ML20147A838
| ML20147A838 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 12/31/1987 |
| From: | DAIRYLAND POWER COOPERATIVE |
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| Shared Package | |
| ML20147A807 | List: |
| References | |
| NUDOCS 8801150078 | |
| Download: ML20147A838 (38) | |
Text
i gl si SUPPLEMENT TO THE ENVIRONMENTAL REPORT FOR THE LA CROSSE BOILING WATER REACTOR (LACBWR)
POST-OPERATING LICENSE STAGE SAFSTOR l
I' December 1987 8801150078 871221' PDR ADOCK O 9
p i'.~...---
o TABLE OF CONTENTS Page No.
1.0 Introduction 1
2.0 Selection of SAESTOR 3.0 Onsite Storage of Fuel 3
4.0 Plant Status and Activities 4
5.0 _Changt in Land Use 4
5.1 Effect on Historical Sites 4
4 5.2 Manpower 6.0 Thermal Effluents 5
7.0 Environmental Effects During SAFSTOR 5
7.1 Effluent Release (10 CFR 50 Appendix I Limits) 6 8
7.2 Liquid Releases 7.3 Gaseous Releases 13 7.4 Solid Radioactive Waste Processing and Shipments 17 7.5 Offsite Dose Calculation Manual (ODCM) 19 7.6 Estimated Doses to Members of the Public During SAFSTOR 19 7.7 Nonradioactive Wastes 20 7.8 Comparison of Expected SAFSTOR Keleases During Plant 20 Operation 8.0 Radiological Environmental llonitoring Program 23 9.0 Potential Accidents During SAFSTOR 25 9.1 Introduction 25 25 9.2 Spent Fuel Handling Accident l
9.3 Shipping Cask or Heavy Load Drop into FESW 28 l
9.4 Loss of FESW Cooling 29 f
I December 1987 f
ER-SUPP w
_. -~
TABLE OF CONTENTS - (cont'd)
Pane No.
29
~ 9.5-FESW Pipe Break-9.6 Uncontrolled Waste Water Discharge 30 9.7 Loss-of Offsite Power 30 P
9.8
' Seismic Event 31 9.9 Wind and Tornado 31 7
32 9.10 References 32 10.0 Regulatory Guide 1.86 I'
l 4
l 11 December 1987 ER-SUPP t
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s LIST OF TABLES Pane No.
-Table 7 Total Fission and Activation Product Activity _(Excluding H-3) Roleased in Liquid 11 Effluents.
~'
Table 7-2 Total H-3 Activity Released in Liquid
~
Effluents..
11 Table 7-3 Major Radionuclide Breakdown in Liquid 12 Releases Table 7-4 Calculated Doses to Members of the Public 13 from Liquid Effluent Releases.
Table 7-5 Total Fission and Activation Froduct Activity Released in Gaseous Effluents 15 Table 7-6 Major Radionuclide Breakdown ~in Gaseous Relcases 16 Table 7-7 Ca'1culated Doses to Members of the Public from Gaseous' Effluent Releases 17 Table 7-8 Low Level Rad l3 active Waste Shipments from 18 LACBWR During Jperation.
Table 7-9 Project-2 Low Level Radioactive Waste 10 Shipmencs from LACBWR During SAFS10R i
Table 8-1 Radiological Environmental Monitoring Program Results Summary'(July-September 1987).
24 l
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L D. comber 1987
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LIST OF FIGURES Pane No.
1 Figure 7.1 Potential Radiation Exposure Pathways and Average Annual Doses to Members of the Public for Liquid Releases from LACBWR 21 During SAFSTOR Figure 7.2 Potential Radiation Exposure Pathways and Average Annual Doses to Members of the Public for Gaseous Releases from LACBWR l
22 During SAFSTOR iv December 1987 ER-SUPP
1.0 INTRODUCTION
This is the Supplement-to the Environmental Report for the La Crosse Boiling Water Reactor (LACBWR) covering the post-operating license period while the reactor is in the SAFSTOR condition.
The Environmental Report was initially dated September 1972 and was submitted to the Atomic Energy Ccmmission on December 8, 1972, in support of the planned application for full-term operating authorization for LACBWR. The Nuclear Regulatory Commission (NRC) issued the Final Environmental Statement related to operation of the La Crosse Boiling Water Reactor in April 1980 as NUREG-0191.
The La Crosse Boiling Water Reactor achieved initial criticality on July 11, 1967. On April 30, 1987, LACBWR was permanently shut down. Final defueling was completed June ll, 1987.
This supplement will address any significant environmental changes asscciated with the shutdown of the La Crosse Boiling Water Reactor and maintenance of the plant in a SAFSTOR condition.
The Decommissioning Plan for the La Crosse Boiling Water Reactor is being submitted to the NRC at the same time as this supplement. An additional Environmental Statement Supplement will be submitted with the detailed DECON Plan, prior to r'.ie end of the SAFSTOR period. This supplament, therefore, considers only +.he SAFSTOR decommissioning activities.
2.0 SELECTION OF SAFSTOR The Nuclear Regulatory Commission (NRC) proposed rule on Decommissioning Criteria for Nuclear Facilities identifies 3 major classifications oi-decommissioning alternatives.
They are DECON, SAFSTOR, and ENTOMB. The proposed rule defines the alternatives as follows:
DECON is the alternative in which the equipment, structures, and portions of a facility and site containing radioactive contaminants are removed or decontaminated to a level that permits the property to be released for unrestricted use shortly after cessation of operations.
SAFSTOR is the alternative in which the nuclear facility is placed and maintained in such condition that the nuclear facility can be safely stored and subsequently decontaminated (deferred decontamination) to levels that parmit release for unrestricted use.
ENTOMB is the alternative in which radioactive contaminants are encased in a structurally long-lived material, such as concrete. The entombed structure is appropriately maintained and continued sarveillance is carried out until the radioactivity decays to a level permitting unrestricted release of the property. This alternative would be allowable for nuclear facilitias contaminated with reistively short-lived radiorsclides such that all contaminants would deca'; to levels permissible for unrestricted use within a period on the ordet of 100 years.
ER-SUPP 1
December 1987
e For a power reactor the choice is either DECON or SAFSTOR. Due to some of the long-lived isotopes in the reactor vessel and internals, ENTOMB, by itself, is not an allowable alternative.
The choice between SAFSTOR and DECON must be made based on a variety of factors including availability of fuel and waste disposal, land use, radiation exposure, waste volumes, economics, safety, and availability of experienced personnel.
Each alternative has advantages and disadvantages.
The best option for a specific plant has to be chosen based on an evaluation of the factors involved.
The overriding factor affecting the decommissioning decision for LACBWR is that the only feasible place to store the activated fuel is onsite. A federal repository is not expected to be available for about 20 years. With the fuel in the Fuel Element Storage Well, the only possible decommissioning option is SAFSTOR. Only limited decontamination and dismantling of unused systems can be performed during this period.
There are other reasons to choose the SAFSTOR alternative.
The majority of piping contamination is Co-60 (5.27 yr half-life) and Fe-55 (2.7 yr half-life).
If the plant is placed in SAFSTOR for 50 years, essentially all the Co-60 and Fe-55 will decay to stable elements.
Less waste volume, with less radioactivity cont mt, will be generated, and exposure to personnel perform-ing the decontamination and dismantling activitiet, will be significantly lower. Therefore, delayed dismantling supports the ALARA (As Low As Reason-ably Achievable) goal. The reduction in dismantling dose exceeds the dose the monitoring crew receives during the SAFSTOR period.
The decommissioning cost estimate is discussed in Section 6.7 of the Decommissioning Plan. The majority of studies show that while the total cost of SAFSTOR with delayed DECON is greater than immediate DECON, the present value is less for the SAFSTOR with delayed DECON option.
The nain disadvantage of delayed 9 ECON is that the plant continues to occupy the land during the SAFSTOR period. The land cannot be released for other purposes. DPC also operates a 350 MWe coal-fired power plant on the site.
Due to the presence of the coal-fired facility, DPC will continue to occupy and contrcl the site, regardless of the nuclear plant's status. Therefore, the continuet commitment of the land to LACBWR during the SAFSTOR period is not a significant disadvantage.
A second disadvantage of delaying the final decommissioning is that the people who operated the plant would not be available for the DECON period.
When immediate DECON is selected, some of the experienced plant staff would be available for the dismantling. Their knowledge of plant characteristics and events could be extremely helpful.
In the absence of these knowledgeable people, all information has to be obtained from plant records. When SAFSTOR is chosen, efforts must be made to maintain excellent records to compensate for the lack of staff continuity.
