ML19351F153

Draft Environmental Statement Related to Steam Generator Repair at Turkey Point Plant Units 3 and 4

ML19351F153
Person / Time
Site:	Turkey Point
Issue date:	12/31/1980
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML17340A519	List: ML17340A520 ML19351F153
References
NUREG-0743, NUREG-743, NUDOCS 8012290655
Download: ML19351F153 (48)

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{{#Wiki_filter:. O NUREG-0743 Draft Environmental Statement related to steam generator repair at Turkey Point Plant Units 3 and 4 Docket Nos. 50-250 and 50-251 Florica Power and Light Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation December 1980 ,.s = = s.,,, 5 O e 90 +\\

i NUREG-0743 4 DRAFT ENVIRONMENTAL STATEMENT RELATED TO STEAM GENERATOR REPAIR AT TURKEY POINT PLANT UNIT NO$. 3 AND 4 Docket Nos. 50-250, 50-251 FLORIDA POWER AND LIGHT COMPANY U.S. NUCLEAR REGULATORY COMISSION - 0FFICE OF NUCLEAR REACTOR REGULATION December 1980 a s e r-c -- m-,-, yme-y 9-g -g --w-m-- 9-e q-w w4 eyg-wp p --y.- en----, t-geyww -&y-- y9wqg-g.- r --yg -w w--- W g -w y-- v

This draft environmental statement was prepared by the U.S. Nuclear Regulatory Comission staff. This statement contains an environmental evaluation of the proposed steam generator repair program for Turkey Point Plant Unit Nos. 3 and 4 and alternatives thereto. For further information regarding this environmental statement, contact: Marshall Grotenhuis, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 (301) 492-7128 Comments on this draft statement must be received by the Director, Division of Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, by February 23, 1980, to be assured that they are taken into account in the preparation of the final environmental statement. ABSTRACT The staff has considered the environmental impacts and economic costs of the proposed steam generator repair at the Turkey Point Plant Unit Nos. 3 and 4, along with reasonable alternatives to the proposed action. The staff has con-cluded that the proposed repair will not significantly affect the quality of the human environment and that there are no preferable alternatives to the proposed action. Furthermore, any impacts from the repair program are outweighed by its benefits. SUPNARY In a letter dated September 20, 1977, Florida Power and Light Company (FPL) proposed to repair the steam generators in Units 3 and 4 of the Turkey Point Plant (the plant) (Sect. 2.0). The Nuclear Regulatory Commission staff (the staff) determined that the proposed program would require amending the FPL operating licenses for the plant, and on December 13, 1977, a Notice of the Proposed Issuance of Amendments to the licenses was published in the Federal i Register (42 FR 6259). A request for hearing has been filed and granted in connection with this proposed action. The staff issued a Safety Evaluation Report on the proposed program on May 14, 1979, and an Environmental Impact Appraisal on June 19, 1979 (Sect. 1.0). On August 3, 1979, Mr. Mark P. Oncavage was granted status as an intervenor. On March 4, 1980 the Commission issued a Memorandum and Order which ordered that an Environmental Impact Statement (EIS) l be prepared for the repair of the steam generators for the Surry Power Station Unit No. 1. In view of this order and the similarity of the two facilities, the staff decided to prepare an EIS for the Turkey Point steam generator repair. The primary impact in this environmental review is the occupational radiation exposure that the unit I repair program will entail (Sect. 4.1.1). The staff comparatively evaluated the environmental impacts of the proposed program and the following alternatives. i

(1) Continuation of the present mode of operation (5.0) (2) Shutdown and replacement of the unit with a generating plant of different design (5.0) (3) Decontamination of the steam generators before cutting (5.1) (4) Retubing the existing steam generators (5.2) (5) Installing tube sleeves in the existing steam generator (5.3) (6) Complete replacement of the steam generators (5.4) (7) Short-term storage of degraded steam generators before removal from site (5.5) (8) Immediate cut-up and shipment offsite (5.5) (9) Long-term storage with cut-up and shipment offsite (5.5) (10) Chemical decontamination followed immediately by cut-up and shipment offsite (5.5) The staff has concluded that the proposed program will not significantly affect the quality of the human environment. Furthermore, the staff found none of the alternatives to be obviously superior to the proposed program. The staff has also concluded that any impacts from the proposed repair program are out-weighed by its benefits (Sections 6.0). O J it

TABLE OF CONTENTS Page

SUMMARY

i 1.0 Purpose of this Environmental Statement.......................... 1-1 2.0 Background....................................................... 2-1 2.1 History of Steam Generator Operation............... 2-1 2.2 Reasons for Steam Generator Repair............... 2-1 2.3 Staff Initial Environmental Review............. 2-2 2.4 Major Environmental Impact............................... 2-2 3.0 Description of the Proposed Repair Method............ 3-1 4.0 Environmental Impacts of the Steam Generator Repair Project...... 4-1 4.1 Radiological Assessment..................................... 4-1 4.2 Economic Costs of Steam Generator Repair.................... 4-13 4.3 Nonradiological Environmental Assessment.................... 4-15 4.4 Environmental Impacts of Postulated Accidents............... 4-19

5. 0 Impacts of Alternatives..........................................

5-1 5.1 Decontamination............................................. 5-2 5.2 Retubing of Existing Steam Generators....................... 5-2 5.3 Installing Tube Sleeves in the Existing Steam Generators.... 5-3 5.4 Replacement of the Entire Steam Generators.................. 5-4 5.5 Alternate Disposal Methods.................................. 5-4 6.0 Conclusions...................................................... 6-1 7.0 References....................................................... 7-1 8.0 Federal, State, and Local Agencies to Whom this Environmental Statement Was Sent............................................... 8-1 Appendix A - Environmental Evaluation of Offsite Shipment of the Turkey Point Steam Generator Power Assemblies.......... A-1 W iii m m--- w++,---e u v' =

i list of Tables Page Table 4.1 - Comparison of Occupational Collective Whole-Body Dose Estimates............................................... 4-3 Table 4.2 - Occupational Dose at Pressurized Water Reactors (PWRs).... 4-4 4 Table 4.3 - Occupational Dose at Turkey Point......................... 4-5 Tabl e 4. 4 - Incidents of Job-Related Fatalities....................... 4-7 Table 4.5 - Radioactive Effluents from Steam Generator Repair and No rmal Ope rati o n........................................ 4-11 Table 5.1 - Options Considered........................................ 5-2 Table 5.2 - Steam Generator Repair Alternatives....................... 5-4 Table 5.3 - Steam Generator Disposal Alternatives..................... 5-7 Table 5.4 - Costs of Alternate Disposal Methods....................... 5-8 List of Figures Figure 3.1 - Typical Steam Generator.................................. 3-2 Figure 3.2 - Dose Rates Around Steam Generator 4A..................... 3-3 Figure 4.1 - Expo sure Pathways to Man................................. 4-9 Figure 4.2 - Turkey Point Plant Site.................................. 4-16 e iv

1.0 PURPOSE OF THIS ENVIRONMENTAL STATEMENT By letter dated September 20, 1977, Florida Power and Light Company (FPL) submitted a reporti entitled " Steam Generator Repair Report Turkey Point Units 3 and 4" (SGRR). This report has been supplemented by revisions 1-7 dated December 20, 1977; March 7, April 25, June 20, August 4, and December 15, 1978; January 26, 1979; and March 20, 1980, respectively. The report describes a proposed program to repair the six steam generators in Units 3 and 4 by replacing the lower assembly, including the tube bundles, of each generator. Our review of this program was completed and the Safety Evaluation (SE) and Environmental Impact Appraisal (EIA) were issued on May 14 and June 29, 1979, respectively. However, on March 4, 1980, the Commission issued a Memorandum and Order which ordered that an Environmental Impact Statement (EIS) be pre-pared for the repair of the steam generators of Surry Power Station Unit No. 1 (Surry Unit 2 steam generator repair was essentially complete at that time). In light of this Order, the staff concluded that an EIS should be prepared for the Turkey Point Plant steam generator repair. In addition, the original EIA (and also SE) was rendered partially obsolete by the issuance of Revision 7 to the Steam Generator Repair Report on March 20, 1980. This document changed several items including the method of cutting the steam generator, the preferred proposed disposal option, and the design of the storage facility. Also, the decision to install full flow condensate demineralizers by FPL and the necessity to assess endangered species required that the environmental evaluation be changed to include the impact of the installation and operation of these demineralizers and the impact of the repair on any endangered species in the vicinity. FPL plans to repair all six steam generators in Turkey Point Units 3 and 4. The Unit 4 steam generators have the most tubes plugged and, therefore, would be repaired first. The repair of Turkey Point Unit 3 steam generators is expected to be started about one year later. Since power demands in the FPL system peak in the summer, and the repair is expected to take from six to nine months per unit, the repair should be started in the fall-in order to be completed before the next summer peak demand. When FPL submitted the repair plan on September 20, 1977, the corporate plan was to be prepared to start the repair for Unit 4 in October 1978. The repair of Unit 4 steam generator is now scheduled to start in the fall of 1981 assuming that the hearing on that matter has been completed favorably. 1-1

