allowed to settle before excess water is decanted to the waste collector tanks. Mixed powdered and bead resin can be pumped to the Waste Dewatering Tank for conditioning, followed by dewatering by in-house equipment (according to SP R3.710.02), or the contractor's mobile unit using procedures referenced in Section 9.0 of this PCP. 2.5 FILTER CARTRIDGES AND MISCELLANEOUS WASTES Solid wastes, such as filter cartridges from the Laundry Drain Syster. and pump suction filters and strainers, solidification samples or other radioactive debris will be immobilized in cement which may be mixed with evaporator bottoms wastes. Thece may also be compacted as Dry Active Waste (DAW) using SP R3.075.01 (provided they are dry). (
Page 4 of 20 (s Miscellaneous solids are placed within a holding device located in the approximate center of the liner and suspended off the bottom so that the cement mixture will totally surround the wastes in the basket. Liners that contain solid objects are specifically identified. 2.6 TRASH COMPACTOR The drum compactor is used to compress low level dry waste such as rags, paper, shoe covers, floor sweepings, dry filters, strainers and plastic gloves into 55 gallon steel drums for shipment offsite. Compaction force is rated at 18,000 lbs. for an approximate 4:1 compaction ratio. A box compactor will be installed which compresses waste into 96 cu. ft. metal boxes. This is a self-contained unit with its own HEPA filtering system. The compaction force is rated at 60,000 lbs. 3.0 IN-HOUSE SOLIDIFICATION SYSTEM SNPS has a permanently installed Atcor radioactive waste solidification system which, when additional test data is available, will be used to package either radioactive evaporator bottoms or resins / sludges for disposal as Class A waste. Until that time, all wastes fcr solidification will be transferred to the contractor's mobile equipment. The following is a description of the in-house system. Waste and cement lf flows are fixed by preset metering pumps using flow rates recommended by Atcor and verified by full scale testing. Flows are also monitored by tachometers installed on the control panel. The resin / sludge is processed from the Waste Dewatering Tank and evaporator bottoms are processed from the evaporator bottoms system. Cement and evaporator bottoms or resin / sludge are introduced into the mixer / feeder unit for thorough mixin. and discharge into a container. The small-volume continuous mixer limits the surface contact of the wet cement and also limits the quantity of wet cement in the system at any time. A manual handcrank is provided to permit emptying the mixer / feeder by the operator in case of power loss or equipment malfunction. Flush water connections are provided inside the mixer / feeder to remove cement residue. Safety features include:
.. An interlock to prevent fillinF unless the fill pipe is properly inserted into the container fill opening.
. An ultrasonic 1cvel sensor, and a timer to monitor waste level in the container to prevent overflowing. Cement-bearing flush water cannot be discharge 9 unless a receptacle is in place.
l I l
- - - - - - ,- - - - , _ , , , - - - , . . , - __. -, - __,._,,__,_,_.,__,__.m y _ _ , _._,
1 Page 5 of 20 ( . Failure to initiate a flush sequence within 20 minutes af ter filling stops prompts an alarm. i
- The system is operated according to SP 23.713.01, " Solid Radwaste System."
A simplified functional diagram appears in Figure 1. 4.0 IN-HOUSE DEWATERING As an alternative to solidification, Class A, B, and C devaterable waste j may be dewatered in containers which are equipped with internal filters to which a pump may be attached. Pumping continues until burial site criteria
! for free standing non corrosive liquid are met as described in < SP R3.710.02. High Integrity Containers will be used for Class B and C waste.
4 4 Dewatering is conducted in accordance with plant procedure SP R3.710.02,
" Dewatering of Spent Radwaste Media," to assure a consistently acceptable -
product. ; i 5.0 MOBILE SOLIDIFICATION AND DEWATERING SERVICES 5.1 Wastes to be solidified must be transferred to the mobile solidification / dewatering equipment which is provided and operated by a qualified contractor. Class A, B or C dewatered wastes may be processed by the mobile services contractor or the in-house dewatering O- system at the discretion of the Radwaste Engineer. Class A solidified
. wastes will be processed by the contractor or by the in-house solidification system after qualification is completed.
5.2 Section 8.2 lists those procedures in use by the mobile services contractor to ensure that vaste products meet all requirements for shipment and burial offsite. I 5.3 The mobile equipment is installed on elevation 15' of the Radwaste Building which is a Seismic Category 1 structure. Spills are contained by installing the equipment in areas where sloping floors will carry liquids to floor drain sumps. The building ventilation i system provides for filtering of particulate airborne contacination l and monitoring of radiation before it enters the station vent'.
- 6.0 SOLIDIFICATION PROCESS CONTROL PARAMETER DETERMINATION i
l When additional samples have been tested for solidification, this section will contain a summary of qualification test results for the in-house
- solidification system. During the interim, the in-house system will not be I used to solidify waste for shipment offsite.
l
Page 6 of 20 ( 7.0 SOLIDIFICATION AND DEWATERING PRODUCT CONTROL 7.1 SAMPLING AND ANALYSIS 7.1.1 Samples shall be obtained and analyzed according to SP 74.002.18. "Radwaste Sampling for Disposal," and SP R4.014.01, "Radwaste Sample Solidification Test," prior to each solidification or dewatering operation.
- 1. The waste tank to be sampled shall be recirculated for a minimum of three tank volumes prior to sampling, unless the tank had been on recirculation continuously since it began filling.
- 2. The Waste Dewatering Tank is equipped with an agitator rather than a recirculating pump. Agitation is continued for at least 25 minutes prior to sampling.
- 3. Funda, Spent Resin Tank and flat bed filter wastes must be sampled from the liner. These will not be mixed prior to sampling.
7.1.2 The waste tank sampled shall remain isolated and in recirculation or agitation, as applicable, until the solidification process is started. If it becomes necessary to add material to the tank being processed, a new batch number O. will be initiated and a new sample will be taken after an appropriate mixing time. 7.1.3 Samples will be analyzed for pH and gamma emitters. Vicual inspection will be made for oil. 7.1.4 For waste streams to be solidified, test solidifications will be performed according to the schedule described in Secticn 7.3. 7.1.5 Sampling requirements apply to all waste, whether f t is being processed by permanently installed equipment or by the contractor's mobile equipment. 7.1.6 The analysis nurber will be added to the Solidification or Dewatering Record Sheet (Section 7.8) which is prepared for each vaste container (liner or HIC). In this way the containers which belong to a particular batch can be easily identified. 7.2 CONDITIONING 7.2.1 Waste conditioning is required when any of the following conditions exist: 1
Page 7 of 20
- 1. A high or low pH condition, outside of the acceptable range according to the contractor's PCP (Section 9.2). l
- 2. Liquid content of the batch is above or below the acceptable envelope for solidification as indicated in l Section 9.2 documents.