ER-SUPP 2
December 1987
w The remaining factor to be discussed is safety. As of August 1987, 43 power reactors have been shut down worldwide, 19 of which are in the United States.
All 3 methods of decommissioning are being used. Experience has shown that all can be used safely.
A possible fourth decommissioning alternative exists which combines some features of SAFSTod and DECON. The possibility exists to use the secondary side of the plant with a new fossil-fired steam supply system. While DPC is not planning on pursuing this option at this time, it should not be eliminated as an alternative.
After evaluating the factors involved in selecting a decommissioning alternative, Dairyland Power Cooperative decided to choose an approximately 30-50 year SAFSTOR period, followed by DECON.
The exact duration of the SAFSTOR period will be dependent on the availability of the federal fuel repository, availability of waste disposal, economics, personne'l exposure, and various institutional factors.
If any major changes are made in DPC's decommissioning plans, a revision to this plan will be prepared.
REFERENCES
- 1) Nuclear Regulatory Commission, proposed rule on Decommissioning Criteria for Nuclear Facilities, Federal Register, Vol. 50, No. 28, February 11,
- 1985,
- 2) "Decommissioning - Demonstrating the Solution to a Problem for the Next Century," Nuclear Engineering International, Vol. 32, No. 399, October 1987, p. 48.
- 3) Proceedings from the 1987 International Decommissioning Symposium, Conf-871018, October 4-8, 1987.
3.0 ONSITE STORAGE OF FUEL Eleven fuel cycles over the 20 years of operation have resulted in a total of 333 irradiated fuel assemblies being stored in the LACBWR Fuel Element Storage Well.
Fifty-two new fuel assemblies remain in New Fuel Storage. It is expected that the new fuel will be sent offsite during 1988.
DPC plans to continue to store the irradiated fuel assemblies in the Fuel Eicment Storage Well (FESW). The FESW is located in the Containment Building. The storage well is an 11-foot square pool, approximately 42 feet deep. The fuel is stored in a two-tier configuration in high density storage racks.
The Nuclear Regulatory Commission issued its Waste Confidence Decision in the Federal Register on August 31, 1984.
In it, the NRC found "reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the expiration of that reactor's operating license at that ER-SUPP 3
December 1987
reactor's spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations." Therefore, DPC's plan to maintain the activated fuel at LACBWR, until a federal repository is ready to accept the fuel, is acceptable from the safety standpoint, as well as necessary from the practical standpoint.
4.0 PLANT STATUS AND ACTIVITIES The LACBWR Decommissioning Plan provides a description of the plant and discusses the status of plant systems.
It also contains an estimation of the plant's radionuclide inventory.
The Decommissioning Plan describes the preparations for SAFSTOR and activities planned for the SAFSTOR period. The Decommissioning Program and its administrative controls are discussed. The Decommissioning Plan also describes the Radiation Protection Program, radioactive waste handling, and various monitoring programs.
These programs are quite similar to those used during plant operation.
5.0 CHANGE IN LAND USE During the SAFSTOR period, LACBWR will continue to occupy its present site.
The land committed to the plant will remain the same; only the function of the facility will change. The site will be used for safe storage of the decommissioned nuclear plant, instead of for an electric power generating facility.
Adjacent to LACBWR is a coal-fired power plant, Genoa Unit No. 3.
Due to the presence of this facility, DPC will continue to use the site for industrial purposes, regardless of the status of the nuclear plant.
5.1 Effect on Historical Sites DPC contacted the State Historical Society of Wisconsin to inquire whether the proposed decommissioning would have any effect on any property listed in the National Register of Historic Places or eligible to be listed.
The reply stated that the first stage of decommissioning (SAFSTOR) would not have any potential to affect such historic places.
The final decommissioning (DECON), however, could have effects on historical and archeological sites.
Therefore, the impact on historical sites will have to be evaluated during the planning for the DECON effort.
5.2 Manpower The number of DPC employees at the nuclear plant during the SAFSTOR period is expected to be approximately 30% of the operating staff. Traffic associated with LACBWR will be less than during plant operation.
ER-SUPP 4
December 1987
o 6.0 THERMAL EFFUIENTS The thermal release from the plant during SAFSTOR will be significantly less than that experienced during plant operation. No longer will the waste heat from the steam cycle be released to the Mississippi River. The main sources of heat remaining are the decay heat from the irradiated fuel, which will continue to decrease with time, and the heat generated by operating equipment, e.g. the air compressors.
WI 11e all open water systems listed in the Environmental Report will c6ntinue to be operated, the water use will be less. The Circulating Water System will be operated only intermittently, rather than continuously.
Less water will also be needed to compensate for system leakage. Thermal discharges will be accomplished primarily by the Low Pressure Service Water System through the same thermal effluent pathway to the Mississippi River as during plant operation.
The shutdown of the 50 MWe LACBWR will not affect the thermal discharges from Genoa Unit No. 3, the 330 MWe coal-fired generating unit onsite. A common discharge into the civer is used by both plants. The original Environmental Report discussed that it was unlikely that both plants would be shut down at the same time, and so the probability of subjecting the fish to cold-shock if LACBWR was shut down was small. With LACBWR permanently shut down, its thermal discharges cannot reduce the effect of a shutdown of Genoa Unit No. 3.
Due to the immediate mixing at the discharge point, however, thermal shock has not been seen. Therefore, the decommissioning of LACBWR will not l
cause a significant greater environmental effect when Genoa Unit No. 3 shuts I
down.
1 7.0 ENVIRONMENTAL EFFECTS DURING SAFSTOR This section describes the environmental impacts during the LACBWR SAFSTOR r
period. The principal potential sources of impact are gaseous and liquid releases of radioactive material, solid radioactive waste processing and shipments, chemical wastes, and noncombustible solid wastes. This section describes limits and methodology to calculate offsite doses.
Estimated doses to Members of the Public from releases of radioactive material during the SAFSTOR period are included in this section. A discussion of the comparative releases during SAFSTOR to releases during normal plant operation are also presented. The Radiological Environmental Monitoring (REM) Program description along with current and projected SAFSTOR results of the REM program are included in this section.
Effluent discharges from LACBWR will be significantly below applicable state and federal regulations during the SAFSTOR period.
Estimated doses to Members of the Public from liquid and gaseous releases of radioactive msterials will be significantly less during SAFSTOR than during normal plant operations, which were below the applicable 10 CFR 50 Appendix I dose limits.
ER-SUPP 3
December 1987
The release of chemicals and hazardous materials from LACBWR during the SAFSTOR period should be insignificant. As discussed in Section 6.0, heat released by the plant's circulating water system during the SAFSTOR period will not be of any significance.
I 7.1 Effluent Release (10 CFR 50 Appendix I) Limits:
The Effluent Release (10 CFR 50 Appendix I) Limits which are enumerated by LACBWR Technical Specifications will still apply during the SAFSTOR period.
Specifically, these required limits for effluent releases during the SAFSTOR period are:
7.1.1 Liquid Releases Requirement #1:
"The concentration of radioactive material released in liquid effluents at any time to areas beyond the EFFLUENT RELEASE BOUNDARY shall be limited to the concentrations specified in 10 CFR Part 20. Appendix B, Table II, Column 2,...for radionuclides other than dissolved or entrained noble gases (Kr-85). For dissolved or entrained noble gases (Kr-85), the concentration shall be limited to 2 x 10*" pCi/ml total activity."
Requirement #2:
"The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to areas beyond the EFFLUENT RELEASE BOUNDARY shall be limited:
During any calendar quarter to 1 1.5 mRems to the total body and a.
1 5 mRems to any organ, and b.
During any calendar year to 1 3 mRems to the total body and 5 10 mRems to any organ."
7.1.2 Gaseous Releases Requirement #1J "The dose rate due to radioactive materials released in gaseous effluents to areas beyond the EFFLUENT RELEASE BOUNDARY shall be limited to the following:
The dose rate limit for noble gasea shall be 1 500 mRems/yr to the total a.
body and f 3000 mRems/ year to the skin, and b.
The dose rate limit for H-3 and for all radioactive materials in particulate form with half lives greater than 8 daya shall be
$ 1300 mRems/ year to any organ."1 f
i ER-SUPP 6
December 1987
Requirement #2:
"The air dose to a MEMBER OF THE PUBLIC due to noble gases released in gaseous effluents to areas beyond EFFLUENT RELEASE BOUNDARY.shall be limited to the followings During any calendar quarter, to 5 5 mRads for gamma radiation and a.
i 10 mRads for beta particle radiation, and b.