2.0 BACKGROUND

The steam generator repair program proposed by FPL is similar to the one proposed by the Virginia Electric Power Company (VEPCO).4.s.s The two plants are similar. Each of the plants contain two Westinghouse three-loop PWRs and commenced commercial operation in 1972 and 1973. Both plants began operation using a sodium phosphate secondary water chemistry treatment and both plants changed to all volatile chemistry treatment (AVT); Turkey Point in late 1974, Surry in early 1975. The repair prog m of the Surry units was approved in January 1979. The Unit 2 repair was completed by May of 1980 and the Unit 1 repair commenced in September 1980. 2.1 History of Steam Generator Operation Turkey Point Units 3 and 4 began commercial operation on December 14, 1972 and September 9, 1973, respectively. Like almost all units with U-tube design steam generators, they began operation using a sodium phosphate secondary water chemistry treatment. Largely to correct a wastage and caustic stress corrosion cracking encountered witn the phosphate treatment, most PWRs with a U-tube design steam generator using a phosphate treatement for the secondary coolant have now converted to an all volatile chemistry (AVT). Both Turkey Point Units 3 and 4 were converted around Augast 1974. In 1975, radial deformation, or the so-called " denting," of steam generator tubes occurred in several PWR facilities following the conversion from a sodium phosphate treatment to an AVT chemistry for the steam generator secon-dary coolant. On September 15, 1976, during normal operation, one U-tube in the innermost row parallel to the rectangular flow slots in steam generator A at Surry Unit No. 2 rapidly developed a substantial primary to secondary leak (about 80 gpm). Subsequent to the 80 gpm leak at Surry Unit 2, the NRC has imposed augmented inservice inspection requirements on Surry Units 1 and 2, Turkey Point Units 3 and 4, San Onofre Unit 1, and Indian Point Unit 2. In addition, operating restrictions and limited periods of operation, typically six months, between inspections have been imposed at Surry Units 1 and 2* and Turkey Poi.,t Units 3 and 4. The augmented inspection requirements include an assessment of the magnitude and progression of tube denting,'and support plate deformation and/or cracking. 2.2 Reasons for Steam Generator Repair The six steam generators at Turkey Point Units 3 and 4 have all undergone a significant amcunt of degradation since they began operation. The wastage and denting phenomena, discussed earlier, have led to tube wall thinning, support " Unit 2 steam generators have been repaired and Unit 1 steam generators are now being repaired. 2-1

plate flow slot hourglassing and plate ligament cracking, tube denting, stress corrosion cracking, and several instances of reactor coolant leakage through cracked tubes. As of November 1980, tube plugging for various reasons has resulted in removing about 20.4% of the steam generator tubes in Unit 3 and about 24% of the tubes in Unit 4 frnm continuing service. Due to the continuing denting related problems, the certainty of additional tube plugging that will result in power derating, and the economic considera-tions for operating with substantially reduced heat transfer capacities on the two units, FPL submitted a proposal for the replacement of the degraded portions of the steam generators. 2.3 Staff Environmental Review In order to provide an independent basis for evaluating the radiological impacts associated with the repair of degraded steam generators at large pressurized water reactors (PWRs), the staff contracted with Battelle Pacific Northwest Laboratories (PNL) to perform a generic radiological assessment of the steam generator repair and disposal operations. This assessment has been published in an NRC report.2 This report has now been revised to include, among other things, an assessment of the method prepared by FPL in Revision 7 to the report, the channel cut method. Information useful to the environmental review was also obtained from the updated NRC staff SE3 on the repair project, particularly the sections evalu-ating (1) the effects of steam generator design changes, (2) the Radiological and As Low As is Reasonably Achievable (ALARA) considerations, and (3) the radiological consequences of postulated accidents. 2.4 Major Environmental Impact The major environmental impact is the occupational radiation exposure associated with the proposed repair of the degraded steam generators of the Turkey Point Plant Unit Nos. 3 and 4. e 2-2

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3.0 DESCRIPTION

OF THE PROPOSED REPAIR METHOD A drawing showing the principal parts of a typical steam generator is presented in Figure 3.1. Figure 3.2 shows the regions where the main cuts are proposed to remove the degraded steam generator. It shows also the radiation levels in these regions. A brief description of the FPL proposed repair procedure follows. A number of changes have been made in the materials, the design, and the operating procedure for the replacement steam generators to assure that the corrosion and denting problems will not recur. Among the more important of these changes are (1) using All-Volatile-Treatment chemistry control in the secondary system from the beginning of operation, (2) using corrosion resistant SA240 Type 405 ferritic stainless steel rather than carbon steel for the support plate material, (3) thermally treating the Inconel 600 heat exchanger tubes for better corrosion resistance, and (4) using a broached hole pattern with a quatrefoil design in the support plates rather than separately drilled flow holes to minimize the accumulation of corrosion products where the tubes pass through the plates. The staff review of the expected effects of the proposed changes is presented in detail in the introductory section of the SER3 for the repair project. We have concluded in the SER that the new steam generator design incorporates features to eliminate the potential for the various forms of tube degradation observed to date. FPL is planning to repair all six steam generators at the Turkey Point Plant Units 3 and 4. The units will be repaired in series; one unit will be conduct-ing normal power operations while the other unit is undergoing steam generator repairs. The repair will consist of replacing the lower assembly of each steam generator including the shell and the tube bundle and refurbishing and partially replacing the steam separation equipment in the upper assembly. The old lower assembly will be removed from the containment building through the existing equipment hatch and transported to a special storage facility that will be constructed on the Turkey Point site or transported by barge to a licensed disposal facility. The new lower assemblies arrived by barge and are now stored on the site. They were transferred to a wheeled transporter and hauled on the existing road to the temporary storage site. Prior to the repair work, the unit will be shut down and all systems will be placed in condition for long-term layup. The reactor vessel head will be removed for refueling. All of the normal procedures for fuel cooling and fuel removal will be followed. The fuel will be removed from the reactor and placed in the spent fuel storage facility, and then the reactor vessel head will be replaced. The equipment hatch will be opened and access control will be established. A special curtain, which would be able to reduce the size of the opening in the containment in case of an accident, will be installed in place of the door for l ease of deployment. A special vent exhausting through a HEPA filter will be l constructed. The biological shield wall and a section of the operating floor concrete and structural steel will be removed to provide access to the steam generator. Guide rails will be installed for transporting the lower assembly through the equipment hatch. 3-1 l

I Figure 3.1 - Typical Steam Generator l I STEAM OUTLET To TURBlft GEMRATOR M015TURE SEPARATOR ??.6. / MANWAY f " l _pUPPER SHELL SWIRL VAM MOISTURE f SEPARATOR t o REDWATER INLET / ~ pANTIVl8 RAT 10N BARS WRAPPER TRAN$lil0N COM iY I' LOWER SHELL f8E SUPPORTS -TU8E BUNDLE PARTITION g-f / TUBE SHEET pu, ' / MANWAY - j SUPPORT FOOT PRIMARY COOLANT OUTLET PRIMARY CO0lANT INLET CHANML HEAD 3-2

Figure 3.2 - Dose Rates Around Generator 4A EL G8 '- C " p Pl.A TE"l5N/ ELD /MG / i Y 50 - 75 me/hr SWiELD HALL -- (CourACT) 4' ', aj a.- 3rEA H 2-5 mr/hr GEUERA rOR (.GENEPAL ARGA) 4.A f (7 5 m r / h r COUTAcr) 30 mr/hr TkBE MEE r (MNERAL AREA) $i CHA NNEL \\ EL 30' 6 " O h "

1. 4 a

4o j 4 9 q + g, RC3 HOTLEG 50rr.r/hr 0 RCS PUMP SUCDOM (CurACT) lCOmr/hr 75 mr/hr (COMrACT) ~ (COprACT) i 75mr/hr l 100mr/hr (courAcr) (COMTAC r) 75 mr/hr ""' ^ '^ EL / 4 '- 0 " 3-3 .., _ = -.