7.2.2 Waste conditioning will be performed in accordance with procedures referenced in Section 9.0. 7.3 BATCH TEST SOLIDIFICATION 7.3.1 Test solidification shall be performed according to the following schedule:
- 1. One sample initially from each typa of wet waste, and then from every tenth batch of each type of wet waste; NOTE: Batch is defined as the total volume of waste contained in a waste mixing tank that has been prepared for solidification.
- 2. When sample analyses fall outside the acceptable envelope established by the mobile services contractor, indicating a change in vaste type;
- 3. If it is believed that some other contaminant may be present (for example, when an unusual chemical is in use at SNFS and may reach Radwaste).
7.3.2 If any test specimen fails to solidify, the solidification of the batch under test shall be deferred until such time as additional test specimens can be obtained, alternative solidification parameters can be determined, and a subsequent test verifies solidification. Solidification of the batch may then be performed using the alternative solidification parameters determined. Representative samples shall be obtained and tested from each consecutive batch of the same type of wet waste until at Icast three consecutive initial test specimens demonstrate solidification. The process control program shall be redified as required to assure solidification of subsequent batches of waste. The contractor shall modify his own PCP as necessary to accommodate unusual waste streems. 7.3.3 The test specimen shall be judged to have solidified successfully if, when its container has been removed, it remains a free standing monolith with no visible free liquid. _,__,_,._m _ _ , _ _ _ _ _ _ _ _ _ , . _ _ _ _ , _ _ , ,, _
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Page 8 of 20 7.3.4 If a cement and water mixture (without waste) is used to solidify miscellaneous objects, this mixture will be tested for solidification prior to use. 7.4 WASTE CLASSIFICATION 7.4.1 In compliance with 10 CFR 20.311, wastes are classified as Class A, E or C, or greater than Class C, based on the presence of particular radionuclides and their activities as specified in 10 CFR 61.55. Plant procedures SP R8.713.05, " Shipment of Radioactive Materials by Exclusive Use,"and SP R2.713.06,
" Calculations for Radwaste Curie Content," or SP R3.713.02 "RADMAN Computer Program" provide the methodology for this determination as used at SNPS.
7.4.2 Waste streams will be sampled based upon the Branch Technical Position Requirements (or more frequently, if plant parameters indicate a change in waste characteristics) and analyzed for fission and activation products, including transuranics. Scaling factors developed from these complete analyses will be used with gamma spectra from each batch of waste to infer the concentrations of non-gamma emitting radionuclides). 7.4.3 During initial plant operation when the results of actual f analyses are not yet availabic, radionuclide concentrations ( 4 vill be used in accordance with Ref. 9.3.12. 7.4.4 The curie content of waste streams (such as trash) for which representative sampling is difficult may be inferred based on gamma analysis of representative smears and an external dose
- rate measurement (SP R2.713.06 or SP R3.713.02).
7.5 CONTAINER CONTROL 7.5.1 A quality assurance program shall be established to inspect the container to be used for dewatering (and solidification) using SP R2.713.30 "Radwaste Container Control." 7.5.2 This program shall assure that prior to use, the containers to be used for dewatering are intact and free of any visual damage that would prevent the dewatering of waste to required limits. 7.6 DECONTAMINATIOM
. Prior to shipeert, containers will be swiped for removable contamination and examined for general condition. Decontamination will be conducted as necessary to meet shipping requirements.
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Page 9 of 20 I 7.7 CHANGES TO THE PROCESS COUTROL PROGRAM Any changes to the Solid Waste Process Control Program for the Shoreham Nuclear Power Station shall be approved by the Review of Operations Committee and the Plant Manager and reported to the NRC in the Semiannual Radioactive Effluent Release Report. 7.8 RECORDS AND INVENTORY CONTROL 7.8.1 A Solidification Record Sheet (Appendix A) shall be completed for each liner filled for solidification. 7.6.2 A pre and post dewatering analysis sheet (see Appendix E and Appendices 12.4 and 12.5 of reference 9.1.2) shall be coupleted for each container filled with dewatered waste. 7.8.3 If more than one liner results from a batch, then the initfel liners will not be shipped until all liners for that batch have verified solidification. Those liners will be identified by a common analysis number (paragraph 7.1.6). 7.8.4 The Solidification and Dewatering Record Sheets and the attached isotopic analysis shall be forwarded to the Radwaste Engineer for retention until such time as the liner identified on the Record Sheet is shipped for final disposition. 7.8.5 When the identified liner is shipped, the Solidification ard Dewatering Record Sheets and other documents concerning the shipment shall be forwarded to SR2 for permanent record storage. 8.0 RESPONSIBILITIES The following outline departmental responsibilities and interfaces to implement and support all activities associated with the SNPS PCP. NOTE: All service contractor procedures implementing the PCP which will be used at SNPS, prior to their utilization and implementation, must be approved by the Review of Operations Committee as per SNPS Tech. Spec. 6.8.1.h. This whole process should take place after implementation of the PCP at SNPS. 8.1 NOC Policy 25 (Management of Low Level Radioactive Waste) ider.tifies
. the following departments as having direct responsibilities for the implementation, raintenance, and licensing and regulatory interface of the SNPS PCP.
8.1.1 Nuclear Engineering Department, as also described in NED 1.02.
r Page 10 of 20 8.1.2 Shoreham Operation Department, as also described in SP R1.001.01. 8.1.3 Nuclear Operations Support Department. 8.1.4 Quality Assurance Department (QA) as described in QAP-1.1 8.1.5 Nuclear Review Board (NRB) as also described in the " Charter of the Shoreham Nuclear Power Station Nuclear Review Board". See NOC Policy 25 for more details. 8.2 The Procedure NED 6.04, " Change Control" in conjunction with the
. Procedure NOSD 6. " Control of License Documents" shall be used to review, approved, control and disposition proposed changes and revisions to the SNPS PCP".
8.3 NED is responsible for preparing and maintaining the PCP current per NED Procedures 6.04 and 6.01. 8.4 The Review of Operations Committee (ROC) is responsible for reviewing and approving any changes to this program and the associated implementing procedures.
- 8.5 The Plant Manager's approval shall also be obtained for every PCP change.
3.6 ROC and Plant Manager approval signatures shall be indicated on the cover page of the PCP. 8.7 PCP implementation is accomplished through station procedures and its implementation is the responsibility of the Radiological Controls Division.