During any calendar year, to f 10 mRads for gamma radiation and
$ 20 :iRads for beta particle radiation."
Requirement #3:
"The dose to a MEMBER OF THE PUBLIC from H-3 and all radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents released to areat beyond EFFLUENT RELEASE BOUNDARY shall be limited to the following During any calendar quarter to i 7.5 mRems to any organ, and a.
b.
Duriag any calendar year to $ 15 mRems to any organ."1 t
7.1.3 Solid Radioactive Waste Processing and Shipments Requirement #1:
"Solid radioactive wastes shall be handled in accordance with a PROCESS CONTROL PROGRAM in order to meet shipping and burial ground requirements."
7.1.4 Radiological Environmental Monitoring (REM) Program Requirement #1:
"The radiological environmental monitoring program shall be conducted as specified in Table 4.4.1-1.
An Interlaboratory Comparison Program for annual analyses of radioactive materials shall be conducted."2 A technical specification submittal to remove I-131 and I-133 from this 1
requirement is planned since these radionuclides have decayed to stable elements and are no longer being produced at LACBWR.
A technical specification submittal to remove reference to a land use 2
census from this requirement is planned, since activity concentrations in gaseous releases have diminished significantly and an annual vegetation sample will be taken.
ER-SUPP 7
December 1987
Cumulative doses to Members of the Public due to liquid, gaseous releases and direct radiation from the plant will be determined as required in accordance with methodology and parameters found in the LACBWR Offsite Dose Calculation Manual (0DCM). The ODCM is described in Section 7.5.
The REM Program supplements the effluent release monitoring and dose assessment program by verifying through sample collection and analysis that the measurable concentrations of radioactive materials and levels of radiation released from the plant effluents are not higher than those expected on the basis of effluent measurements.
7.2 Liquid Releases The purpose of the liquid waste processing and treatment system is to (1) collect and temporarily store radioactive liquid waste generated in the plant, and (2) to process and discharge liquid wastes in compliance with 10 CFR 50 Appendix I limits.
The total fission and activation products in liquid releases from the plant are summarized in Table 7-1 for the years 1974-1984, 1985 and 1986, the period January-March 1987, and the period July-September 1987. The plant was operational during 1985, part of 1986 and for the period January-March 1987. The reactor was shut down on April 30, 1987, and reactor defueling was completed in June 1987. The liquid releases of tritium (H-3) are summarized in Table 7-2.
Major radionuclide. breakdowns (isotopic analyses) for 1985 through 1987 are summarized in Table 7-3.
After liquid waste is collected in a tank, the contents are recirculated in the tank using the tank transfer / recirculating pump with the required recirculation piping line-up. After a recirculation period, a sample is taken from the tank and analyzed for radioactivity concentrations and effective MPC (Maximum Permissible Concentration in water ratio). A batch y
discharge record is documented, including necessary circulating water dilution flow rates and effective MPC upon entrance into the Mississippi y
River water. The tank's contents are discharged through one of two backwashable filters and pass full flow through a shielded hemispherical stainless steel marinelli chamber, housing a sensitive radiation detector which reads out in the control room. This radiation detection system has a conservatively set ala m set point, which if reached would activate a trip which would terminate the discharge by closing the automatic liquid waste discharge flow control valve. The total discharge volume is measured by a flow totalizer. The liquid effluent discharges into the circulating water system. The discharged waste water is further diluted by mixing with the Genoa-3 circulating water discharge and with the waters of Thief Slough. The waters of the slough mix with the mainstream of the Mississippi River and effect further dilution.
Using individual batch discharge radionuclide analyses and flow rates, offsite dose calculations to Members of the Public are periodically performed. No credit is taken for further dilution in Thief Slough or the mixing zone of the mainstream of the Mississippi River in performing these calculations.
l ER-SUPP 8
December 1987
s a
'l Theoretically, Members of the Public engaged in recreational activities, such as swimming or boating, could receive direct external exposure from the beta and gamma radiation emitted from the radionuclides in the water. However, the doses associated with these recreational endeavors should not exceed approximately 0.005 mrem whole body dose in any year, assuming a very conservative exposure time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> in the vicinity of the circulating water discharge and based upon the highest concentrations of radionuclides discharged in liquid waste. The dose from this pathway is not considered in calculations, since the dose is insignificant and boating and swimming do not normally take place directly at the outfall.
Theoretically, a Member of the Public could receive an internal dose from consumption of Mississippi River water which is processed for a potable water supply. However, the drinking water pathway is not considered in the offsite dose calculations, since the nearest community which obtains its drinking water supply from the Mississippi River is located in Davenport, Iowa, 195 miles downstream. -The dose from this source would be immeasurable.
The only significant dose pathway used for offsite dose calculations to Members of the Public for liquid effluents is the dose commitment due to the adult ingestion of fish from the Mississippi River.
These dose calculations for the adult-fish consumption pathway, very conservatively assume that the I
Adult Member of the Public consumes 21 kilograms / year of the edible portions of fish, which have continuously resided at the circulating water discharge outfall.' Results of dose calculations for this pathway are found in Table 7-4.
These dose pathways are illustrated in Figure 7.1.
Af ter LACBWR was permanently shut down in April 1987, marked changes in the magnitude and composition of the radioactive liquid effluents were observed.
i The total tritium (H-3) activity discharged in the 1st quarter of 1987 decreased from 25.2 Curies to 1.47 Curies in the 3rd quarter of 1987. The H-3 activity levels discharged during the SAFSTOR period will continue to decrease due to lack of production through fission and activation of douterium and due to radioactive decay.
Cobalt-60 and Iron-55 continue to represent about 75-80% of the discharged activities and contribute the most toward calculated whole body doses. Cesium-134 and Cesium-137 represent about 5-6% of the discharged activity. They are major contributors toward organ doses. These radionuclides will continue to be the predominant activity in the liquid releases during the SAFSTOR period. Although the total activity released per calendar quarter may initially change due to variations in tasks performed at the plant, the overall trend in release activities and offsite doses from liquid discharges will decrease l
significantly during the SAFSTOR period. This will be due to reduced plant radioactivity concentrations because of lack of production and because of radioactive decay. Shorter-lived radionuclides, such as I-131, noble gases.
l Ba(La)-140 and Tc-99m, which contribute to doses during operation, were nondetectable in liquid releases during the 3rd quarter 1987. Most of these radionuclidos have completely decayed to stable elements and will no longer be produced during SAFSTOR.
Other radionuclides, such as Mn-54, Cr-51 Co-58 l
and Ru(Rh)-106, will be nondetectable in liquid discharges within a year or i
two, because of decay.
)
ER-SUPP 9
December 1987
5 The activity concentrations in batches which are currently being discharged range from about 2 x 10 5 to about 6 x 10'3 pCi/ml with an average concentra-i tion of about.8 x 10 " pCi/ml. The average batch concentrations will con-tinue to vary initially but will decrease in the longer term SAFSTOR period.
The total volume of liquid discharges during the period January September 1987 was 276.842 gallons. The total volume discharged from July-September 1987 was 70,253 gallons. The estimated total volume'to be discharged in 1987 is 350,000 gallons..The initial SAFSTOR volumes of liquid discharges will vary, depending upon plant tasks, but should be an average of about 150,000 gallons per. year.
Eventually, the volumes discharged shoald decrease.during the long-term SAFSTOR period.
Normal and potential sources of liquid wastes processed during SAFSTOR include the following:
4
- 1) Turbine Building floor drains and sumps - laundry water, laboratory water, decontamination liquids.
- 2) Waste Treatment Building sump - decontamination liquids, resin dewatering liquids.
i
- 3) Containment Building sumps and Retention Tanks - FESW leakage, air conditioner condensate. decontamination liquids and system draining evolutions.
i 2
4 i
4 i
1 3
]
a ER-SUPP 10 December 1987
., - ~
. ~.,...