After this preparatory work, the cutting of system piping will begin. This will include cutting and removal of sections of steam lines, feedwater lines, and miscellaneous smaller lines for the service air and water and the instru-mentation system. The steam generator will then be cut at the transition cone, and the upper shell will be removed and will be refurbished inside containment. After the channel cut at the bottom (see Figure 3.2), the lower assembly will be lifted from its support to the working level where it will be welded shut. Following this, the steam generator lower assembly will be lowered and placed in position on a transport mechanism. This mechanism will carry the assembly through the equipment hatch. A mobile crane will lift the lower assembly onto a transporter that will carry it to the steam generator storage facility on the site. The other two steam generator lower assemblies will be lifted from their location, welded shut, and lowered through the same hatch where the first steam generator was removed. After removal and storage of all three steam generator lower assemblies, their replacements will be transported from the temporary storage location to the equipment hatch. The same machinery used to remove the lower assemblies will be used to install the new assemblies in their cubicles. The steam generator lower assembly will be reinstalled and rewelded to the old bottom section. The upper assembly with its refurbished internals will be mounted on the lower assembly. Af ter welding the two assemblies together, the piping will be reconstructed. Following these major repair activities, there will be cleaning, hydrostatic testing, baseline inservice inspections, and preoperational testing of instruments, components, and systems. Then the reactor will be refueled and startup tests will be performed. The performance of the repaired steam generators will be tested for moisture carryover and verification of thermal and hydraulic characteristics. o 3-4

4.0 ENVIRONMENTAL IMPACTS OF STEAM GENERATOR 4.1 Radiological Assessment 4.1.1 Occupational Exposu_re 4.1.1.1 Introduction Our original evaluation 3 dated May 14, 1979 (and EIA dated June 29, 1979) was based on the pipe cut method of separating the steam generator from the reactor coolant system. Revision 7 to the Report,1 submitted on March 28, 1980, changed this to a channel cut method. The radiological assessment was changed signifi-cantly by this change in cutting method as well as by the input of Surry experience with a similar repair program. This evaluation represents an update based on this new information. 4.1.1.2

Background

The generic radiological assessment 2 of steam generator repair, prepared for the NRC by PNL, provides an upper bound estimate of the occupational doses likely to be associated with the repair of steam generators at a typical large PWR. Taking into account the conservatisms in PNL's method of assessment, described below, and the possibility of reducing occupational doses at Turkey Point by ALARA implementation, actual doses could be below the generic estimate. The PNL generic estimates of occupational dose (person-rem) were derived by taking into account expected dose rates and the person-hours needed for each maintenance activity. Descriptions of specific activities were developed by PNL as a composite of the work descriptions for removal and replacement of the l steam generators at Surry, Turkey Point, and Palisades as determined by VEPCO, l FPL, and Consumer Power Company. Person-hour estimates for each activity were developed by PNL based on prior experience with similar activities, using standard estimating techniques. Dose rates were based on information from several sources including data from several operating PWRs including the Turkey Point Units. PNL usually selected exposure rate values on the high end of the range of values measured at the several plants. The generic estimate of the total collective occupational wnvle body dose for the repair of three steam generators was presented 2 as a range of values, 2400 to 4700 persen-rem. Both ends of this range were conservatively estimated and l represent upper bound values for two sets of assumptions. The upper value, 4700 person-rem, was estimated assuming no credit for dose-saving techniques. l The lower value, 2400 person rem, was estimated taking credit for three specific l dose-reduction methods: (1) increased shielding by raising the steam generator water level; (2) use of remote cutting and welding devices; and (3) increased source-to-receiver distance. i 4-1 1

4.1.1.3 Effect of Cutting Method and Surry Experience on Occupational Exposure In an earlier submittal, FPL has developed an occupational exposure estimate of 1300 person-rem per unit to replace the steam generator by using the pipe cut method. This involved removal of the upper assembly by cutting the steam generator in the transition cone area and removing the lower assembly by cutting the reactor coolant piping (Figure 3.2). However, based en experience at the Surry Plant which resulted in 2140 person-rem, FPL increased their estimate of occupational exposure from 1300 person-rem to 2985 person rem per unit. The Surry experience has proven that the removal, refurbishment, and replacement of reactor coolant piping is time-consuming and dose-intensive. In addition, Turkey Point's pipe alignment problems are more severe than those at Surry. The steam generator supports at Turkey Point are at the base of the steam generators, thus hindering the movement of both the pipes and workers during cutting and refitting of the reactor coolant pipes. Taking these problems and the actual Surry experience into account, FPL recalculated the estimated occupational exposure to be incurred in making the repairs. Primarily due to the resulting significant increase in the time required for alignment of the reactor coolant pipes, the predicted dose for the pipe cut method was increased from 1300 person-rem to 2985 person-rem per unit. On this basis, they reevaluated the methodology and person-rem exposures associated with the pipe cut method and determined that an alternative approach would be desirable. 4.1.1.4 The Channel Cut Method The FPL alternative method is called the channel cut approach. It consists of removing the upper assembly, as before, by cutting in the transition cone area, and removing the lower assembly by cutting the lower assembly near the junction of the steam generator channel head to the tubesheet (Figure 3.2). This approach results in estimated doses of 2084 person-rem per unit. A major part of the person-hours for the channel head cut will be spent in radiation fields which are an order of magnitude smaller than those involved in the reactor coolant pipe cuts. The lower radiation fields will result in a savings of 900 person-rem per unit. The FPL total estimate of 2084 person-rem per unit for the channel cut method takes into account the dose reduction measures described in Regulatory Guide 8.8, Revision 3, "Information Relevant to Ensuring That Occupational Radiation Exposures At Nuclear Power Stations Will Be As Low As Is Reasonably Achievable,"7 which include local decontamination, temporary lead shielding, pre-job planning, pre-job training, and use of remote tools where practicable. Taking into account the FPL commitment to undertake dose-reducing activities, the lower end of the generic range, 2400 person-rem per unit, is the appropriate estimate for comparison with the FPL estimate of 2084 person-rem per unit. A summary comparing FPL estimates with our generic estimates 2 for the four main phases of the project is given in Table 4.1. 4-2

TABLE 4.12 COMPARISON OF OCCUPATIONAL COLLECTIVE WHOLE-BODY DOSE ESTIMATES Phase NRC Generic Estimate FPL Estimate Dose, person-rem / unit Dose, person-rem / unit O Preparation 730-830 283 Removal 780-1200 1016 Installation 620-2300 644 Miscellaneous

• 260-390 141 TOTAL 2400-4700 2084
• Miscellaneous - includes cleanup and storage.

The range of values given in the above estimates result from differences in procedures and radiation dose rates from site to site. FPL used commonly accepted practices for calculating doses and took into account the dose reduc-tion measures proposed to maintain doses as low as is reasonably achievable (ALARA), including local decontamination, temporary lead shielding, pre-job training, and use of remote tools where practicable. In Section 3 of its report,1 FPL has documented its consideration of the guidance with regard to ALARA issued in Regulatory Guide 8.8, Rev. 3, and is responsive to the Regula-tory Staff Positions in Regulatory Guide 1.8, "Personr.el Selection and Training," 8.2, " Guide for Administrative Practices in Radiation Monitoring," and 8.10, " Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable." We conclude that the FPL efforts to maintain occupational doses ALARA during the repair efforts meet our positions in Regu-latory Guide 8.8 and are therefore acceptable. In summary, the above discussion shows that the differences between the lower generic estimate (2400 person rem per unit) and the FPL detailed estimate (2084 person-rem per unit) are well within variations expected due to differing pro-cedures and radiation exposure rates from site to site. We therefore believe that the FPL detailed estimate of 2084 person-rem per unit is a realistic estimate for the repair of the steam generators in one Turkey Point unit. Consequently, in the remainder of the statement, we have used 2100 person-rem

• per unit as the projected occupational dose for the steam generator repair work at Turkey Point.
• The number 2084 does not indicate four significant figure accuracy; rather, it represents a sum of several numbers of varying magnitudes.

Therefore, we use 2100 person-rem as the projected occupational dose. 4-3

4.1.1. 5 Occupational Dose due to Normal Plant Operation To put into perspective the occupational doses to be incurred in repairing steam generators, it is helpful to compare these doses with (1) those from the normal operation of nuclear plants, (2) the projected long-term person-rem reduction resulting from steam generator repair, and (3) the doses from major maintenance operations at other plants. Although the Atomic Energy Commission (predecessor to the Nuclear Regulatory Commission) (AEC), was starting to compite occupational exposure estimates for nuclear power plant operation at the time that the Turkey Point FESS was pre-pared in 1972, such exposures were not specifically considered in the Turkey Point FES. In recent environmental statements for pressurized-water reactors (PWRs), we have estimated an annual average of 410 person-rem per reactor, from 1975 thru 1979, with particular plants experiencing a lifetime average annual dose as high as 1300 person-rem. These dose averages are based on widely varying yearly doses at PWRs; for example, annual collective doses for PWRs have ranged from 18 to 5262 person-rem per reactor. The annual average is based on reported data from operating power reactors. A summary of the data is provided in Table 4.2. TABLE 4.217 OCCUPATIONAL DOSE AT PRESSURIZED WATER REACTORS (PWRs)* (person-rem per reactor unit) Year Average Low High 1975 318 21 1142 1976 400 74 1583 1977 396 87 1154 1978 428 48 1621 1979 510 30 1792 AVG. 410 52 1458

• For those sites with more than one reactor, the number of person-rem per reactor is obtained by dividing the number of person-rem reported for the site divided by the number of reactors on the site.