9.0 REFERENCES
9.1 LILCO OPERATING PROCEDURES 9.1.1 SP 23.710.01, Low Conductivity Liquid Radwaste 9.1.2 SP R3.710.02, Dewatering of Spent Radwaste Media 9.1.3 SP 23.711.01 High Conductivity Liquid Radwaste 9.1.4 SP 23.712.01, Pegenerant Chemical Liquid Radwaste ' 9.1.5 SP 23.713.01 Solid Radwaste System 9.1.6 SP 23.718.01, Liquid Radwaste Spent Resin 9.1.7 SP 23.719.01, Liquid Radwaste Evaporator Bottoes 9.1.8 SP 72.002.01, General Sampling Techniques 9.1.9 SP R8.713.01, Shipment of Radioactive Materials 9.1.10 SP R8.713.02, General Packaging, Marking and Labeling Requirements for Radioactive Materials for Shipment l l _--,y,-, p --- - - - - - - - -- w- --. ,
l I Page 11 of 2C l 9.1.11 SP R8.713.05, Shipment of Radioactive Materials by Exclusive Use 9.1.12 SP R2.713.06, Calculations of Radwaste Curie Content. 9.1.13 SP R2.713.24 Sampling, Treatment and Disposal of Radioactive Waste Oil 9.1.14 SF R2.713.30, Radwaste Container Control 9.1.15 SP R2.713.35, Storage of Packaged Radwaste Liners and DAW 9.1.16 SP R4.014.01, Radwaste Sample Solidification Test 9.1.17 SP 72.002.18 Radwaste Sampling for Disposal 9.1.18 SP 78.030.30, Lab pH Meter Standardization and Use 9.1.19 SP 73.033.10, Gamma Spectrometer System Operation 9.1.20 SP R3.713.02, RADMAN Conputer Program 9.1.21 SP R3.075.01, Compaction of Contaminated Waste 9.2 MOBILE SERVICE CONTRACTOR DOCUMENTS 9.2.1 STD-R-05-007, Topical Report, Cenent Solidified Wastes to Meet the Stability Requirements of 10CFR61, Westinghouse Hittman Nuclear Inc., April 1984. 9.2.2 STD-R-05-005, Waste Qualification Program Report for Cement Solidified Wastes, k'estinghouse Hittman Nuclear, Inc. , October 25, 1983. 9.2.3 F458-P-001, Process Control Program for the In-Container ( Solidification of 20-25 Weight Percent Sodium Sulfate Slurries 9.2.4 1458-P-002, Operating Procedure for Mobile Incontainer Solidification of Sodium Sulfate Slurries 9.2.5 F458-P-003, Process Control Program for Incontainer Solidification of Bead Resin - Powdered Resin Mix 9.2.6 F458-P-004, Dewatering Powdered Resin Slurries in Hittman HN-100 Steel Liners with a Three Layer Flexible Underdrain Assembly to Less Then 1/2% Drainable Liquid 9.2.7 F458-P-005, Dewatering Bead Resins Mixed with Powdered Resin in 111ttman HN-100 Steel Liners with a Three Layer Flexible Underdrain Assembly to less Than 1/2% Drainable Liquid 9.2.8 F458-Pp06,DewateringContainerResinSlurriesinHittmcr 9 -100 Container with a Three Layer Flexible Underdrain RADLOK Assembly to Less Than 1% Drainable Liquid 9.2.9 F458-P-007, Deyp'tering Bead Resin Mixed with Powdered Resin in Mittman RADLOK -100 Containers with a Three Layer Flexible Underdrain Assembly to less Than 1% Drainable Liquid 1
Page 12 of 20 9.2.10 F458-P-008 Process Control Program for the Incontainer Solidification of Pewdered Resin i 9.2.11 F458-P-009, Process Control Program for the Incontainer i Solidification of Radwaste Filter Cake 9.2.12 F458-P-010, Operating Procedure for Mobile Incontainer i Solidification of Mixed Bead Resin - Powdered Resin Slurry i 9.2.13 F458-P-011, Operating Procedure for Mobile Incontainer : Solidification of Powdered Resin Slurry 9.2.14 F458-P-012, Operating Procedure for Mobile Incontainer Solidification of 50% Powdered Resin /50% Diatomacious Earth Filter Sludge i 9.2.15 F458-P-013, Dewatering Filter Sludge Cakes in Hittman HK-100 Steel Liners with a Three Layer Flexible Underdrain Assembly to Less than 1/2% Drainable Liquid 9.3 GENERAL REFERENCES 9.3.1 NRC Standard Review Plan 11.4, " Solid Waste Management Systems" (NUREG-0800) 9.3.2 NRC Branch Technicci Position ETSB 11-3, " Design Guidance for C Solid Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor Plants", July 1981 9.3.3 Code of Federal Regulations, Title 10, Part 61, " Licencing l Requirenents for Land Disposal of Radioactive Waste" j i 9.3.4 Code of Federal Regulations, Title 49, "Transportatier." ; 9.3.5 South Carolinc Department of Health and Environmental Control. Radioactive Material License No. 097, as amended. - l 9.3.6 NRC Special Nuclear Material License No. 12-13536-01, se l amended, for Barnwell, SC. 9.3.7 State of Washington Radioactive Materials License #WN-IO19-2, as arended, for Richland, Washington. 9.3.8 NRC Special Nuclear Material License No. 16-19204-01, as amended, for Richland, Washington. , 9.3.9 ANSI /ANS-55.1/1979, American National Standard for Solid Radioactive Waste Processing System for Light Water Cooled Reactor Plants. )
Page 13 of 20 9.3.10 AIF/NESP-027, Methodologies for Classification of Low-level Radioactive Wastes from Nuclear Power Plants, Impell I Corporation, January 1984. 9.3.11 NRC Loa-Level Waste Licensing Branch, Final Waste , Classification and Waste Form Technict1 Position Papers, , Kay 11, 1983 i 9.3.12 RADMAH- Data Base Analysis Report - Shoreham Nuclear Power Station - Waste Management Group, Inc., August, 'S , 4 0 . l I _ _ _ . . . _ , - . , . . _ ~ . . . _ . _ _ . . -
Page 14 of 20 Appendix A Page 1 of 4 SOLIDIFICATION RECORD SHEET PART I Sampling and Pre-Selidification Analysis
- 1. Type of Waste
- 2. Waste tank placed on recire.
Date/ Time
- 3. Waste Tank r ampled analysis ID#
Date/ Time
- 4. Waste Stream pH !
- 5. Oil Coutent )
Less than 1% (Visual Inspe is I
- 6. Isotopic Ana cck l
- 7. Estimated Curi on (SP R2.713.06) uci/cc
- 8. Test Solidificat Required
'le s No
-9. decepttble Test Solidification performed (if required)
Initials
- 10. The abeve waste tank has beer analyzed and ic acceptable for solidification.
Radiochemistry Engineer Date or designee I l , l l l l f l l l l l i l i
Page 15 of 20 Appendix A Page 2 ef 4 SOLID!PICATION RECORD SHEET PART II Systee Preparation and Processing (Use Part lla. if vendor supplied system is used)
- 1. Container and (SP R2.713.30).