TABLE 7-1 TOTAL FISSION AND ACTIVATION PRODUCT ACTIVITY (EXCLUDING H-3)
RELEASED IN LIQUID EFFLUENTS (Curies) l l
l l
l l
1974-1984 l
1985 l
1986 l
Jan-Mar 1987 l
July-Sep 1987 l
l (avg /yr) l 1
~l l
l l
l l
l l
l l
l l
l l
l 7.6 1 6.1 l
1.84 l
5.00 l
0.157 l
0.320 l
I I
I I
I l
i i
e i
i i
l I
i i
i g
TABLE 7-2 TOTAL H-3 ACTIVITY RELEASED IN LIQUID 2FFLUENTS (Curies)
I e
4 4
1 a
e g
g g
l 1985 l
1986 l
Jan-Mar 1987 l
July-Sep 1987 l
l
\\
\\
\\
l l
l l
l l
l l
l l
l l
128.0 l
57.5 l
25.2 l
1.47 l
l l
l l
l l
l l
l ER-SUPP 11 December 1987
h TABLE 7-3 MAJOR RADIONUCLIDE BREAKDOWN
)
IN LIQUID RELEASES (C1) 4 I
I l
I l
l Jan-Mar 1987 ! July-Sep 1987 'l j Radionuclide 1985 l
1986 l
l l
l l
l l
Co-60 l
0.908 l
1.892 l
0.067 i.
0.216
.)
l Fe-55 l
0.298 l
1.381 l
0.053 l
N.A.
l l
Mn-54 l
0.115 l
0.516 l
0.014 l
0.045 l
l Cs-137 l
0.083-l 0.094 l
0.006 l
0.019 l
l I-131 l
0.069 l
0.031 l
<0.001 l
N.D.
l j
I-113 j
0.059 l
0.072 l
<0.001
-l N.D.
l l
.Ba(La)-140 l.
0.056 l
.0.019 l
0.002 l
N.D.
l l
Tc-99m l
0.044 l
0.020 l
0.001 l
N.D.
l
[
l Xe-135 l
0.029 l
0.009 l
<0.001 i
N.D.
l l
l Np-239 l
0.027 l
0.008 l
<0.001 i
N.D.
l i
l I-135 l
0.021 j
0.007 l
N.D.
l N.D.
l
)
1 l
Co-58 l
0.020 l
0.020 l
<0.001 l
0.002 l
i l
Sr-91 l
0.015 l
0.004 l
<0.001 l
N.D.
j l
Xe-133 l
0.015 l
0.017 l
<0.001 l
N.D.
l j
Mo-99 l
0.013 l
0.034 l
N.D.
N.D.
l l
l Cr-51 l
0.011 l
0.095 l
<0.001 l
<0.001 l
l Xe-131m l
0.011 l
0.050 l
N.D.
l N.D.
l l
Sr-90 l
0.004 l
0.005 l
<0.001 l
N.A.
l l
Ru(Rh)-105 l
0.004 j
0.044 l
<0.001 l
<0.001 l
l Ru(Rh)-106 O.004 l
0.009 l
0.001 l
0.002 l
l Cs-134 0.003 l
<0.001 l
<0.001 l
<0.001 l
j Sr-89
'l 0.002 l
0.005 l
N.D.
l N.A.
l l
Ag-110m j
0.002 l
0.017 l
0.008 l
0.014 l
l l
l N.D. = Not detectable above LLD.
N.A. = Not available at the time of the report.
t L
i ER-SUPP 12 December 1987 t
_a
i t
f TABLE 7-4 CALCULATED DOSES TO MEMBERS OF THE PUBLIC FROM LIQUID EFFLUENT RELEASES I'
(mrem) 2 l
l l
I l
I i
1985
- j. 1986 l
Jan-Mar 1987 l
July-Sep 1987 l
l l
i l
l I
l l
\\
l
! Whole Body l
0.803 l
2.033 l
0.093 j
0.611 l
i i
i I
i l
l
\\
l l
l l
l l
I l Highest Organ l 1.211 3.136 l
0.139 l
0.829 l
i I
I l
1 r
7.3 Gaseous Releases The purpose of the gaseous waste holdup, filtration and treatment systems was to (1) reduce radioactivity discharged to the environment through the gaseous release systems. (2) allow for sufficient decay of shorter-lived noble gases, activation gases, and radioiodines prior to discharge to the environment, and (3) control radioactive gaseous emissions within the requirements of 10 CFR i
[
Appendix I limits.
The total fission and activation product activity released in gaseous effluents is shown in Table 7-5 for the years 1985 and 1986, the period
[
January-March 1987, and the period July-September 1987. As previously discussed, the plant was operational during 1985, part of 1986, and for the period January-March 1987. The period July-September 1987 reflects the current conditions of LACBWR as a defuelled-nonoperational facility. The major radionuclide breakdowns (isotopic analyses) for the periods of time described are listed in Table 7-6.
i j
During normal plant operation, the majority of the radionuclides discharged via gaseous releases to the environment were from offgas (non-condensible gases) from the condenser hotwell, which was maintained at a vacuum. The maj'rity of these radioactive gases were noble gas radionuclides. These gas, passed through offgas piping from the air ejector. To reduce their environmental impact, an augmented piping and tank system was utilized to hold up these gases to allow for sufficient radioactive decay of a number of radionuclides, which reduced the total emission rate from the plant stack.
The gaseous release ftom the offgas system remained well within applicable federal and state limits during normal operation, f
i t
ER-SUPP 13 Decamber 1987
+
l
The remaining sources of radioactive gaseous releases are from various ver.tilation exhaust systems. These include the Containment Building, Turbine Building and the Waste Treatment Building Exhaust Ventilation Systems which, essentially, join at the stack plenum prior to discharge up the stack. One or two stack blowers, rated at about 35,000 cfm each, act as a driving force for stack releases and add dilution air to reduce the effective MPCA within the stack prior to discharge.
The offgas system from the condenser is no longer in operation since the plant is shut down. Therefore, the majority of gaseous releases are from the various ventilation exhaust systems. The Containment Building and Waste Treatment Building Exhaust Systems are equipped with full flow HEPA filter banks. The stack currently is continuously monitored for beta particulate, radiciodine and noble gas activities, with at least one stack monitor, which l
isokinetically samples the stack exhaust flow.
Filter papers are periodi-cally replaced in the monitor and isotopically analyzed for gamma emitting radionuclides to determine individual radionuclide total releases.
Filter composites are periodically analyzed for pure beta emitting radionuclides (e.g. Sr-89/90).
Theoretically, Members of the Public could receive radiation doses from radioactivity released in gaseous effluents in three principal pathways, inhalation, ingestion and immersion.
During normal plant operations, the major dose pathway is the immersion dose (external exposure) pathway. This was principally due to noble gas activity constituting over 99% of the radionuclides being released in gaseous effluents.
The second major dose pathway was from inhalation of small quantities of radionuclides released in gaseous effluents.
The radionuclides which constituted the main fraction of dose via this pathway were the radioiodines (I-131. I-133) and Ba(La)-140. The last major dose pathway was from the infant ingestion of milk containing minute quantities of radionuclides, principally I-131 and Ba(La)-140, produced during plant operation. These pathways are illustrated in Figure 7.2.
After LACBWR was permanently shut down in April 1987, marked changes in the magnitude and composition of the radioactive gaseous effluents were observed.
This is shown in Table 7-5 and Table 7-6.
There are essentially no noble gas (Kr. Xe, Ar) nor radioiodine (I-131, 133) releases, since they are no longer being produced by the fission process, and any residual activity for these radiontalides (except Kr-85) within the stored spent fuel or plant systems has completely decayed to stable elements.
In addition, many particulate-type radionuclide releases have diminished significantly due to radioactive decay and non production.
The Ba(La)-140 and Ru-103 have essentially completely decayed to stable elements.
The particulate releases in gaseous A technical specification submittal to remove the iodine monitoring A
requirement is planned since the iodines are no longer being produced and have decayed to stable elements.
ER-SUPP 14 December 1987
..s effluents have decreased by a factor of about 60-100 times since shutdown.
The tritium releases in gaseous effluents have significantly decreased.
During the SAFSTOR period, the noble gas, radioiodine and tritium releases in gaseous effluents will not be evident under normal cenditions. Although the i
particulate release activity may initially vary due to changes in plant tasks f
performed, it will continue to be lower than during plant operation, and the i
overall trend in releases activities and doses due to inhalation will continue to decrease during the long-term SAFSTOR period. The most important dose pathway for gaseous releases during SAFSTOR will be the inhalation pathway, although doses to Members of the Public from gaseous release would remain bnlow 0.001 mrem por year, as illustrated in Table 7-7.