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Those data show that 410 person-rem per reactor unit is roughly the average of the wide range of annual doses incurred at all PWRs. This annual dose is highly dependent on the amount of major maintenance required in a given year, which results in doses well above the average of 420 person-rem; i.e., in 1979 Surry was the first nuclear plant to replace a steam generator. The total occupational dose received by workers at Surry Unit 2 during the repair program was 2140 person-rem. The VEPC0 original estimation was 2070 person-rem. The 2100 person-rem is within the range of doses for one unit a year, and is a small fraction of the occupational dose to be expected over the life of the plant. Table 4.3 shows that workers at Turkey Point received whole-body doses of 335 to 600 person-rem (combined totals for both units), respectively, during the inspection and plugging of degraded steam generator tubes. The total occupa-tional doses for the two units combined from 1976 to 1979 are also given. These doses are comparable to the 410 person-rem per unit per year average for U.S. PWRs in 1979. TABLE 4.317 OCCUPATIONAL DOSE AT TURKEY POINT Annual Person-Rem Avg. Steaml Exposure Total Total Generator Rem per Year Personnel Rem Operations Maintenance Inspection Person 1976 1647 1184 89 495 600 0.72 1977 1319 1036 94 492 450 0.78 1978 1336 1032 90 607 335 0.77 1979 2002 1680 299 1046 335 0.84 ISteam gene-ator inspection and plugging of degraded steam generator tubes. 4.1.1.6 Risks due to Occupational Exposure The individual risks associated with exposures involved in the repair program will be controlled and limited so as not to exceed the limits set forth in 10 CFR Part 20 for occupational exposure. These limits are intended to assure that the hazard to any exposed individual is extremely small. The following estimates of the risks to workers are based on conservative assumptions (i.e., the estimates are probably higher than the actual number). The risk estimates are derived from the recommendations of the National Academy of Sciences Biological Effects of Ionizing Radiation Committee and WASH-1400.18'19 4-5

The average annual dose per nuclear plant worker at Turkey Point (about 0.8 rem) has been well within the limits of 10 CFR Part 20. However, for compara-tive purposes, the NRC staff has estimated the risk to nuclear power plant workers at Turkey Point. The nuclear plant workers' risk is equal to the sum of the radiation-related risk and the nonradiation-related risk. The occupa-tional risk associated with the industry-wide average radiation dose is about 11 potential premature deaths /105 person years.* The nonradiation-related fatality incidence of nuclear plant workars is expected to be no greater than the fatality incidence for similar types of work. The average nonradiation-related risk for 7 U.S. electrical utilities over the period 1970-1979 is about 12 premature deaths /105 person years.3 Adding the nonradiation-related risk to the radiation-related risk, the comparable risk to a nuclear power plant worker receiving the average annual dose would be about 23 premature deaths per 105 person years. The risks of various occupations, including nuclear plant workers, are shown in Table 4.4. In terms of job-related fatalities, the occupational risk to a nuclear power plant worker (i.e., about 23 potential premature deaths /10s person years) is higher than the average private sector risk (i.e., 10 premature deaths /105 person years). However, the risk to nuclear plant workers is lower than the risk for a number of other groups. It should be pointed out that the fatality incidence rates due to radiation exposure in Table 4.4 for the nuclear power plant.<orkers are conservative estimates (i.e., the actual risk may be much less than the estimate), whereas the fatality incidences for other groups are based on known instances of job-related fatalities. In addition, it should be noted that the values in Table 4.4 reflect the degree of study of the respective occupational groups. For example, the underground metal miners, the uranium miners, the smelter workers, and radition exposures of nuclear plant workers have been much more extensively studied than the other occupational groups. The amount of data on job-related fatalities due to long-term stress in the other accupations is less complete because these groups have not been as extensively studied. However, based on the above comparisons, the staff concludes that the risk to nuclear plant workers from the steam generator repair is comparable to the risks associated with other occupations. The following risk estimators were used to estimate health effects to the entire work force for this repair program: 135 potential deaths from cancer per million person-rem and 258 cases of all forms of genetic disorders per million person-rem. Multiplying the population dose by the risk estimators, the NRC staff estimates that there may occur 0.6 cancer deaths in the exposed population and 1.1 genetic disorders in all future generations of the exposed population. These health impacts will not be measurable when spread over the lifetime of the entire work force. 4.1.1. 7 Conclusinn In summary, we have drawn the following conclusions. The FPL estimate of 2084 person-rem per unit for the repair of the steam generators is reasonable. This dose falls within the range of annual occupational doses which have been

• Exposure to individual workers will vary from the average; however, exposure to individual workers will be limited so as not to exceed the limits in 10 CFR Part 20 for occupational exposure.

4-6

TABLE 4.4 INCIDENCE OF JOB-RELATED FATALITIES Fatality Incidence Rates Occupational Group (premature deaths /105 person year) a Underground Metal Miners 1275 a Uranium Miners 422 a Smelter Workers 794 c Mining 61 c Agriculture, Forestry, and Fisheries 35 c Contract Construction 33 b Nuclear Plant Worker 23 d Transportation and Public Utilities 24 c Manufacturing 7 c Wholesale and Retail Trade 6 c Finance, Insurance and Real Estate 3 c Services 3 e Total Private Sector 10 a"The President's Report on Occupational Safety and Health," May 1972. The fatality incidence rate for nuclear plant workers is based on an annual exposure of 0.8 rem to the average worker, and the nonradiation-related fatalities for 7 large U.S. electrical utilities over the period 1970-1979.38 About half of the estimated fatality incidence rate for nuclear plant workers is potential, rather than actual, premature deaths that might be caused by radiation exposure. c" Occupational Injuries and Illness in the United States by Industry, 1975," Bureau of Labor Statistics, Bulletin 1981, 1978. 4-7

observed in recent years at operating reactors. Our review concludes that FPL is taking the necessary steps to insure that occupational doses will be main-tained ALARA. FPL will use some experienced personnel from the Surry Unit 2 steam generator removal and replacement. These individuals will provide added expertise to the FPL for dealing with health physics eroblems associated with the repair. Communication of this knowledge gained A. ring the Surry Unit 2 operation is a key ingredient in an effective ALARA piogram. Based on the above evaluation, we determine that the programs and procedures proposed by FPL in making the steam generator repairs demonstrate that it will meet (1) 10 CFR Part 20 limits and requirements, including efforts to maintain radiation exposure as low as reasonably achievable; (2) Regulatory Guide 8.8 as it relates to management policy and organization; personnel qualifications and training; design of facilities and equipment; radiation protection program, plans, and procedures; and the availability of supporting equipment, instrumen-tation, and facilities; and (3) Regulatory Position C.1.f of Regulatory Guide 8.10 on modifications to reduce radiation exposures. 4.1.2 Public Radiation Exposure This section contains conservative estimates of the impacts on the public from the proposed steam generator repair project. The major sources of radiation and environmental pathways were considered in preparing this section; these sources of radiation and environmental pathways are shown in Figure 4.1. The section includes doses from radioactive effluents released during the steam generator repair, doses from the storage or disposal of solid radioactive wastes and the impacts due to solid waste storage. 4.1.2.1 Doses from Effluents Public radiation exposure from the Turkey Point steam generator repair can be estimated by comparing the estimated quantities of radioactive effluents from the steam generator repair with annual average releases and dose estimates from normal operations at Turkey Point. Estimates of the gaseous and liquid releases of radioactivity from the steam generator repair at Turkey Point can be obtained from three sources: (1) the generic report,2 prepared by PNL for NRC; (2) esti-1 and (3) measured releases mates made by FPL for the steam generator repair: from the steam generator repair at Surry Unit 2.8 Estimates of annual average releases and doses are documented in the Turkey Point FES.9 Radicactive effluent releases from the steam generator repair and normal reactor operations are presented in Table 4.5. Table 4.5 indicates that, in general, the estimates of radioactive effluents from the repair from three sources of information!'2 23 are less than the releases from normal operations at Turkey Point. The Turkey Point FES9 did not contain separate estimates of the quan-tities of airborne particulates and tritium. However, the FES (Table V-6) indicated that airborne effluents contributed a small fraction of the total dose to individuals (i.e., less than 5% of the total body dose for the highest dose pathways of exposure to individuals in the general public). Therefore, the total doses to maximum individuals from the steam generator repair at Turkey Point will be no greater than the doses estimated in the Turkey Point FES. The Turkey Point FES estimated an annual total body dose of about 6 mrem, and 4-8