Type ID#
- 2. Contriner Properly Positioned Under Fill Pipe Check
- 3. Till Flatige Properly Mated to Container Check
- 4. Sufficient Cement Available Check
- 5. Waste Devntering Tank Level in Evap. Ecttoes Tank Leve es
- 6. Authorization c
( Date Radwaste Superv' . 9 F
- 7. Time process sta E. Time process stopped
- 9. Tachometer reading (Petering Purp)
- 10. Waste Dewatering Tank Level inches
- 11. Evap. Bottoct Tank Level inches l
l
- 12. Process Completed Operator Date Time
- 13. Waste Class: A B C i 14. Liner Chcck for free standing water Initials Date/ Time l
- 15. Liner Capped Date Tire
- 16. Container Weight lbs.
(
Page 16 of 20 Appendix A Page 3 of 4 SOLIDIFICATION RECORD SHEET PART iia Contractor System Preparation and Processing
- 1. Container ID Number Type
- 2. Applicable connections made to liner for transferring vaste and cecent to liner Initials
- 3. Connections made to liner for mixing contents, if applicable Initials
- 4. Process parameterc Waste to be added to liner .
Cerent to be added to liner . Water to be added to lin ft
- 5. Authorization t Radweste Supervis Time
( jg1pdTe
- 6. Time processing staked
- 7. Time processing stopped
- 8. Waste Class Class A Class B Class C
- 9. Container checked for free standing water Initials Date/ Time
- 10. Liner capped Date Tire
- 11. Liner Weight lbs.
Page 17 of 20 Appendix A Page 4 cf 4 SOLIDIFICATION RECORD SEEET PART III Filled Liner Data
.1. Filled Liner Radiation levels
- a. Centact Dose Rete 1 Meter Dose Rate mr/hr ar/hr
- b. Smeerable Activity 2 4 Quadrants 1 dpm/100 cm 2
2 dpn/100 cm 2 3 dpm/100 cm 2 4 dpm/100 e
- c. Liner decon performed
- d. Smearable if formed) dpm/100cn
- e. Liner rea transfer to storage n
Health PhysicF' Supervision Date Time (
- 2. Storage Location Radwaste Supervision Forward this document to Radwaste Shipping Folder (per SP R8.713.01)
I i l (
i Page 18 of 20 I [ Appendix B Page 1 of 2 l DEWATERING RECORD SHEET PART I Sampling and Analysis
- 1. Type of Waste
- 2. Waste Tank (or Linar) Sampled Analysis ID!
Date/ Time
- 3. Oil Content (Visual Inspection)
Verify Less than 1% _ Initials
- 4. Isotopic Analysis Attache
- 5. The above waste a a scepled and found to contain the isotopes a nd ated on the attached data sheets.
O e1 cs > pr1- - l I I i
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Page 19 of 19 i-
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Page 19 of 20 Appendix 5 Page 2 of 2 DEWATERING RECORD SHEET PART 11 Dewatered Container Data
- 1. Container and Type ID#
- 2. Container Radiation Levels
- a. Contact Dose Rate Mr/hr 1 Meter Dose Rate Mr/hr
- b. Smearable Activity 2
4 Quadrants 1 / dpm/100 c 2 a M adpm/100 cm 22 3 ope /100 cm 3 100 cm
- c. Contain d Yes/No 2
- d. S washdown/decon dmp/100cm
- e. Li tipping or transfer to storage
( - Healt sics Supervision Date Tin.e
- 3. Waste Class A B C Initials
- 4. StoraEe Location Radwaste Supervision l
Forward this document to Radwaste Shipping Folder (per SP R8.713.01). l l . l l t . _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ .
Attachment 2 1 l OFFSITE DOSE CNLCULATION MANUAL Revision 8 - February 1986 i Insertion Instructions Remove Old Pace (s) Insert New Pace (s) Title Page Title Page Signature Page Signature Page , EP 1 thru EP-2 EP-1 thru EP-2 1- 1 1- 1 2.1-1 2.1-1 2.1-3 thru 2.1-7 2.1-3 thru 2.1-7 3.1-1 thru 3.1-5 3.1-1 thru 3.1-5 4 4- 1 4- 1 Appendix A Appendix A l l l
OFFSITE DOSE CALCULATION MANUAL List of Effective Pages Page. Table (T), Revision or Figure (F) Number EP-1 5 EP-2 , 8 i 3 111 thru iv 3 v thru vii 4 viii . 1 1-1 8 2.1-1 8 2.1-2 2 2.1-3 8 2.1-4 8 2.1-5 8 2.1-6 thru 2.1-7 8 F2.1-1 thru F2.1-4 3 2.2-1 1 2.2-2 thru 2.2-5 3 2.2-6 thru 2.2-8 6 T2.2-1 1 F2.2-1 3 F2.2-2 thru F2.2-4 1 F2.2-5 thru F2.2-6 7 4 3.0-1 3.1-1 8 3.1-2 8 3.1-3 thru 3.1-5 8
- 3.1-6 1 T3.1-1 3 T3.1-2 3 F3.1-1 thru F3.1-2 1 3.2-1 1 1
F3.2-1 4 3.3-1 3.3-2 thru 3.3-4 3 3.3-5 thru 3.3-6 6 3.3-7 4 6 3.3-8 1 F3.3-1 4 3.4-1 3.4-2 3 3.4-3 6 3.4-4 3 3.4-5 6 1 T3.4-1 3.5-1 4 3.5-2 thru 3.5-6 3 3.5-7 4 3.5-8 thru 3.5-9 3 4 3.5-10 3.5-11 3 EP-1 Revision 8 - February 1986
OFFSITE D05E CALCULATION MANUAL List of Effective Pages (Cont'd.) Page, Table (T), R vision or Figure (F) Number 3.5-12 , 6 3.5-13 3 T3.5-1 1 T3.5-2 1 Ti.5-3 1 T3.5-4 1 T3.5-5 3 T3.5-6 1 T3.5-7 1 T3.5-8 4 , T3.5-9 3 T3.5-10 3 T3.5-11 . 3 T3.5-12 3 T3.5-13 3 T3.5-14 3 T3.5-15 3 T3.5-16 3 T3.5-17 4 3.6-1 - 1 F3.6-1 1 3.7-1 2 4-1 8 4-2 thru 4-3 4 T4-1 4 T4-2 4 5-1 1 T5-1 4 T5-2 4 TS-3 3 T5-4 4 T5-5 6 F5-1 3 F5-2 6 6-1 3 A-1 3 B-1 3 EP-2 Revision 8 - February 1986
l l LONG ISLAND LIGHTING COMPANY OFFSITE DOSE CALCULATION MANUAL
\
Revision 8 - February 1986
l
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V TITLE OF DOCUMENT Shoreham Nuclear Power Station-Unit 1 Offsite Dose Calculation Manual l NOll. B
SNPS-1 ODCM SECTION 1 INTRODUCTION The purpose of thir manual is to show the calculational methodology and parameters used to comply with the Radiological Effluent Technical Specifications (RETS). , Section 2 establishes methods to calculate the Liquid Effluent Monitor set point and the Gaseous Effluent Monitor set points in order to comply with RETS Sictions 3.11.1.1'and 3.11.2.1, respectively. Section 3 establishes dose calculational methods for liquid and gaseous effluents. The liquid effluents dose calculation methods are used to show compliance with RETS Sections 3.11.1.2 and 3.11.1.3. For liquid pathways, the. dilution factor of 8.85 used in Section 3.1 is a calculated value based on a , submerged, multiport diffuser with a port discharge velocity of 12 fps, a 300 ft radius mixing rone, and 4 circulating water pumps discharging. If only service water pumps are discharging, the dilution factor is one (1.0). The gaseous effluent dose calculation methods are used to show compliance with RETS Sections 3.11.2.1, 3.11.2.2, and 3.11.2.3. The atmospheric dispersion and deposition factors used in calculation methods were calculated based on onsite meteorological data for the 2-year period,of October 1, 1973 through September 30, 1975. Regulatory Guide 1.109 Rev. 1 (October 1977), Methodology and Parameters, with the exception of the dilution factor of 8.85, when circulating water flow exists, were used in Method 2 (the Backup Method) dose rate and dose conversion factors. Tables 3.5-10, 3.5-12 and 3.5-13 are incorporated only for future use if there is a change in the land use census which requires considering any ccmbination of cow's milk and meat pathways. Section 4 identifies the receptor locations which represent critical pathway locations, water dilution, atmospheric dispersion, and deposition factors used in calculation Method 2. Table 4-1 summarizes the above factors for the gaseous effluent pathways. Sectien 5 indicates locations at which environmental sampling may be conducted. Section 6 addresses the Interlaboratory Comparison Program. l l l l BN1-11600.02-92 1-1 Revision 8 - February 1986 i l