TABLE 7-5 i
TOTAL FISSION AND ACTIVATION PRODUCT ACTIVITY RELEASED IN GASEOUS EFFLUi!NTS (Curies) i i
i i
e i
i i
l Radionuclide l
l l
l l
l j
Category 1985 l
1986
! Jan-Mar 1987 ! July-Sep 1987 l l
l l
l l
l l
l Noble Gases l
8580 l
3530 l
1793 l
N.D.
l i
i l
l l
l l
l l
l I-131 and I-133 l 1.1x10'8 j 6.76x10 8 l
1.24x10 8 N.D.
e l
l l
l l
l l
l l
l H-3 l
34.8 l
12.1 l
8.71 i
N.D.
l l
l l
I l
l l
l l
l l
l l
Particulates l
l l
l l
l (T1/2 >8 days) l 2.63x10-8 l 5.47x10-*
l 2.97n10
l 5.13x10 5 i
i l
including alpha l l
l l
l l
(includes natural!
l l
l l
i l
radionuclides) l l
l l
l
[
l
[
N.D. = Not detectable above LLD.
i i
I i
e i
l i
l i
ER-SUPP 15 December 1987
TABLE 7-6 MAJOR RADIONUCLIDE BREAKDOWN IN GASEOUS RELEASES (Curies) l I
I
\\
l l
l Radionuclide 1
1985 1986 Jan-Mar 1987 !
July-Sep 1987 l
l l
l I
I l
l Xe-135 l
5320 l
2110 l
988 l
N.D.
l l
Xe-133 l
1120 l
428 l
252 l
N.D.
l l
Kr-88 l
926 l
347 l
166 l
N.D.
l l
Kr-85m l
395 l
154 l
83.8 l
N.D.
l l
Xe-131m l
383 l
71.8 l
61.1 l
N.D.
l 4
l Kr-87 l
373 l
156 l
83.6 l
N.D.
l l
Xe-135m i
152 l
50.6 l
38.7 l
N.D.
l l
Xe-138 l
133 l
129 l
109 l
N.D.
l l
Ar-41 l
60.4 l
44.2 j
6.1 i
N.D.
l l
Xe-137 l
21.3 l
22.3 l
4.2 l
N.D.
l Xe-133m j
1.4E-1 l
13.4 l
1.8E-3 l
N.D.
l l
I-133 l
6.3E-3 l
2.1E-3 l
5.1E-4 l
N.D.
l l
I-131 l
4.9E-3 l
3.5E-3 l
7.2E-4 l
N.D.
l l
Ba(La)-140 l
2.0E-3 l
3.0E-4 l
1.3E-4 l
N.D.
l l
Co-60 l
2.3E-4 l
9.5E-5 l
1.3E-5 l
N.D.
l l
Mn-54 l
8.2E-5 l
4.2E-5 l
6.1E-6 l
5.5E-7 l
l Cs-137 l
5.8E-5 l
3.5E-5 l
4.3E-6 j
4.1E-7 l
l Sr-89 l
4.7E-5 l
1.6E-5 l
1.3E-4 l
N.A.
l Ce-144 l
3.2E-5 l
1.5E-6 l
5.3E-6 i
N.D.
l l
Co-58 l
3.2E-5 l
1.2E-5 l
1.8E-6 l
N.D.
l l
Cr-51 l
1.1E-5 l
6.1E-6 l
4.6E-6 l
N.D.
l l
Ru(Rh)-106 l
4.3E-5 l
1.1E-5 l
N.D.
I N.D.
l j
Ce-141 l
7.8E-6 l
2.3E-6 l
1.5E-6 l
N.D.
l l
Ru-103 l
3.4E-6 l
4.2E-6 l
5.7E-7 l
N.D.
l Sr-90 l
7.5E-7 l
N.D.
l 1.6E-7 l
N.A.
l l
Alpha l
1.2E-4 l
2.1E-5 l
5.1E-6 l
4.1E-6 l
l (including l
l l
l l
l naturals) l l
l l
l l
34.8 l
12.1 l
8.7 l
N.D.
l l
l l
N.D. = None detectable above LLD.
N.A. = Not available.
ER-SUPP 16 December 1987
[
TABLE 7-7 CALCULATED DOSES TO MEMBERS OF THE PUBLIC FROM GASEOUS EFFLUENT RELEASES (mrem) i c
l l
l l
t i
1985 l
1986 l Jan-Mar 1987 July-Sep 1987 l
r l
l l
l i.
l l
l l
l l
Whole Body l
l l _
0.166 l
0.000 l
4 l (Noble Gass-) -l 1.232 l
0.662 l
l l
l' l
l l
l l
l Skin' l
l l
l l
l (Noble Gases) l 1.176 l
0.632 l
0.158 l
0.000 l
f l
l l
l l
l-l l
l Highest Organ l l
l l
l (I-131, I-133 l
0.563 l
0.470 l
0.083 l
0.001 l
1 j
l and/or H-3 &
i (Child l (Child l
(Child l
(Child j
l Particulates) l Thyroid) l Thyroid)l Thyroid) l Bone) l
)
I l
l e
I
..i' 7.4 Solid Radioactive Waste processing and Shipments
?
During the SAFSTOR period, the principal types of radioactive solid waste which will be processed and shipped to a suitable disposal facility will be low level radioactive wastes principally with radioactivity content less than t
Class C (10 CFR 61) wastes. This will include
- 1) Dry Active Wastes (DAW), normally Class A unstable.
[
r
- 2) Dewatered Spent Demineralizer resins and filtration media, normally Class A, B or C stable, and
- 3) Contaminated or irradiated plant system components, normally Class B or C, stable.
r All solid radioactive wastes will be processed and shipped in accordance to i
an approved Process Control Program.
The average volumes of low level solid radioactive waste shipped during the period 1977-1986 and 1987, when LACBWR was operating, are listed on Table 7-8.
Projected low level solid waste L
shipments for the SAFSTOR period are listed on Table 7-9.
Although the waste volumes and activities should initially vary, they will be below historical l
operating average annual volumes and activities. This will be principally I
because of reduced DAW generation, reduced demineralizer resin and filtration i
component usage, and because of radioactive decay of Co-60 Fe-55 Hn-54, Ce-141/144, and other radionuclides.
During the longer term SAFSTOR period, f
the average waste volumes generated may be approximately 7.5 m /yr, which 8
I ER-SUPP 17 December 1987 i
~
would be a factor of about 3.5 less than the historical operating annual The average annual activity in waste should be a factor of up to 6 average.
times less than in waste shipped during periods of operation.
It is estimated that the average annual dose to any Member of the Public, due to direct radiation exposure from waste processing and shipment activities during the SAFSTOR period, will not exceed 1 mrem / year.
TABLE 7-8 LOW LEVEL RADIOACTIVE WASTE SHIPMENTS FROM LACBWR DURING OPERATION l
l l
l 1977-1986 1987 l
I l
l l
l Volume /yr l
l l
l (m )
j 25.22 21.53 l
29.31 l
3 l
l l
l l
Activity /yr l
l l
l (Ci) l 158.78 i 162.33 l
95.0 l
l l
TABLE 7-9 PROJECTED LOW LEVEL RADIOACTIVE WASTE SHIPMENTS FROM LACBWR DURING SAFSTOR I
I l
l I
l l
1988 l
1989 1990 1
1990-2010 l
l I
I l
l I
l Volume /yr l
l l
l l
l (m )
l 10 l
21 l
5 l
5 - 7.5(1) l a
l l
l l
l Activity /yr l
l l
l l
l (Ci) l 140 l
65 l
40 l
15 - 20 l
l (1) Assumes periodic shipments of contaminated and/or irradiated components no longer required for SAFSTOR maintenance and surveillance.
ER-SUPP 18 December 1987
s 7.5 offsite Dose calculation Manual (ODcM)
The purpose of the Offsite Dose Calculation Manual (ODCM) is to describe the methodology and parameters to be used in the calculation of instantaneous release rate monitor alarm setpoints and the dose commitments to Members of the Public beyond the Effluent Release Boundary due to the release of radioactive materials in liquid and gaseous effluents from LACBWR. The ODCM contains calculational methods for determining periodically the effluent radiation monitors' alarm setpoints, and also provides site specific dose factors for the LACBWR site. The site specific dose factors enable the calculation of curaulative dose contributions to Members of the Public on a quarterly and annual basis for principal pathways of exposure, as required by Technical Specifications, to demonstrate compliance with the requirements listed in Section 7.1.
The ODCM also contains a description of the Radiological Environmental Monitoring (REM) Program.
7.6 Estimated Doses to Members of the Public During SAFSTOR The estimated average dose to any Member of the Public for the initial few years of the SAFSTOR period is expected to be less than 2 mrem / year.