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a highest annual dose to any organ of about 16 mrem (thyroid) from the operation of Units 3 and 4. Consequently, the dose to the maximum individual from radio-active effluents released during the steam generator repair of one unit should be less than about 3 mrem and 8 mrem to the total body and thyroid, respectively. Since the estimated dose to the maximum individual from the repair work is a small fraction of both the annual dose (about 63 mrem)20 and lifetime doses (about 4400 mrem) from exposure to background radiation in the Florida area, we conclude that the dose to individuals offsite due to radioactive effluents from the repair work will not be environmentally significant. The FES9 for Turkey Point (Table V-7) estimated an annual total body dose to the population within 50 miles (about 2.1 million person) of about 12 person-rem from operation of Units 3 and 4. Since the effluent releases from the steam generator repair are less than those estimated in the FES for normal operation of one unit, and even allowing for a 50% increase in population in the past 10 years, the population dose from the repair effort should be less than about 9 person-rem to the total body. Since the population dose from the repair work is a small fraction of both the annual population dose (about 130,000 person-rem) and the lifetime population dose (about 9,300,000 person-rem) from exposure to background radiation, the staff concludes that the dose to the offsite population within 50 miles due to radioactive effluents from the repair work will not be environmentally significant. 4.1.2.2 Impacts from Solid Wastes (Not Including the Steam Generator Assemblies) The environmental impact of solid radioactive wastes from the Turkey Point steam generator repair can be estimated by comparing the estimated quantity of solid wastes from the steam generator repair with annual average releases from normal operations. FPL has estimated that the repair effort will generate about 1100 cubic meters of solid waste per unit containing about 130 to 270 curies of radioactivity depending on the effectiveness of the decontamination of the channel head.22 The generic report 2 estimates that about 760ma of low activity wastes would be generated per steam generator; this is equivalent to about 2300m3 of wastes for a three-steam generator unit such as Turkey Point.2 The steam generator repair at Surry Unit 2 produced about 1600m3 of solid wastes containing about 65 Ci of activity.23 None of the preceding estimates include the radioactivity on the inside surfaces of the old steam generators. In the years 1975 through 1977, Turkey Point generated an annual average of about 575 cubic meters of solid waste per unit containing about 170 curies per unit.10 Consequently, the activity of solid wastes from the repair will be about the same as the annual amount from normal operations. Since the estimated activity j of wastes from the repair are comparable to the amount of wastes from normal operations, the impact from the solid wastes should be about the same as that t from normal operations, and not environmentally significant. On the basis of long-term onsite storage of the degraded steam generators until the reactors are decommissioned, there will be essentially no radioactive i effluents from the steam generators for 30 years. After 30 years of storage l of the steam generators, final disposal will result in very small offsite gaseous and liquid radioactive releases because most of the radionuclides in the steam generators will have decayed to very low levels. l l l 4-10 l I

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Table 4.5 RADI0 ACTIVE EFFLUENTS FROM STEAMGENERATOR REPAIRS AND NORMAL OPERATIONS Radioactive Effluents a b Steam Generator Repairs, Ci/ Unit Normal Operations, Ci/Yr/ Unit Estimates Measured FES Estimates Measured Type of Radioactive NUREG/ Surry Annual Effluent FPL CR-1595 Unit 2 Average AIRBORNE Noble Gases Negligible 100 3650 8700 Halogens (Iodines) 0.0105 included in 0.0000069 0.8 0.3 c particulates j Particulates 0.0428 0.00021 0.0013 0.025 ~~ Tritium Negligible 4.3 2.1 LIQUID Mixed Fission & 0.55 0.23 0.5 28 3.4 Activation Products Tritium 185 190 8.5 1000 415 aValues for radioactive effluents from steam generator repairs are taken from References 1 (sect. 5.2.2), 2 and 3. i bValues for radioactive effluents from normal operations are taken from References 9, 10, 8, and 21'. The measured effluents from normal operations are average values based on effluent monitoring for the years 1975-1977. cThe FES8 value includes both iodines and particulates.

4.1.2.3 Imoacts from Solid Wastes (Steam Generator Assemblies) The environmental impact of the steam generator assemblies can be estimated by comparing that volume of waste to the average annual amounts of waste resulting from normal operation. The steam generator assemblies represent a volume of 3 3 about 20,000 ft (540m ). This is about equal to the 1975-77 annual average of 575m3 from one unit. The 1973-79 annual average is about the same, 980m3 f.om both units. The activity contained in the steam generators at the time of shipment is about 1500 curies. This is about four times the annual average of 340 curies (1975-77) or 430 curies (1973-79) but small compared to the 9,100 curies projected in table 5 of CFR Part 51. Considering that this activity level is over nearly two years and is very well contained for the thirty-five years it takes to decoy the impact is considered acceptable. The dose to the public received during the s.G ment of these assemblies is ? conservatively estimated to be least than 2 person-rem. (The details of the estimate may be found in Appendix A.) Based on the above evaluation the staff concludes that the dose to the offsite population due to the shipment of the steam generator assemblies will not be environmentally significant. 4.1.2.4 Doses from Onsite Storage of Steam Generator In addition to considering the environmental impacts of effluents and solid wastes to be shipped offsite, the staff has estimated doses from the storage of the steam generators at the site (should that alternative be necessary). FPL estimates that each steam generator will contain about 250 Ci of gamma radioactivity. This estimate of total gamma activity is based on measured levels of contamination in the steam generator primary side and allowing about 80 days of radioactive decay prior to removal from containment. The staff estimated a dose rate of less than 0.0001 millirem per hour at the nearest site boundary due to the onsite storage of six steam generators. An individual spending an entire year at this location would receive less than 1 millirem of radiation exposure. This dose would decrease by a factor of about 2 every five years because of the decay of Co-60. FPL made similar calculations and reached similar conclusions concerning offsite doses. Since the dose conser-vatively estimated from storage of the steam generators represents less than 2% of the annual dose from natural background radiation, the staff concludes that the impact to the public from the stored generators will be very small, and not environmentally significant. ~ 4.1.2.5 Effect of Repair on Future Normal Oceration The repair effort will return the plant to the design conditions on which the staff's evaluation in the FESS was based. Therefore, the staff concludes that the quantity of radioactive materials released from normal operations after the repair should not be significantly greater than those presented in the FES. Therefore, the potential doses to the public and the impact on biota other than man from those materials will be no greater than the doses and impacts presented in the FES. 4.1.2.6 Summary In summary, the offsite doses resulting from the steam generator repair will be less than those from recent plant operation since the estimated releases of 4-12

small compared to the annual doses from natural background radiation. Therefore, the radiological impact of the repair project to the public will not significantly affect the human environment. 4.2 Economic Costs of Steam Generator Repair Steam generators generally are built with more tubes than necessary as a margin to allow for any tubes that may have to be plugged. We have evaluated the Turkey Point Plant and find that each unit can operate safely with up to 25% of the steam generator tubes plugged. If the percentage of plugged tubes gets high enough so that there is not enough heat transfer surface, the unit will have to be operated at some level of power less than 100%. If the unit is required by physical restraints such as the heat transfer to operate at some lower level of power, the operation is referred to as derated. In addition to the percentage of plugged tubes, the nuclear peaking factor, Fq (a nt.mber which is related to the uniformity of the neutron flux over all positions in the reactor core), imposes limitations on the unit, and depending upon the fuel burnup can also cause the plant to operate in a derated mode. Based on the above discussion and the latest amendments to the Technical Specifications,"'15 it is possible that Unit 4 may operate in a derated mode for at least part of cycle 7, which is expected to begin in January 1980. Unit 3 has usually had about 3% fewer plugged tubes than Unit 4 and therefore is likely to be operated in a derated mode about one year after Unit 4. Over the life of the plant, the proposed steam generator repair project will result in a net dollar savings of at least $500,000,000 compared with the cost of centinued operation of the existing steam generators, with an optimistic assumed scenario of tube plugging and derating. The cost of purchasing and installing the steam generator lower assemblies and associated activities is estimated at about $119,000,000 for the two units. A full-flow condensate polishing demineralizing system will be installed during the repair program. This system will demineralize 100% of the condensate system's circulated flow to the steam generators..The estimated cost of the condensate polishing system is $3,000,000 for both units. The cost of disposal of the six degraded lower assemblies is expected to be about $3,000,000. The estimate for replacement power during the outage for repair is about $145,000,000. The total project cost is therefore about $270,000,000. The cost of replacement power during the outage is based on the FPL estimate of $535,000 per day per unit and an outage duration of 270 days per unit. The FPL estimate of $535,000/ day / unit based on differential fuel costs is reasonable in view of the fact that the replacement power would be provided by oil and gas-fired units which FPL would press into service (690,000 kW x 24 hrs / day x a fuel differential cost of about 0.038/kW hr* x 0.85% capacity factor = j $535,000/ day / unit). We consider this replacement power cost estimate reasonable. The estimated net saving of $500,000,000 is based largely on the cost of replacement capacity. The replacement power cost for both units would be about " Based on oil at $25/ bbl, 6.1 x 108 Btu /bbi and nuclear fuel cost of $4/MWh. l i 4-13