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SNPS-1 ODCM SECTION 2 SET POINTS 2.1 LIQUID EFFLUENT MONITOR SET POINTS (Compliance with Section 3.11.1.1 of the Radiological Effluent Technical Specification (RETS)) The radionuclide concentrations released via liquid effluents to unrestricted areas shall be limited to the concentrations specified in 10CFR20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved4 r entrained n ble gases, the total concentration shall be limited to 2 x 10 Ci/ml. The set points of the effluent monitors are dependent on circulating or service water as follows:
- 1. a. With the circulating water system (a once-through system) in use, the circulating water flow rate (the circulating water system is composed of four pumps and circulates sea water at a rate of 574,000 gpm).
- b. The service water flow rate, if the circulating water system is not in use but service water is in use. (The service water system is composed of four reactor building service water pumps, each having a capacity of 8600 gpa and three turbine building service water pumps each having a capacity of 8000 gpm.)
- 2. Flow rates of effluents from tanks and/or from the RHR heat exchanger service water outlet, and/or yard piping drain sump.
- 3. Individual concentrations of gamma emitters (other than dissolved or entrained noble gases) and Sr-89, Sr-90, Fe-55, and H-3; and the total concentration of dissolved or entrained noble gases and gross concentration of the alpha emitters in the liquids to be discharged.
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- 4. Maximum allowable concentration of 2 x 10 C1/ml for the total concentration of dissolved or entrained noble gases and maximum permissible concentrations (MPCs) of other gamma emitters, Sr-89, Sr-90, Fe-55, H-3, and alpha emitters in the effluents as specified in 10CFR20, Appendix B. Table II, Column 2 for an unrestricted area.
NOTE: Precautions should be taken to assure that the circulating water system flow rate or the service water system flow rate used in determining the set point remains constant during the period of discharge. If the circulating or service water flow rate during discharge becomes less than the flow rate that was used in calculating the discharge set point, the discharge must be terminated and a new set point calculated. Service water via the RER heat exchanger service water outlet will be released continuously to the environment when the RHR heat exchanger is in operation. Reactor building salt water drain tank contents may be released to the environment either as a batch process or continously. The discharge waste sample tanks, recovery sample tanks, and yard piping drain sump contents will always be released to the environment as batch processes. l BN1-11600.02-92 2.1-1 Revision 8 - February 1986
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i I SNPS-1 ODCM i The above methodology will ensure that a representative sample will be obtained for batch releases. l Set Point Philosophy The philosophy of the set points will be based on the sum of the ratios of isotopic concentrations to MPCs being less than 1 for discharges into unrestricted areas. Specifically:
.; C i M C " i li MpC i
C, Cg a b n *
" MFC a
+ E +b ***** + MPC +n EC" #a WPCC s t Fe (2.1-1)
+ ypes + gp t + FC- Fe 11 where: ,
C ,C ,......C = Concentration of the individual gamma emitting a b n radionuclides identified (MC1/ml) C = The gross alpha concentration (pCi/ml) , Cg = The total concentration of dissolved or entrained noble gases (pCi/ml) i
- The Sr-89 and Sr-90 concentrations (pC1/ml)
C,. ' C = The H-3 concentration (pCi/ml) C = The Fe-55 concentration (pci/ml) Fe MPC g = WC,, NCb ****** n' o' G' s' t' Fe
= the maximum permissible concentration of the respective radioisotope i (UCi/ml) from 10CFR20, Appendix B. Table II Column 2. For dissolved or entrained noble gases, the maximum allowable concentration (MPC g ) will be 2.00E-04
. (pCi/ml). For gross alpha, the MPC assumed will be 3.00E-08 (pCi/ml).
then no release is possible. The If the C[MPC calculated is greater than 1, normalization factor (as defined in Section 2.1.1) must be greater than 1 to i permit releases. To permit releases, this factor can be increased i to a value l ting greater or service than water 1 bypumps increasing in the dilution flow Fapplicablecharge structure), dis (byand/or running more c rcu decreasing theeffluent flow rates f , f , f ,etc. (defined in Section 2.1.1), and recalculate C/MPC using D new O 1NA, fEqu!kion2.1-1. g BNI ,ll600.02-92 2.1-3 Revision 8 - February 1986 5 e
SNPS-1 ODCM 2.1.1 Radiation Effluent Monitor (RE-13) High/ Trip Alarm Set Point for Discharge Waste Sample Tanks, Recovery Sample Tanks, or Yard Piping Drain Sump The function of this monitor set point is to ensure that the sua of the ratios of the discharge concentrations to the NPCs of the corresponding radionuclides of the discharges monitored by this monitor and, other liquid waste discharges, if any, does not exceed 1. If the monitor count rate is higher than the calculated set point, the radiation monitor will terminate the release. Msample is taken from any of the following tanks or sump which is to be discharged along with any streams which are in the process of being discharged.