For the long term SAFSTOR period, the estimated average dose is expected to decrease well below 1 mrem / year, due to nonproduction and continued radioactive decay of predominant residual radionuclides (e.g., Co-60, Fe-55, Ag-110m, Mn-54, Cs-137). The initial estimated average annual dose is summarized as follows:
Direct External Radiation Dose (Primarily from LLRW shipments) 1.000 mrem /yr Liquid Releases (Fish - Adult Ingestion Pathway) 0.500 mrem /yr Liquid Releases (Immersion or External Dose) 0.005 mrem /yr Liquid Releases (Ingestion) 0.001 mrem /yr Gaseous Releases (Inhalation) 0.001 mrem /yr Gaseous Releases (Ingestion) 0.000 mrem /yr Gaseous Releases (Immersion or External Dose) 0.000 mrem /yr TOTAL 1.507 mrem /yr ER-SUPP 19 December 1987
7.7 Nonradioactive Wastes 7.7.1 Chemical Wastes Chemical wastes which could be produced during the SAFSTOR period will be processed and handled according to appropriate hazards classification, and discharged to the environment in accordance with the requirements of the Wisconsin Pollution Discharge Elimination System (WPDES) Pernit. The discharge to the Mississippi River comes from one source--the plant circulating cooling water. No chlorine or other chemicals are used to treat cooling water, and therefore none are discharged. Decontamination chemicals, laboratory chemicals (reagents), and minute quantities of residual Hg Cr0
- 4 K Cr2 7 and Na, Cr04 which have been used in selected plant systems for 2
0 corrosion inhi5ition are periodically discharged with the liquid radioactive waste batch discharges. Morpholine, whit:h is added to the heating boiler during boiler usage, may evaporate to the atmosphere. SulfurJe Acid and Sodium Hydroxide used for makeup demineralizer regenerations are not discharged to the river, but are neutralized in a settling pond. Waste solvents, paints and degreasers will be disposed of in hazardous material collection containers. Any chemicals used for system decontamination during SAFSTOR will be processed using appropriate demineralizers or filtration systems which will be processed as solid wastes.
No sanitary wastes are discharged to the river. All of the waste from toilets and sinks (except for decontamination sinks) are discharged to underground septic tanks located within the Owner Controlled Area.
Plant shower discharges (except for decontamination showers) are discharged into the septic system.
Decontamination sinks and showers discharge to the liquid radioactive waste system.
7.7.2 Noncombustible Solid Wastes The disposal of noncombustible solid wastes has changed from that described in the original Environmental Report.
Nonradioactive noncombustible solid wastes are now deposited at a landfill in Genoa, instead of at the Genoa Unit No. 3 ash disposal area, with the exception of metal wastes, which are collected by a scrap metal dealer. Any asbestos will be disposed of in accordance with the Dairyland Power Cooperative asbestos control program.
7.8 Comparison of Expected SAFSTOR Releases to Releases During Plant Operation The release of radioactive material to the environment during the SAFSTOR period will be significantly less than during plant operation. There will be no noble gas nor radioiodine releases in either liquid or gaseous effluents during SAFSTOR. The shorter-lived particulate-type radionuclides such as Ba(La)-140, Zr(Nb)-95, Sr-89, Ru-103, Ce-141, Fe-59 and Cr-51 will not be available for release to the environment about 6 months to 1 year after shutdown. The intermediate-lived, particulate-type radionuclide sources such as Ag-110m, Ce-144, Mn-54. Ru-106 and Cs-134 will decrease by factors ranging from 29 to several thousand within ten years after shutdown. Cobalt-60 and Iron-55 which are predominant radionuclides in liquid waste discharges and solid wastes will decrease by a factor of 3.7 and 13 respectively, within ten years after shutdown, due to radioactive decay.
ER-SUPP 20 December 1987
l l
AO U ATIC R E LE ASES V
L SUBMERSION FISH W ATE R 0.005 mrem
<0.001 mrem 0.500 mrem U
MAN i
(
l l
Potential Radiation Exposure Pathways i
and Average Annual Doses to Members of the Public for Liquid Releases f rom LACBWR During SAFSTOR FIGURE 7.1 ER-SUPP 21 December 1987 l
1
ATMOSPHERE V
INH AL ATION PASTURAGE IMMERSION V
DAIRY ANIMALS V
FRESH FLUID MI L K
<0.001 mrem 0.000 mrem 0.000 mrem V
MAN
}
Potential Radiation Exposure Pathways and Average Annual Doses to Members of the Public for Gaseous Releases from LACBWR During SAFSTOR FIGURE 7.2 i
ER-SUPP 22 December 1987
O 8.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 8.1 A Radiological Environmental Monitoring (REM) Program will be conducted to comply with Technical Specifications and 10 CFR 50 Appendix I during the SAFSTOR period.
Environmental samples will be taken within the surrounding areas of the plant.
These samples will be analyzed to determine any effects plant effluent releases may have on the environment and the public. These samples will include (1) direct radiation measurement devices, (2) air particulate samples, (3) river water and sediment samples, (4) fish samples, (5) vegetation or foliage samples, taken on a periodic basis as specified by Technical Specifications.
8.2 The results of the REM program sample analyses for the period July-September 1987 are listed on Table 8-1.
Most of the results demonstrate that no plant attributable radioactivity nor radiation is detectable above the normal background radiation levels. Direct radiation levels in the environ-ment, as measured by TLD's, ranged from 14.8 3.0 mrem to 26.0 1 9.5 mrem / quarter. The average direct radiation measurement was 19.8 i 3.0 mrem / quarter. The beta particulate activities in environmental air samples remained at approximately 0.02 pCi/m on the average, with no significant 8
difference between the La Crosse air sample activities, nor samples obtained cicser to the site. No airborne I-131 activity above the MDA was detected t,
any charcoal cartridge samples. No detectable gamma emitting radionuclide activity was found in air particulate composite samples. No detectable gamma emitting radionuclide activity was found in fish samples nor milk samples.
No significant activity was detected in river water samples.
Mississippi River bottom sediment samples exhibited a small amount of plant attributed radionuclide accumulation, principally at the outfall of the circulating water system. The radionuclide adsorption into sediments is very localized and does not exhibit an increase over previous years' data. This will continue to be monitored and is expected to decrease during the SAFSTOR period.
These results are similar to results obtained in environmental sample analyses during plant operation.
These results should be typical o# results of environmental sample results for the entire SAFSTOR period.
In other words, no detectable plant attributed radioactivity should be present in environmental samples obtained during SAFSTOR.
ER-SUPP 23 December 1987
a TABLE 8-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM RESULTS
SUMMARY
(July-September 1987) l l
Average Sample Type Range of Sample Results i
Sample Result l
l Direct Radiation l
14.8 1 3.0 l
(TLD's) l to 26.0 i 9.5 mrem j
19.8 i 3.0 mrem l
l Air Particulate l
l Filter l
0.009 t 0.002 l
8 8
(Gross Beta) l to 0.050 t 0.004 pCi/m l
0.023 pCi/m l
l Air Particulate l
l Filter l
l 8
8 (Gamma isotopic) l
<3.66 E-4 to 4.43 E-3 pCi/m l
<7.00 E-4 pCi/m I
I Air Charcoal l
l 8
8 Cartridge (I-131) l
<1.86 E-3 to <4.11 E-3 pCi/m l
<3.00 E-3 pCi/m l
l Milk Sample l
l (Gamma Isotopic) l
<2.25 to <25.4 pCi/l l
<5.5 pCi/l I
I River Water l
l (Gamma Isotopic) l
<3.65 to <24.14 pCi/1 l
<10 pCi/l l
Fish - Edible l
l portion l
l (Gamma isotopic) l
<6.56 to <24.81 pCi/kg l
<12.5 pCi/kg i
River Sediments -
l l
Outfall +
l l
River Proper j
l 2.32 E3 i (Gamma isotopic) l 5.90 El to 6.81 E3 pCi/kg l
3.89 E3 pCi/kg ER-SUPP 24 December 1987
9.0 POTENTIAL ACCIDENTS DURING SAFSTOR 9.1 Introduction The probability of an accident occurring during the SAFSTOR period is considerably less that. during plant operation. The focus of the potential accidents has also changed. During operation, the focus was on minimizing the plant transient and cooling the reactor core.
During SAFSTOR, the only major concern is protecting the fuel in the Fuel Element Storage Well.
The fuel in the well, while not benign, is not as much a hazard as the fuel in the operating reactor was.
Since April 30, 1987, the fission product inventory has decreased and the decay heat generation is significantly less.
These factors reduce the consequences of any accident affecting the fuel. As time passes, the consequences will continue to decrease.
The reactor's design basis accidents were reviewed to determine which could still occur during SAFSTOR.