$858,000,000 for only 10 years of derated operation at an assumed derating rate of 3% per year beginning when 25% of the tubes were plugged. If the derated period lasted longer, the cost would be larger.* The calculation was made as follows. For the first year of derating the cost would be, f(1b5)x 0,000,000 = 15,600,000 The 0.03 corresponds to 3% derating per year, the 1-0.25 term corresponds to the number of remaining sound tubes after 25% are plugged. The $390,000,000 is the yearly cost of replacement power due to fuel differential cost for two units at $535,000 per day per unit. By the end of the second year of derating, the cumulative cost would be three times as high ($46,800,000) since the first 3% derated batch of tubes would have been out for two years and the second 3% derated batch of tubes would have been out for one year, for an effective total of three years of 3% derating. By the end of the third year of derated opera-tion, the cumulative cost would be six times the first year cost (3+2+1). After ten years the cumulative cost would be, (1b25)x$390,000,000,000x55=$858,000,000. (55 = 10 + 9 + 8 + 7 + 6 + 5 + 4 + 3 + 2 + 1) Therefore, the estimate that $500,000,000 would be saved over the life of the plant, even after spending $119,000,000 for the steam generator repair, is conservative. The present value of the replacement power assuming a net discount rate of 3% (corresponding to a discount rate of 10% minus an inflation rate of 7%) would be about $690,000,000. The estimate of $3,000,000 for final disposal of the degraded steam generators assumes either onsite storage for 30 years followed by sectioning and shipment to a licensed burial facility for low level waste or immediate shipment by barge to a licensed land burial facility. Radiological and cost considerations associated with short-term barge shipment and long-term storage and decommis-sioning are approximately the same. This consideration of costs does not take into acenunt the continuing costs of tube inspection and plugging services, nor the costs of possible future modifi-cations to control corrosion if the repair is not done. It also does not consider the cost of the reduced generating capacity and the current lack of reliability and availability. In 1978, the approximate outage times for steam generator tube inspection and plugging were 10.5 days for Unit 3 and 27 days for Unit 4. Experience at the Turkey Point Plant indicates that such an inspection takes about 10 days when combined with a refueling cutage and about 21 days when not combined with a refueling outage. Inspections have been carried out about two times per year. " Assuming all degradation has occurred since 1975, the annual degradation rate (based on plugging rate) has been at an average of 4.1% and 4.8% for Unit 3 and Unit 4 respe. :tively. This annual rate was larger from 1977 to 1978, about 5.5% and 5.7% respectively, but has decreased to about 2.5% in 1979 and 1980. 4-14

4.3 " nradiological Environmental Assessment 4.3.1 Construction Impacts Nonradiologically related construction activities have been evaluated for their potential to impact both aquatic and terrestrial species occurring at the Turkey Point site. The following presents a discussion of these activities and an assessment of their potential impact on organisms inhabiting the site. An area approximately 2 acres in size located along the south side and inside the site compound (Fiqure 4.2) will serve as the steam generator storage cc: pound (SGSC) where the existing steam generators may be kept in permanent stor a after removal. The SGSC will be at elevation 17.5' (present elevation +d ) with side slopes of 1:3. Fill required to bring this area up to grade will bt obtained from onsite spoils piles created during canal construction. An existing at grade area within the site compound also at the southern boundary is presently being used as a laydown area for the new steam generators. No additional site preparation is planned. Both the SGSC and the new steam generator laydown area are surrounded on three sides by the canal system and on the fourth side by the station. Standard construction practices, including erosion control procedures and revegetation where appropriate, will be employed. Site runoff during and after construction is towards Loch Rosetta, north of the site, and East Canal, east of tne site. These canals flow into the station intake. No runoff from the disturbed sites will flow into Biscayne Bay. Filling the SGSC to grade may result in some slight increase in turbidity in the East Canal and Loch Rosetta for a brief period after rainfall during the construction phase; however, this is expected to be minor and temporary in nature. Once the site is brought to grade the cement-like properties of the limestone marl used as fill and the bank slope will reduce erosion due to runoff to an insignificant level. The use of dredge spoils located on the banks of the cooling canal system for fill may result in increased turbidity in the canal system in the vicinity of the activity for a short period after rainfall; however, this also is expected to be minor and temporary in nature. FPL estimated 1 (Section 5.4.1) the order of magnitude sound pressure levels (SPL) typical of construction noise sources at the nearest site boundary (0.8 mile). Estimated values of SPL at the nearest site boundary ranged from 40 to 58 dBA for a variety of construction equipment. Noise resulting from the repair program for the steam generators may deter avian fauna from utilizing the l immediate vicinity of the plant; however, the large size of the site, the ability of the organisms to acclimate to higher than normal sound pressure level (SPL),24 and the temporary nature of the repair program leads one to conclude that no long-term adverse impact is expected. i l Dust created by movement of vehicular traffic in unpaved areas will be sbated by periodically spraying the road surface with water. The bringing of the SGSC up to grade will result in (1) some slight increase j in turbidity in the cooling system after rainfall in the vicinity of the plant and the source of the fill material, (2) increased constructional noise, (3) increased dust due to vehicular traffic. Each of these potential impacts are 4-15 l 1

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expected to be minor, temporary, and to a great extent controllable by the use of standard erosion abatement methods, inspections, and the use of sprinkler trucks to minimize dust. Due to the limited use of the site compound by orga-nisms known to inhabit the Turkey Point site, the rather small area of the compound in relation to the entire Turkey Point site that will be affected by such activities and the presence of an abundance of nearby suitable habitat, no detrimental effect on any species is expected. During the constructional phase of the steam generator repair approximately 300 additional people will augment the existing site workforce. This is typical of the number of additional people required for refueling and major maintenance activities associated with normal operation. The repair activities will occur in areas that lack readily accessible sanitary facilities. Portable sanitary units will be utilized and all sanitary wastes will be removed from the site.2s No significant increase in sanitary wastes discharged from the station or modification to the existing sanitary facility will occur; therefore, no incre-mental impact from sanitary facilities to species inhabiting the Turkey Point site is expected. FPL has indicated 1 (Section 5.2.2.4) that during the repair procedures laundry releases will be approximately 22,000 gal / day for about 300 days for each unit. Although 22,000 gal / day was the conservative value used to estimate radioactive wastes, it is expected that on the average only 10,000 gal / day will be discharged. Laundry waste water is and will be discharged without processing. The volume of discharge of laundry wastes, though not significantly different from that during normal refueling outages, but discharged over a significantly longer period of time (6-9 months / unit), even under the most conservative value of 83,277 1/ day (22,000 gal / day) represents only 0.0008% of the total daily plant flow for four units. Even with multiple unit outage, the laundry waste stream would still represent considerably less than 0.01% of the total discharge flow. Due to turbulent mixing in the discharge channel, the small volume of the waste in relation to station flow, the large volume of water in the cant.1 system, the short duration (s6-9 months per unit) of the increased laundry waste flow, and the use of a phosphate-free biodegradable laundry detergent, no detrimental impact on any species is anticipated. It is concluded that construction activities associated with the steam generator repair program will not have a significant adverse or detectable impact on species known to inhibit or utilize the Turkey Point site. 9 4.3.2 Operational Impacts Included in the steam generator repair program is the installation of the full-flow condensate polishing demineralizer. This system will become operational once the repair has been completed and the units begin to operate. This is the only nonradiological modification to the plant that will be materially different from those identified in the Turkey Point FES.S The function of this system is to demineralize the entire volume of condensate water prior to its reentry into the steam generators. It is anticipated that the removal of suspended solids and ionic species from the condensate water will reduce corrosion-related phenomena. The demineralizers will employ a mixed-bed ion exchange resin 2s (Powdex) in a series of vessels. Condensate water will be circulated into the vessels and through the exchange beds. Periodic replacement j of the resins will occur due to a buildup of suspended solids and exhaustion 4-17 4

1 1 ,I l t 4 r 4 i of the ion exchange resins. Replacement of spent resins in the full-flow l condensate polishing demineralizer vessels will result in the periodic discharge 3 J of a waste steam int.) the canal system. Resin vessels will be backflushed to a backflush receive? tank and resins replaced periodically depending upon the 3 buildup of susperced solids and ion-exchange capacity exhaustion. Replacement i of spent resins in an on-stream vessel will occur about every two to three j weeks.2s A total of eight vessels will be employed for both units, four per l unit, three operating at a given time, with one per unit in reserve. Maximum resin loading is approximately 136 kg (300 lb) dry weight per resin vessel per cycle. Backflushing will occur at an anticipated frequency of one per week per unit or two vessels per week for the plant. Approximately 24,000 1 (6,350 gal) of high purity water from the condensate storage tank will be used to back-flush each cemineralizer resin vessel.28 The spent resins will be backflushed i 1 to a receiving vessel for resin water separation. The spent resin handling i subsystem is cesigned to process the backwashed resin slurry discharged from the resin vessels using a filtration system. After resin-water separation, ] the supernatant liquid (approximately 22,700 1 [6,000 gal]) will be discharged to the canal system. The quality of the waste water prior to discharge is t j predicted to have a pH between 8.5 and 8.7, a dissolved oxygen concentration t of 0.08 pom, and a conductivity about 1 paho/cm.2s.27 The concentration of total suspenced solids in the waste stream is expected to be significantly less than low volume waste source limits (Title 40 CFR 32.22 - 100 mg/l instantaneous max., 30 ppm monthly average). Tce waste water will be released at a rate of less than 0.0009 m /sec (15 gps) 3 from the discharge structure to the discharge canal that leads into Lake Warren, j a receiving pond,2s and then into the canal cooling system. t Uncer normal operating conditions full cooling water flow for all four units is approximete 114 m /sec (1.8 x 105 gpm). The waste stream from the demineral-3 izer will discharge for about 7 hours per week for eacn unit. The waste steam will represe-t about 0.0008% of the total discharge flow into Lake Warren at full four-uni t operation. Even with multiple unit outage and simultaneous l i discharge of waste stream from both units, the waste would still represent less than 0.01% of the total discharge flow. Turbulent mixing in the discharge channel leading to the receiving pond is expected. Mixing is further enhanced i by the geometry of the receiving pond. Due to the high purity of the cackflush j waste water stream and the anticipated water quality previously du cribed, no measurable effects on aquatic organisms are anticipated.29*3o The total demineralizer backflush discharge volume for both units over the life of the ~ 4 plant would represent less than 0.5% of the total present volume of water in i the canal system. No long-term degradation of water quality in t M tanal system ^ l due to concentration of pollutants in the waste stream system over tr;e life of i the plant is expected. The small amount of resins that may fail to be removed from the waste water prior to discharge poses no environmental threat: the resins are highly insoluble, resistant to biological degradation, have no effect-i on biological oxygen demand (SOD) or total organic carbon (TOC), and are nontoxic j at concentrations expected.32 . Due to (1) the extremely small volume of water in relation to the discharge i flow and the volume of the canal system, (2) the high purity of the waste stream, (3) the anticipated good mixing of the waste stream and the discharge flow, and (4) the nontoxic aspect of the ion exchange resin, no impac.t to organisms inhabiting or utilizing the canal system or surrounding water bodies due to the discharge of the backflush waste water is anticipated. l 4-18 l i -, ~. -., - w.,,,- -- n ~ m. .,.m.,-,,nn.nn,..,,.,,,. ,,,-,--,,.n,.., . nan.,.,,,.,,,.n..,.nn.r.---,w,-n.,-,..n.--,,.