- 1. Discharge waste tanks ,
- 2. Recovery sample tanks
- 3. Yard piping drain sump
- 4. Reactor building salt water drain tank
- 5. Residual heat removal heat exchanger service water Only one of the first three items above is discharged at any one time, which can be combined with releases from item 4 and/or 5.
Obtain the circulating or service water flow rate from the control room (see NOTE in Section 2.1). Define Normalizing factor [fp+fMA + fRB + fs + (F -f RA _fD)] *0.8
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F = . (C DifD f f f [ g
+ biA HA + HiB HB + Csi s))
i=1 MPCg An isotopic analysis of each sample is performed. This analysis includes isotopic analysis for gamma emitters; gross alpha emitters; total dissolved or entrained noble gases; and Sr-89, Sr-90, Fe-55, and H-3. This should be done for all monitors. Then the set point (NOTE: the background (cpm), if it can be determined, is also added to the set point value. If, however, it cannot be determined, it is con-t sidered as zero) for detector RE-13 is calculated as: , 1 ! N S33 < F* 1 CDi*Ei (cys) g3 l l L l BN1-11600.02-92 2.1-4 Revision 8 - February 1986
SNPS-1 ODCM ! where: f C = c ncentration of radioisotope (i) (pCi/ml) in any of the following Di tanks or sump that is to be discharged:
- 1. discharge waste tanks ,
- 2. recovery sample tanks
- 3. yard piping drain sump
= Discharge flow rate (gpm) from any of the following tanks or sump that f*f is to be discharged:
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- 1. discharge waste tanks
- 2. recovery sample tanks ,
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- 3. yard piping drain sunp.
(Maximum design discharge flow rate = 150 gpm) C gg = Reactor building salt water drain tank concentration of radioisotopc(i) (UCi/ml) C = RHR heat exchanger service water outlet concentration of radioisotope (i) Hi/' (uci/ml) from loop A. C = RHR heat exchanger service water outlet concentration of radioisotope 4 HiB (i) (pCi/ml) from loop B. ! f = Reactor building salt water drain tank discharge flow rate (gpm). g 3 (Maximum design discharge flow rate = 100 gpm) i f = RHR heat exchanger service water outlet discharge flow rate (gpm) froc l RA loop A (Maximum design discharge flow rate = 9340 gpm) 7 f = RHR heat exchanger service water outlet discharge flow rate (gpm) from RB loop B (Maximum design discharge flow rate 9340 gpm) F = Total circulating or service water flow rate (gpm) (this includes FHA ""d C HB) E g
= Gamma counting efficiency of RE-13 for radionuclide (i) (cpm /pci/ml).
Figure 2.1-1 shows the energy response. For non-gamma emitters, E g=0. 0.8 = Safety factor MPC, is defined in Section 2.1. The above calculation is made for each batch to be rellased. l After each batch release, the high alarm set point should be reset as close to the background as practical to prevent spurious alarms and yet assure an alarm should an inadvertent release occur. BN1-11600-02-92 2.1-5 Revision 8 - February 1986 i I
SNPS-1 ODCM 2.1.2 Radiation Effluent Monitor (RE-79) High Alarm Set Point for Reactor Building Salt Water Drain Tank The function of this monitor set point is to ensure that the sum of the ratios of the discharge concentrations to the MPCs of the corresponding radionuclides of the discharges monitored by this monitor and other liquid waste discharges, if any, does not exceed 1. If the monitor count rate is higher than the calculated set point, the radiation eenitor will alarm in the control room.
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A sample will be taken from the reactor building salt water drain tank discharge, along with individual samples of any of the following streams which may be in the process of being discharged:
- 1. Discharge waste sample tanks
- 2. Recovery sample tanks
- 3. Yard piping drain sump
- 4. Residual heat removal heat exchanger service water In the case of continuous release, samples will be taken as per requirement RETS Table 4.11.1.1.1-1.
Obtain the circulatirk or service water flow rate from the control room (see NOTE l in Section 2.1). The set point for cr>ntinuous or batch release (see NOTE in Section 2.1.1) will be calculated as follous: N Syg i F*[{,C3 E] (cpe) where: Eg = Gamma counting efficiency of RE-79 for radionuclide i (cpm /uC1/ml) . Figure 2.1-2 shows the energy response. For non-gamma emitters, E =0 All other parameters are as defined in Section 2.1.1. When the tank operates in a batch mode, the above calculation is made for each
, batch to be released.
Af ter each batch release or continuous release period, the high alarm set point should be reset as close to the background as practical to prevent spurious alarms and yet assure an alarm should an inadvertent release occur. BN1-11600-02-92 2.1-6 Revision 8 - February 1986
SNPS-1 ODCM 2.1.3 Residual Heat Removal Heat Exchanger Service Water Outlet Monitors (RE-23A, RE-23B) High Alarm Set Points The function of this monitor set point is to ensure that the sum of the ratios of the discharge concentrations to the MPCs of the corresponding radionuclides of the discharges monitored by this monitor and other liquid waste discharges, if any, does not exceed 1. If the monitor count rate is higher than the calculated set point, the radiation monitor will alarm in the control room. , Monitors RE-23A and RE-23B are independent. Each is dedicated to monitor its r'espective RHR loop. A sample will be taken from the RHR heat exchanger service water outlet (A and/or B), along with individual samples of any of the following streams which may be in the process of being discharged: -
- 1. Discharge waste sample tanks
- 2. Recovery sample tanks
- 3. Yard piping drain sump
- 4. Reactor building salt water drain tank discharge Obtain the circulating or service water flow rate from the control room (see NOTE in Section 2.1).
The set points for RE-23A and RE-23B are calculated as follows: 5 23A i1 BiA LA 235 b 11 bib iB where: E g
= Gamma counting efficiency of RE-23A for radionuclide 1 (cpm /pC1/ml).
Figure 2.1-3 shows the gr.mma energy response. For non-gamma emitters, E =0 E " 8mm8 C un nB e Cien y f - f r ra nuc e (cpm M /mD. iB Figure 2.1-4 shows the gamma energy response. For non-gamma emitters, E =0 iB All other parameters are as defined in Section 2.1. i BN1-11600.02-92 2.1-7 Revision 8 - February 1986
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SNPS-1 ODCM 3.1 LIQUID EFFLUENT, DOSE CALCULATION To comply with Section 3.11.1.2 of the RETS, the liquid effluents released to unrestricted areas shall be limited:
- 1. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ.
- 2. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.