Some other accident scenarios which were not previously considered design basis accidents were also evaluated. A list of 8 postulated accidents was identified. These events are:
Spent Fuel Handling Accident Shipping Cask or Heavy Load Drop into FESW Loss of FESW Cooli.ng FESW System Pipe Break Uncontrolled Liquid Waste Discharge Loss of Offsite Power Earthquakes Wind and Tornado Each of these postulated events was evaluated based on the revised plant status to identify their potential consequences during the SAFSTOR period.
The following sections discuss these accidents.
One additional event was examined - a fire. The potential safety consequences of any fire fall within the scope of other evaluated events.
9.2 Spent Fuel Handling Accident This accident postulates a fuel assembly falling from the hoist into the Fuel Element Storage Well. The probability of this accident is extremely small, since minimal fuel handling will be performed during the SAFSTOR period until the fuel assemblies are rsmoved from the FESW.
Periodic inspections may be conducted during the years the fuel remains onsite.
In the almost 20 years of operation and associated fuel handling at LACBWR, no fuel assemblies were ever dropped.
In this event, it is assumed that the cladding of all the pins in 2 fuel assemblies ruptures. The fuel handling crew evacuates when the local area radiation monitor alarms.
Containment Building ventilation would isolate on high activity, but for this analysis, no containment integrity is assumed.
l ER-SUPP 25 December 1987
L The assumptions used in evaluating this event during SAFSTOR were similar to those used in the FESW reracking analyses.1 The fuel inventory calculated 2
for October 1987 was used. The only significant gaseous fission product available for release is Kr-85. The plenum or gap Kr-85 represents about 15%
(215.7 Curies) of the total Kr-85 in the fuel assemblies. However, for conservatism and commensurate with Reference 1, 30% of the total Kr-85 activity, or 431.4 Curiea, is assumed to be released in this accident scenario.
No credit was taken for decontamination in the FESW water or for containment integrity, so all the activity was assumed to be released into the environment. Meteorologically stable conditions at the Exclusion Area
~
-Boundarv (1109 ft, 338m) were assumed, with a release duration of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> commensurate with 10 CFR 100 and Regulatory Guides 1.24 and 1.25.
A stack release would be the most probable, but a ground release is not impossible given certain conditions. Therefore, offsite doses were calculated for 3 cases. The first is at the worst receptor location for an elevated release, which is 500m E of the Containment Building.
The next case is the dose due to a ground level release at the Exclusion Area Boundary.
The maximum offsite dose at the new proposed EmerFency Planning Zone boundary for a ground level release is also calculated. Adverse meteorology 3
8 1
is assumed for all cases.
Elevated Release Average Kr-85 Release Rate i
6.00 E-2 Ci/see 431.4 Curies
=
2 hrs. x 3600 sec/hr L
X 8
Worst Case 6 for 0-2 hours at 500m E = 2.3 E-4 sec/m Kr-85 average concentration at 500m E 3
1.38 E-5 Ci/m8 6.00 E-2 Ci/sec x 2.3 E-4 sec/m
=
I Immersion Dose Conversion at 500m E Kr-85 Gamma Whole Body Dose Factor (Regulatory Guide 1.109) i i
i l
1.61 E+1 mrem /yr x 10' pCi x 1.142 E-4 yr = 1,839 mrem /hr 1
pCi/m' Ci hr Ci/m*
Whole Body Dose at 500m E j
8 x 2 hr = 0.05 mrem 1839 mrem /hr x 1.38 E-5 Ci/m l
Ci/m 3
i l
T ER-SUPP 26 December 1987
[
\\
>Kr-85 Beta /Gamme Skin Dose Factor (Regalatory Guide 1.109) 1.24 E+3 mrem /*;r x 10' uCi x :.142 E-4 yr = 1.53 E5 mrem /hr uC1/m Ci-hr Ci/.,3 3
i Skin Dose at 500m'E 1.53 E5 adem/hr x 1.38 2.-5 Ci/m' x 2 hr = 4.2 mrem
~51/m 3
Ground Level Release at EAB 3
Worst Case Q for 2 hrs at 338m NE or 339ai SSE, using Regulatory Guide 1.25 2.2 E-3 see 3
m Whole Body Dose at 338m 0.49 mrem Skin Dose at 338m 40.4 mrem Ground Level Release at Proposed Emergency Planning Zone Boundary E
Worst Case Q ior 2 hrs at 100m E 1.02 E-2 see 3
m W5-le Body Dose at 100m E 2.25 mrem Skin Dose at 100m E 187 mrem As can be seen, the estimated maximum whole body dose is more than a factor of 11,000 below the 10 CFR 100 dose limit of 25 Rem (25,000 mrem) to the whole body within a 2-hour period.
'ER-SUPP 27 December 1987
~
9.3 Shipping' Cask or Heavy Load Drop into FESW This accident postul'ates a shipping cask or other heavy. load' falling into the Fuel Element Storage We.11.
Reference 1 stated that extensive local rack' deformation and fuel damage would occur during a cask drop accident, but with an additional plate (that was installed during the'reracking) in place, a dropped cask wou?d not damage the pool liner or floor sufficiently to adversely-affect'the leak-tight integrity of the storage well (i.e., would not cause excessive water leakage from the FESW).
For th'is accident, it is postulated that all 333 spent fuel assemblies located in the FESW are damaged. 'The cladding of all the fuel pins ruptures.
The same assumptions used in the Spent' Fuel Handling Accident (Section 9.2) are used here. A total of 35,760 Curies of Kr-85 is released within the 2-hour period. The doses calculated are as follows:
Elevated Release Whole Body Dose at 500m E 4.2 mrem Skin Dose at 500m E 350 mrem Ground Level Release at EAB Whole Body Dose at 338m 40.2 mrem Skin Dose at 338m 3,34 Rem Ground Level Release at Proposed Emergency Planning Zone Boundary Whole Body Dose at 100m E l
186 mrem Skin Dose at 100m E 15.6 Rem l
As can be seen, the estimated offsite doses for the cask drop accident are below the 10 CFR 100 limits. The postulated maximum whole body dose is more than a factor of 100 below the 10 CFR 100 limit of 25 Rem (25,000 mrem).
ER-SUPP 28 December 1987
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9.4 Loss of FESW Cooling This accident postulates a loss of FESW cooling. The most likely causes of a loss of cooling are:
- 1) Both FESW pumps fail or FESW piping has to be isolated for maintenance;
- 2) The Component Cooling. Water (CCW) System is out of service due to failure of both pumps or other reason. The CCW System removes heat from the FESW cooler.
- 3) The Low Pressure Service Water (LPSW) System is out of service due to failure of both pumps or other reason.
The LPSW System removes heat from the CCW coolers.
If the third possibility is the cause, cooling to the CCW coolers can be restored by cross-connecting the High Pressure Service Water System to the coolers, in lieu of LPSW.
A calculation was perforaed to determine FESW heatup rate if active cooling were lost. The decay heat generation rate for January 1, 1Po8, was used.
Pool boiling would commence approximately 5 days after the loss of cooling.
The tops of the control rods stored in the fuel racks would become uncovered after 13.6 days.
After 23.1 days, the top of the fuel would be exposed. If water is added to the FESW at sny time during this period, these consequences would be delayed.
Substantial time is therefore available for restoration of FESW cooling. No immediate action is necessary during this postulated accident.
9.5 FESW Pipe Break This accident postulates a break in the FESW system piping, other than in the pump discharge piping between the redundant check valves and the pool liner.
A load analysis was performed on this approximately 20 feec of piping.
It was concluded that all stresses are within ASME Code allowable.
(Reference 1 calls this line the spent fuel pool drain line.) The series check valves were added during the 1980 FESW reracking.
If the postulated break occurs, the lowest the FESW could drain is approximately 679'.
At this level all spent fuel will remain covered. The control rods which are currently stored in the fuel racks will be partially uncovered. The tops of the control rods are about elevation 686'.
The operator would be alerted to this accident by receipt of the FESW Level Lo/High alarm.
Any makeup water added may run out the break, depending on the size of the break.
A calculation has been performed to determine the radiation levels due to the exposed control rods.
In the vicinity of most of the FESW piping and isolation valves, the radiation dose would not be substantially increased due to the loss of water.
ER-SUPP 29 December 1987
A repair team should be able to a: cess the break location or piping isolation valves and either isolate the break or effect temporary repairs.
FESW level could then be restored to normal.