It is concluded that changes in the operational characteristics of the station due to the steam generator repair will not have an adverse or detectable impact on species known to inhabit or utilize the Turkey Point cooling canal system. 4.3.3 Endangered or Threatened Species q In compliance with Section 7 of the 1978 Amendments to the Endangered Species Act (Act), the NRC requested information32 from the U.S. Fish and Wildlife Service (FWS) concerning those federally recognized threatened and endangered species, both listed and proposed to be listed, and on designated critical FWSresponded3,g4 habitat which mi ht be affected by the steam generator repair progra The 3 to this request informing the NRC that the station site is within the known range of several species and, under provisions of the Act, requiring the NRC to perform a biological assessment for the listed species. The results of this assessment have been transmitted to the FWS for review.as s The assessment concluded that no adverse impact to any threatened or endangered species known or suspected to inhabit or utilize the Turkey Point site will occur. Furthermore, it was concluded that no destruction or modification of designated critical habitat to the detriment of the American crocodile Crocodylus acutus will occur due to this action. 4.4 Environmental Imoact of Postulated Accidents As discussed in the staff Safety Evaluation,3 the design and plant operating parameters which are relevant to accident analyses will not change as a result of the steam generator repair effort. Therefore, the assessment of the environ-mental impact of postulated accidents presented in the FES9 will be unchanged and remain valid. The Safety Evaluation 3 also considers accidents which are unique to the repair effort. The accidents considered were: release of activity during decontami-nation, accidents during cutting operations, and lifting accidents inside and outside of containment both from the point of view of damage to the steam generator itself and to nearby tanks and structures. All of the accidents were evaluated with large factors of conservatism. The one with the largest calcu-lated dose was found to be release of activity following the drop of the st(1m generator outside of the containment building. "\\ About 33.5 curies of activity would be released if the welded end plate over the primary side of the steam generator failed during the drop, releasing half of the " loose" activity. The limiting potential receptor from the point of l view of both breathing rate and dose conversico factors is the teenager's lung. Average meteorological conditions were assumed, and the potential receptor was l assumed to be at the exclusion area boundary. The radiological consequences evaluated for these conditions is about 15 mrem. Several areas of conservatism are present in this evaluation: (1) the drop accident itself is unlikely, I (2) it is unlikely that the welded end plate could completely fail, (3) the amount of activity that is considered to be loose was conservatively estimated, and (4) the size of particles that could be knocked loose was assumed to be such that they all would be carried to the boundary in adverse conditions while a r?nge of particle sizes would actually be produced with different propensities j to remain airborne or to be resuspended. 4-19

Some amount of activity, primarily Co-69, would probably be carried to the site boundary and deposited as " contamination" in case of an actual accident. It is not possible to evaluate quantitatively the amount of activity that actually might be transported to the boundary, mainly because the effects of the factors of conservatism listed above cannot be evaluated. It is the staff judgment that the amount would be small. We conclude that the potential environmental impact +.o t.h public from accidents during the repair effort is well within the 10 CFR 20 timits for normal opera-tion and that, combined with the probability of occurrence of accidents, results in a risk to the public that is acceptably small. 9 4-20

5.0 IMPACTS OF ALTFRNATIVES The basic choices of future action regarding the tube degradation problem are \\ (1) repair of the degraded steam generators, (2) continuation of the present mode of operation, (with increasing costs in plant efficiency and occupational exposure), and (3) shutdown of the Turkey Point Units 3 and 4 and replacement by generating plants of different design. The option of operating the FPL system without Turkey Point 3 and 4 is not feasible in light of our review of the power demand in the FPL service area.s.12.is.ts FPL opted for repairing the degraded steam generators, with changes in design, materials, and operating procedures calculated to eliminate the tube denting problem. The units can be run in a derated mode and, based on economic considerations, no doubt would De operated in a derated mode in preference to shutting the units down with no replacement power. In the absence of methods to arrest or greatly reduce denting, the continua-tion of operation for an extended period in the present mode is impractical. With tube degradation and plugging continuing at the present rate, the units would of necessity likely be derated as discussed earlier in Section 4.2. FPL has estimated the cost of replacement power, based on fuel differential costs, to be about $535,000 per oay for the shutdown of a unit. Consequently, as. discussed in Section 4.2, the cost of derating the Turkey Point Units would be about $858,000 in ten years. Also, the person-rem cost of occupational exposure during the inspection and plugging of degraded tubes would continue. Laboratory test programs on the denting phenomenon are currently underway to define the corrosion process more precisely and to develop preventive measures such as corrosion inhibitors. While the combination of steam generator secon-dary side cleaning and corrosion inhibitors is being studied by some utilities to combat denting in its early stages, the denting phenomenon at Turkey Point is too advanced for such measures to be practical. Therefore, FPL cannot count on a greatly reduced future rate of tube degradation to justify con-tinuing the present mode of operation. The option of shutting down the Turkey Point Units and replacing them with units of different design is easily shown to be much more costly than that of repairing the steam generators. FPL estimates 1 (Section 7.7) that the capital cost of new nuclear units with improved steam generators would be about $2.0 billion and would require about 10 years to build. New fossil units l l \\ would cost about $1.5 billion in capital and require about 7 years to build. l The capital cost for gas turbine units would be about $310 million and would l require about 3 years to build. VEPC0 made similar comparisons for the steam generator repair program at the Surry Station and found that the cost com- ~ parison overwhelmingly favored the repair option. Based on our review of the above figures, we find that the time and cost estimates are reasonable. We conclude that the plant replacement option is not economically feasible. In addition, there would be significant environ-mental impacts frca such a large-scale construction cperation. The most i l practical overall optSn is therefore to repair the degraded steam generators l (Table 5.1). i 5-1

Table 5.1 CPTION CONSIDERED l. Repair of Generators Chosen alternative 2. Continued Operation With No Repair Impractical because of cost and ultimately FPL would face the same decision 3. Replacement by Plant of Another Not economically feasible Design In the remainder of this section, we shall consider the radiological and economic costs of several alternative ways of repairing and disposing of the degraded steam generators. An important item in estimating economic costs is the cost of replacement power during unit outage. The FPL cost estimate of $145,000,000 for the replacement power needed during the 270-day outage of each unit corresponds to a replacement power cost of about $535,000 per unit per day of outage. The replacement cost of $535,000 per day is based on the availability of fossil-fired fuel capacity which normally would be used only during periods of peak demand. The repair program was planned to be carri d e out during the seasonal periods of relatively low demand. However, if,uut-down is required during peak demand periods, c.- if long-term derating is necessary, new replacement capacity would have to be installed resulting in replacement power costs about 50% higher. 5.1 Decontamination FPL has estimated 1 (Section 8.2) that chemical decontamination of the steam generators before cutting would likely result in a net saving in occupational exposure compared to shielding. This would apply to the decontamination of the channel head or of the reactor coolant pipe and not to routine decontamina-tion outside tne reactor coolant pressure boundary. In a letter from R. E. Uhrig dated Stptember 24, 1980, additional information was provided regarding decontamination, but the specific method was not named. In a letter from R. E. Uhrig dated November 21, 1980, he indicated that FPL preferred a grit blast method of decontamination. The staff is familiar with the grit blast method because of its application to the Oconee 1 steam generator decontamination. Our evaluation 3 is given in the related Safety Evaluation 3 for this action. We conclude that the grit blast method of decontamination of the channel head is acceptable with certain conditions which we propose be made a part of the license amendment should this repair program be approved. 5.2 Retubing of Existing Steam Generators The retubing operation would involve (1) removing the upper or dome portion of the steam generator, (2) removing the lower assembly internals and tubes,- (3) replacing the latter with state-of-the-art internals and tubes, (4) refur-bishing the upper internals, and (5) welding the dome back in place. FPL has 5-2

estimated 1 (Section 7.3 and 7.4) that the cost o? this operation in both dollars and occupational exposure would be higher than the proposed replace-

nent of the complete lower assembly.