The site boundary for liquid effluents is shown in Figure 3.1-1. The liquid radwaste system model is shown in Figure 3.1-2. 3.1.1 Method 1: (computerized Method) The equations which follow are used by the computer software to calculate the offsite doses due to release of liquid radwaste. For this dose calculation the actual concentration to be discharged by isotope, the total volume of liquid to be discharged, and the number of circulating water pumps running, supplied by the operator, shall be used. The scftware computes the isotopic releases by multiplying the lab measurements by the volume of the liquid to be released:
-3 y q Qg = 3.785 x 10 where:
Q = total inventory of isotope i in the liquid to be released (Ci) q = concentration of isotope i in the liquid to be discharged (as f measured in laboratory) (UC1/cc) V = volume of liquid to be discharged (gallons)
-3 3.785 x 10 = [(Ci/UCi) (cc/ gal)]
The dose equations which follow are from Regulatory Guide 1.109, with minor modifications. They are employed for the computation of dose from any single batch discharge (continuous discharges are handled as batch discharges in the computerized method). Weekly and quarterly cumulative doses are also calculated and stored in data files for reporting. (a) Organ Dose due to Ingestion of Saltwater Fish R ing Hsh ns
= 0.389 Uap (3jg1 ) 7 q 3 Hs DFI e
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i for cire, water ja part+I I ip ija n E R ,g Hsh = 57.4/(K2 + 0.930 K3 ) U, I 10 8 part+I for service water BN1-11600.02-92 3.1-1 Revision 8 - February 1986
SNFS-1 ODCM where: R 8
$" = dose to organ j of individual in age group a due to ingestion of fish contaminated with particulates and radio-iodines (mrem) (Ref. Reg. Guide 1.109, with the following special values:
oF(flowrateof}iquideffluent)isrepresentedbyproduct K *F [ft /sec], where K is the number of pumps of systemgaggjratingandF P Ss the flow rate per unit pump. For the circulating watef"E[dPem, m=1; for the reactor building service water system, a=2; for the turbine building service water systen, m=3 3 oF = 1.435x105 [gpm] = 319.7 [ft /sec] oF Pump 1 = 8,600 [gpm] = 19.16 [ft3 /sec] . 3 oF Fu2P,2 = 8,000 [gpm] = 17.82 [ft /sec]. o MP 7Eixing ratio at the point of exposure) = 0.113 = (1/8.85) iE circulating water is in use;
= 1.0 if service water is in use.
oD see below) nsittimerequired$$r(nuclidestoreachthepoint o t'IThra(dose factor) =i DFI oE exposure) = 24 [hr] (see pg 1.109-12 of the Regulatory Guide) s = fish consumption rate by individual in age group a [kg/yr] U ap (from Table E-5 of the Guide, for maximum individual) K, = number of pumps of system m operating Qg = total activity of isotope i released [Ci], from above
= Bioaccumulation factor for saltwater fish [(pCi/kg)/
(pCi/ liter)] (from Table A-1 of the Guide) DFI = dose conversion factor for nuclide i to organ j of I$" individual in age group a due to ingection [ mrem /pCi ingested] (from Tables E-11 through E-14 of the Guide)
= radionuclide decay constant [hr- ] (from Table 3.1-1) k 0.389 = 1100 M /F = 1100 x 0.113 / 319.7 [(pci/t)/(Ci/yr)] for cire. EatgymP,1
= 1100/F mp.2 = 1100/19.16 [(pci/t)/(Ci/yr)] for reactor building
- 57.4 service water 0.930 =F pump,3 pump.2 part+I = 68 particulates and 5 iodines in the summation sign (b) Organ dose due to ingestion of saltwater invertebrate
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R 8 I"Y = 0.389 U (1/Kg ) part+1 $" R 8 " 91 DFIg ),e l
= 57.4/(K2 + 0.930 K3 ) a ,g,7 for service water BN1-11600.02-92 3.1-2 Revision 8 - February 1986 l
SNPS-1 ODCM where: R "E = dose to organ j of individual in age group a due to the d* ingestion of saltwater invertebrate contaminated with radioactive particulates and iodines [ mrem] (Ref.: Reg. Guide 1.109 Eq. A-3 with the special values identified in the fish-ingestion equation above) B"I
= Bioaccumulation factor for saltwater invertebrate [(pC1/kg)/
iP (pCi/ liter)] (from Table A-1 of the Guide) U"ap
= invertebrate consumption rate by individual in age group a
[kg/yr] (from Table E-5 of the Guide, for maximum individual) . (c) Total body dose from shoreline deposits
-t A'-) for cire, water R ," = 0.561 U (1/Kg ) E QgDFG g ( ""
= 82.9/(K2 + 0.930 K3 ) U ** E Qg DFGgy ( service water part+I A g
where: 8 #" R = total body dose to individual in ags group a from shoreline wb.a deposits [ mrem] (Ref.: Reg. Guide 1.109 Eq. A-7 with the following special values: o F(flow rate cf liquid gffluent) is represented by the product K *F ft /sec], where K is the number of operating"puEpjP g s[ystem m and F is the flow rate per p unit pump - For the circulating water system, m=1; for the reactor building service water system, m=2; for the turbine building service water system, m=3 oF = 1.435 x 105 [gpm] = 319 7 [ft3 /sec] Pump,1 oF = 8,600 [gpm] = 19.16 [ft3/sec] P 2 = 8,000 [gpm] = 17.82 [ft /sec] oF"'gg'gingratio)=0.113ifcire.waterisinuse oMPuy P = 1.0 if service wcter is in use o W (shore-width factor that describes the geometry of the , exposure) = 0.5 for ocean site (from Table A-2 of the l guide) ot (transit time from source to shoreline) = 0 (see Reg. GEide pg 1.109-69, for Eq. A-7) oT g (radionuclide half life, days) = 0.693/(24),)wgere 0.693 = log,2 and A, is the decay constant in Ihr ] oD = DFGg (see oelow)
= shoreline exposure time for individual in age group a
[hr/yr] (from Table E-5 of the Guide, for maximum
- individual)
BN1-11600-92 3.1-3 Revision 8 - February 1986
SNPS-1 ODCM K, = number of pumps of system a operating Qg = total activity of isotope i released [Ci], from above DFG = total body conversion factor for standing 2 on contaminated il ground (shore) [(ares /hr)/(pCi/m )) (from Table E-6 of the guide) e t b = time period oger which accumulation is evaluated (15 years, or 1.314 x 10 hours) O.561 = 110,000 M W (Tg1A )/ Fpy,p y [(pCi/1)/(Ci/yr)] p
= 110,000 x 0.113 x 0.5 x (0.693/24)/ 319.7 for cire water 82.9 = 110,000W (Tgg A)F 2 [(pC1/E)/(Ci/yr)]
= 110,000 x 0.5 x (0.tJ3/24)/19.16 for reactor building service water 0.930 =F pump.3/Fpump.2 part+I = 68 particulates and 5 iodines in the summation sign (d) Skin dose from shoreline deposits 8 **
R = 0.561 U p
** (1/K3) E Qg DFG12 (
) f r cire, water part+I A g
-t A I )
= 82.9/(K2 + 0.930 K3) U ap Q1 DFG12 (
part+I i I for service water where: 8 R sMn,a
#* = skin dose to individual in age group a from shoreline deposits [ ares]; (Ref.: Reg. Guide 1.109 Eq. A-7 with the special values listed for the total body dose)
DFG ,, 1'