There would be no immediate urgency to restore the level. The partially uncavered control rods only create a local problem. No offsite release is associated with this event..ictive FESW cooling would be lost during this accident, but as discussed in Section 9.4, considerable time is available to
.take action. Due to the lesser water volume to act as the heat sink and reduced fuel coverage, 12ss time would be available to restore cooling during this accident scenario than in just a loss of FESW cooling event, buc boiling
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would not commence for more than 1 day. As with the loss of FESW cooling event, if water is added to the FESW, any consequences of water heatup can be delayed er prevented. Water can be added from the Demineralized Water System or the Overhead Storage Tank.
9.6 Uncontrolled Waste Water Discharge This accideat postulates that an operator starts punping an unsampled or incorrectly analyzed Wsste Water or Retention Tank to the river. If the contents of the tank are of normal activity, this event will not be detected unt il the lineup is being secured af ter pumping, if then.
F If the liquid in the tenk is of high activity, the waste water monitor will alarm
.J the Auto Flow Control Valve will automatically close, terminating the discharge. The Turbino Condenser Cooling Water Monitor will also alarm, if the activity is high enough.
If the automatic valve does not close, an operator will try to close it from the Control Room.
If it cannot be closed, an operator will close a local valve or secure the pump to terminate the discharge.
After the discharge is terminated, a sample of the tank will be taken to analyze the radioactivity concentrations and effective MPC for the y
uncontrolled release. Waste water is diluted by LACbWR Circulating Water and Low Prcssure Service Water flow, in addition to circulating water from the adjacent coal-fired plant, prier to being discharged into the river.
The environmental consequence of this postulated event would be less severe thaJ that discussed in the original Environmental Report. The probebility of an event resulting in an actual discharge has been reduced by the installation of the Auto Flow Control Valve.
9.7 Loss of Offsite Power This accident postulates a loss of offsite power.
If both Emergency Diesel Generators and a High Pressure Service Water (HPSW) Diesel start, adequate FESW cooling can be provided and adequate instrumentation is available to monitor FESW conditions from the Control Roor All that is needed is for an operator to cross-connect HPSW to the Component Cooling Water (CCW) coolers.
If an HPSW Diesel and IB Emergency Diesel Generator start, EESW cooling can be provided.
If 1A Emergency Diesel Generator (EDG) starts, but 1B does not, adequate cooling can be provided only if the essential buses are tied tohether.
ER-SUFP 30 December 1987
l If one or more EDG's start, but neither HPSW' diesel starts, no ultimate heat sink for the FESW would be available. The consequences would be the same as in the Loss of FESW Cooling Event (Section 9.4).
If neither EDG can be started, neither FESW or CCW pump can run. The consequences again are the same as a Losa.of FESW Cooling Event, with the additional complication that some instrumentation will be lost immediately and others after the station batteries are depleted.
The operator would have tc check the FESW locally periodically.
As discussed in Section 9.4, the FESW would not start boiling for about 5 days, the tops of the control rods would become uncovered in 13.6 days'and the tops of the fuel would be exposei in 23.1 days.
Therefore, no immediate action needs to be taken and sufficient time is available to take corrective actions to restore power.
9.8 Seismic Event This accident postulates that a design basis earthquake occurs.
The magnitude of the seismic event and damage incurred is the same as that assumed during the Systematic Evaluation Program (SEP) and the Consequence Study prepared as part of the SEP Integrated Assessment (References 4-7).
The major concern of the previous evaluation was to safely shut down the plant and maintain adequate core cooling to prevent fuel damage. The focus now is to prevent damage to the fuel stered in ine Fuel Element Storage Well.
Seismic analysis has shown the Containment Building structure, LACBWR stack and Genoa Unit 3 stack are capable of withstanding the worst postulated seismic event at the LACBWR site.
Reference 1 documented that the storage well, itself, the racks and the bottom-entry line between the check valves and the storage well can withstand the postulated loads.
The potential consequences of most interest due to a seismic event could include loss of all offsite and onsite power and a break in the FESW System piping. This event, therefore, can be considered ar a combination of a Loss of Power Event (Section 9.7) and FESW Line Break (Sect $on 9.5).
As with these individual events, considerable time is available for response to a seismic event, with the FESW System pipe break requiring the earlier response. Access to the break location may be more difficult following a seismic event due to failure of other equipment in the plant. The time available, though, should be more than sufficient to initiate mitigating actions.
(Refer to Section 9.5).
9.9 Wind and Tornado This accident postulates that design basic high wind or tornado event occurs.
The magnitude of the event and damage incurred is the same as that assumed during the Systematic Evaluation Program (SEP) and the Consequence Study prepared as part of the SEP Integrated Assessment (References 4-9).
The major concern of the previous analyses was to ensure that adequate cooling of the reactor core was maintained. The focus now is to prevent damage to the fuel stored in the Fuel Element Storage Well.
ER-SUPP 31 December 1987
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The previous evaluations determiaed that the Containment Building would withstand this event.
The Turbine Building, Diesel Building, Cribhouse and switchyard may be damaged. The probability of the LACBWR or Genoa Unit 3 stacks failing and. impacting the Containment Building was_ determined to be low enough that it need not be considered.
Personnel outside the Containment Building may'not survive.
The potential plant consequence of primary concern is the loss of all offsite and onsite power. As discussed in Section 9.7, Loss of Offsite Power, considerable time is available before action must be taken to protect the fuel.
9.10 References
- 1) NRC Letter, Ziemann to Linder, dated Fnbruary 4, 1980.
- 2) NRC Letter, Reid to Madgett, dated October 22, 1975.
- 3) DPC Letter, Taylor to Document Control Desk, LAC-12377, dated September 29, 1987.
- 4) DPC Letter, Linder to Paulson, LAC-10251, dated October 11, 1984.
- 5) NRC Letter, Zwolinski to Linder, dated January 16, 1985.
- 6) DPC Letter, Linder to Zwolinski, LAC-10639, dated March 15, 1985.
- 7) NRC Letter, Zwolinski te Taylor, dated September 9, 1986.
- 8) DPC Letter, Taylor to Zwolinski, LAC-12052, dated January 14, 1987.
- 9) NRC Letter, Bernero to Taylor, dated April 6, 1987.
10.0 REGULATORY GUIDE 1.86 Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors," issued June 1974 currently provides the only NRC-approved guidance for license termination and acceptable releasable surface contaminstion levels. The NRC proposed rule on Decommissioning Criteria for Nuclear Faa.ilities redefines the decommissioning options listed in Regulatory Guide 1.86 and discusses the submittals required for plant decommissioning.
It does not, however, propose revised acceptable surface contamination levels, l
l.
Any material released from LACBWR for unrestricted use will therefore meet the established criterion of Table 1 of Regulatory Guide 1.86, which is reproduced here for reference.
l l
l l-l ER-Sl!PP 32 December 1987
a.
a ACCEPTABLE SURFACE CONTAMINATION LEVELS NUCLIDE a AVT \\GE b,c MAXIMUM b,d REMOVABLE b l
s s
U-nat. U-235, U-238, and 5,000 dpm a/
l 15,000.dpm a/
l' 1,000 dpm a/
associated decay products l.
100 cm2 j
100 cm l
100 cm2 2
l l
l
. 2
.Transuranics, Ra-226, l
100 dpm/100 cm l 300 dpm/100 cm l 20 dpm/100 cm2 2
l
'l Pa-231, Ac-227, I-125,.
l l
l I-129 ~
l l
.l l
l l
Th-nac, Th-232, Sr 90, l
1000 dpm/
l.3000 opm/
l 200 dpm/100 cm 2 Ra-223,-Ra-224, U-232, j
100 cm2 j
100 cu l
2 I-126, 1-131, I-133 l
l l
l Beta gamma emitters l
5000 dpm s-y/
l-15,000 dpm s-y/ l 1000 dpm s y/
l 100 cm2 (nuclides with decay modes l 100 cm l
100 cm2 2
other than alpha emission l l
l or spontaneous fission) l l
l
+
except Sr-90 and others l
l l,
noted above.
l l
l t
I a Where surface contamination by b(.,th alpha-and beta-gamma-emitting.nuclides exists, the limits established for alpha-and beta gamma-emitting nuclides should apply independently.
b As used in this table, dpm-(disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per min'ute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation.
c Mecsurements of average contaminant should not be averaged over more than 1 square meter. For objects of less surface area, the average shculd be derived for each such object.
d The maximum contamination level applies to an area of not more than 2
100 cm,
2 of surface area e The amount of removable radioactive material per 100 cm should be determir.ed by wiping that area with dry filter or soft absorbent paper, applying moderate pressure and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency.
When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.
D ER-SUPP 33 December 1987
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