= skin dose conversion factor for gtanding on contaminated ground (shore) [(arem/hr)/(pCi/m )) (from Table E-6 of the Guide)
Other parameters are as defined earlier for the total body dose from shoreline deposits. (e) Total doses The individual dose components described in items (a), (b), (c), and (d) above are summed in the following way for the computation of total doses: R ja = Rja i
"E fi*h + R'"E ja '""
ng E "# ** R =R vb.a wb.a sh + R w "b.a + R*b.a w shore skin,a
=R skin,a BN1-11600-92 3.1-4 Revision 8 - February 1986
SNPS-1 ODCM where: R = total dose to organ j (exclusive of the total body) of d# individual in age group a due to the ingestion of fish and invertebrate (mrem) R = t tal dose to the total )ody of individual.in age group a "b'" due to the ingestion of fish and invertebrates, and direct radiation from shoreline deposits and R = total dose to the skin of an individual in age group a from skin,a shoreline deposits (mrem). 3.1.2 Method 2: (Backup Method) , The dose contributions for the total release period shall be calculated for all radionuclides identified in liquid effluents released to unrestricted areas using the following expression: N M D7 = y1 [Ag yy AtgC gFg] (3.1-1) I where: D = the cumulative dose or dose commitment to the total body or an 7 organ from the liquid effluents for the total release M (mrem), period g[1 At g At g = the length of the ith release period over which Cgg and Fg are averaged for all liquid released (minutes), C gg = the average concentration of radionuclide C f in undiluted liquid effluent during release period Atg from any liquid release (pci/cc). Ag = the site-related ingestion dose or dose commitment factor to the total body or any organ for each identified principal gamma and beta emitter listed in Table 3.1-2 (mrem / min per pCi/ce), see Appendix A for derivation Fg = Undiluted liquid effluent flow rate F /M F c
= totalcirculaEin$waterflowratewiththenumberofcirculating pumps in use
= total service water flow rate if the circulating water is not in use MP (Mixing factor) == 1.0 0.113 if circulating water is in use; if service water only is in use BN1-11600-92 3.1-5 Revision 8 - February 1986
I SNPS-1 ODCM SECTION 4 i METEOROLOGICAL AND HYDROLOGICAL PARAMETERS UTILIZED IN THE CALCULATION OF DOSES
4.1 INTRODUCTION
This section specifies the liquid pathway dilution factor and the dispersion and deposition factors utilized for atmospheric releases. A description is given of the meteorological methodology and parameters 4 utilized in the computerized method for atmospheric release. Critical locations for receptors and their respective dispersion and deposition factors are provided for the backup method for atmospheric releases. For liquid effluent pathways a calculated dilution factor of 8.85 is , used if circulation water is utilized. If service water is in use the , dilution factor is one (1.0). 4.2 PARAMETERS AND METHODOLOGY USED IN THE COMPUTERIZED METHOD 4.2.1 Meteorological Data Hourly average values (based upon 60 one-minute values) of temperature, , wind speed, wind direction and temperature difference from the 33- and 150-ft levels of the Shoreham meteorological towers are used in the computerized method, to determine.X/Q and D/Q values at the locations given in Table 3.5.-8. 4.2.2 Long-Term X /Q and D/Q Values Sector-average atmospheric coneggtration dispersion factors (x/Q) A , gamma dispersion factors (X/Q) and relative deposition factors (D/Q) are calculated every hour using 60 one-minute meteorological data values obtained from the meteorological towers. The methodology utilized is described in the report "Shoreham Nuclear Power Station EMSP Software (Rev. B.1)" (Entech Engineering Inc., P104-R3, Section 2.0, July 1983, by J. N. Hamawi). General site specific data values that may be required for the calculation of dispersion parameters are given in Table 4-2. The basic methodology used to obtain the (X/Q) ^ and D/Q values is the straight-line trajectory model with Gaussian dispersion described in Regulatory Guide 1.111, Rev. 1. The list of selected options and variations from the Regulatory Guide is as follows: (a) Plume depletion due to dry and wet depositions, as well as to en-route radioactive decay is conservatively ignored. BN1-11600.02-112 4-1 Revision 8 - February 1986
SNPS-1 ODCM APPENDIX A DERIVATION OF A 7 A p$us(arem/minperyCi/cc)isthedoseconversionfactorforthecombinedfish seafood pathways due to a liquid radwaste system discharge. The doses to an organ, due to ingestion of fish and, seafood (contribution from shoreline deposit is considered insignificant) containing isotope, 1, were calculated by a computer code based on Regulatory Guide 1.109, Rev. 1 methodology and default parameters. The computer isotopic dose rates output were normalized to unit intake concentration with the following equation: A II
=
~-
Fi + si Cf /F where: 1 D = Calculated fish ingestion dose rate (mrem / min) to an organ, from Fi isotope, 1, (Ref. Reg. Guide 1.109, Eq. (A-3) assuming a dilution factor of 8.85. D,f = Calculated seafood ingestion dose rate (mrem / min) to an organ, from isotope, i, (Ref. Reg. Guide 1.109 Eq. (A-3)) assuming a dilution factor of 8.85. Cg = Discharge concentration of isotope, i (pCi/cc) F = Near field dilution factor, 8.85 (unitiess) i I j BN1-11600.02-112 1 of 1 Revision 8 - February 1986
ggg _ _ _ _ . . _ E LONG ISLAND LIGHTING COMPANY SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY ROAD e WADING RIVER N.Y.11792 JOHN D. LEON AR D, JR. VICE PRESIDENT. NUCLEAR OPERATION 3 AUG 281986 SNRC-1275 Dr. Thomas E. Murley Regional Administrator Office of Inspection and Enforcement Region 1 U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Semiannual Radioactive Effluent Release Report Shoreham Nuclear Power Station Docket No. 50-322
Reference:
Facility Operating License NPF-36 (Shoreham)
Dear Dr. Murley:
Enclosed is a copy of our Semiannual Radioactive Effluent Release Report covering the period from January 1, 1986 through March 31, 1986 and April 1, 1986 through June 30, 1986. As specified in Technical Specification 6.9.1.7, this report includes information for each type of solid waste shipped offsite and changes made to the Offsite Dose Calculation Manual during the period. It also includes an up-to-date copy of the Process Control Program. If you require additional information, please contact this office. Very truly yours, f( [] ( n 9 /l / ('
,WGQ jul4 2 Jo n D. Le6tiar J r . /* Q Vi e President - Nuclea Operations N
GP:ck Enclosure cc: R. Lo J. A. Berry Document Control Desk [/[,,}}