Further, it should be noted that shop fabrication of new lower assemblies, as contrasted to the in place fabricatian necessary for the retubing option, could provide more positive assurar'e that s the quality of the repaired generators was acceptable. On the other hand, we are aware of recent developments by Westinghou'e in the technology of in place refurbishment which shows some promise of red :ing unit outage and personnel exposure below the values for the FPL proposed repair method. A detailed proposal of the Westinghouse in place refurbishment is being reviewed.11 The review raised a number of questions which have been sent to Westinghouse. NRC is awaiting the response to the questions. If our assessment is favorable, in place retubing may be an alternative for steam generator repairs in the future. Estimates of the time required to wait for the NRC approval of retubing for the Turkey Point Plant indicate that it would likely take a minimum of two years for this approval to be granted. This includes time for the NRC staff to approve the Westinghouse Plan, time for FPL to adapt t'e Westinghouse plan to Turkey Point and to prepare a report for the NRC to review and approve, and time for the NRC review. It does not include time for any additional technical problems to be solved. The economic cost of derating was discussed in Section 4.2. In the time frame contemplated for the proposed licensing action, this is not considered to be an available alternative to the proposed action. Contrib-uting to this judgment are the following facts: (1) the NRC approval of the retubing technique is not assured, (2) ability to reuse the tube sheet is not assured, (3) continued operation with the present steam generaters until the retubing process is approved would extend the period of higher industrial exposure rate, (4) Unit 4 and Unit 3 both would likely be operating in a derated mode before the retubing process is implemented, and (5) a change to the retubing process would change only a fraction of the proposed repair program, probably less than one half.

5. 3 Installing Tube Sleeves in the Existing Steam Generators We have examined the tube sleeve alternative and find tube sleeving is not a feasible or practical repair option for the steam generator problem at Turkey l

Point. Our reasons can be bri stated as follows: (1) The tube sleeving l option, where it has been appl is been primarily intended for application to tubes containing through-wall Jefects (wastage, cracking, etc.) rather than significant dents; (2) The sleeves would be subject to the denting phenomena, i.e., ' e sleeving program does not assure that the generators will last the life of the plant; (3) Installation of sleeves in dented tubes would lead to other problems unique to plants with dented tubes such as (a) excessive strain at the upper and lawer ends of the sleeves due to expansion, (b) multiple l sleeves would likely be required in many tubes; (c) tubes might not be properly inspectable betweea sleeves, etc.; (4) A large-scale sleeving program does not l assure a significant saving of occupational dose compared to that experienced during the Surry repair; and (5) In order to accommodcte the denting in the Turkey Point Plant, the flow area in the repaired steam generator tubes would be reduced by as much as 65% relative to a new tube. l 5-3 1 l

5.4 Replacement of the Entire Steam Generator FPL considered this alternative in two ways. Based on FPL analysis, which we have reviewed and found reasonable, a construction opening in the containment wall about 20 feet wide and 40 feet high would be required since the upper ~ assembly of the steam generator could not pass through the existing equipment hatch. An alternative plan also considered was removal of the steam generator through a 20-foot diameter hole in the containment dome. Tha personnel exposure for these alternatives would be about the same as for the proposed repair, because essentially the same high-dose operations will be required in each case. Elimination of the cut across the diameter at the transition core of each steam generator results in only a small saving of radiation exposure. The capital costs are estimated to be about 15% higher. The principal cost difference is due to an estimated additional outage of about 100 days per unit for the alternative. This corresponds to an additional requirement of about $110,000,000 worth of replacement power during the repair of both units calcu-lated at the rate of about $535,000 pet day of outage per unit. In summary, this alternative would have essentially the same environmental inpact as the FPL proposal (primarily occupational dose) and greater economic cost. Also, there would be significant structural rcpairs involved to assure that the containment is returned to the original state after completion of I this repair program. For these reasons, we conclude that the FPL proposed recair method is preferable (Table 5.2). Table 5.2 STEAM GENERATOR REPAIR ALTERNATIVES 1. Retubing (5.2) Not NRC approved method. Dose reduction not assured 2. Tube Sleeving (5.3) Not appropriate for dented tubes 3. Replacement of Entire Generator Personnel exposure about the same (5.4) Hole in containment presents significant structural problem 4. Replacement of Lower Assembly Proposed method 5.5 Alternate Disposal Methods The stes: generator icwer assemblies will comprise the largest source of radioactive waste requiring disposal. Several alternatives for the disposal of the lower assemblies were considered: (1) Immediate intact shipment to a licensed burial facility; 5-4 i .n..

(2) Delayed intact shipmant to a licensed burial facility; (3) Immediate cut-up and shipment to a licensed burial facility; (4) Delayed cut-up and shipment to a licensed burial facility; n' (5) Onsite storage unit facility decommissioning. Beacuse of the size and packar'r1 involved, the only practical method for immediately shipping the asserNies intact would be by barge. FPL considers this a viable option and is actively pursuing the resolution of the problems involved such as barge offloading, available disposal space, etc. We have evaluated this option (Appendix A) and find that if FPL follows applicable laws regarding such shipments, this option will provide resonable assurance that the health safety of the public will not be endangered and is acceptable. Immediate cut-up and shipment is possible with transoortation by truck or rail. The assemblies could be cut into suitable sized segments and packaged and transported. Cutting of the assemblies and subsequent handling would result in increased occupational exposures due to the activity on the surfaces exposed to reactor coolant Some dose reduction could be achieved by decontamina-tion of the reactor coolant surfaces. However, effective decontamination factors may not be achievable due to presence of a significant number of plugged tubes which would prevent decontamination chemicals from entering 20% or more of the tubes. Reduced exposures due to decontamination would be accompanied by a significant increase in decontamination solution liquid radioactive wastes. These wastes would have to be processed and solidified. We conclude that immediate cut-up and offsite shipment will cause an ennecessary person-rem burden on the workers without providing a significant operation benefit to FPL and to the public as compared to onsite storage as discussed below. FPL has proposed that the immediate intact offsite shipment is a preferred method of disposal; however, plans also include long-term onsite storage which would allow for decay of radioactivity to relatively low levels to minimize radiation exposures before processing for shipment. The lower assemblies l we ;1d be stored in an engineered storage facility specifically constructed for j this purpose. Such storage would provide for FPL responsibi'.ity and control l of access and exposure to the assemblies until the radiation has decayed to l levels that will allow easy disposal (e.g., Unit decommissioning). Based on I decay of the expected radioactive corrosion products, it is estimated that storage for 30 years can reduce the radiation levels to less than 1% of those will ce sealed with steel plates or plugs prior to removal from containment to expected when the assemblies are removed from containment. The assemblies l I eliminate airborne particulates from being released f om internal surfaces. Internal decontamination will not be necessary because of the seals. Some l surface contamination will be present on the outside of the assemblies. FPL has stated that the external surfaces will be decontaminated such that removable l contamination levels will be less than 2200 dpm/100 square cm prior to removal from containment. We will require the welds to be coated to seal out moisture. Therefore, any release to the environment from transport of the assemblies to the onsite storage facility should be negligible. 4 5-5

We have reviewed the FPL storage building. Because the external contamination levels will be <2200 dpm/100 square cm airborne releases from the external surfaces of the generators are not expected. FPL has proposed quarterly surveillance of the facility consisting of visual inspections and randem swipes of the generators and area radiation surveys to assure that no airborne contaminants are being released from the facility. There will be a limited amount of direct radiation which penetrates the storage building walls. Based on the maximum expected radioactive inventory of the steam generators and the shielding of the storage facility, FPL has estimated, using commonly accepted practices, an annual dose of less than one mrem to ma individual at the site boundary. We have reviewed the bases for this estimate and consider the bases acceptable. We conclude that the expected radiation levels on contact with the outside of the facility walls are approximately the levels for unrestricted areas specified in 10 CFR Section 20.105. If upon completion of the storage phase FPL finds levels in excess of 10 CFR Section 20.105, FPL will be required to provide adequate control and posting pursuant to 10 CFR Section 20 203. We have reviewed the FPL proposed surveillance program for the storage facility and find it acceptable. Our review of the five disposal options available is summarized in Table 5.3 and the FPL estimates are summarized in Table 5.4. Based on this summary, we have concluded that immediate offsite shipment would be an acceptable method of disposal provided the necessary State approvals a.e obtained. We have reviewed the FPL proposed method of storage, should the offsite shipment option not be available, and conclude that there is reasonable assurance that this storage will not endanger the health and safety of the public and is acceptable. In addition, we conclude that the measures to be taken to control and monitor thic storage will keep occuoational exposures and radioactive effluents as low as reasonably achievable. e 5-6

Table 5.3 STEAM GENERATOR DISPOSAL ALTEP. NATIVES 2 Approximate Approximate Person-Rem per Airborne Release, Option Steam Generator Ci per Generator c b Immediate intact shipment

2. 4 Negligible 8

b Long-term storage (including 10 Negligible surveillance) with intact shipment a Long-term storage with cut-up 16 0.005 and shipment Short-term storage with cut-up at 5 yr 230 0.026 at 15 yr 60 0.015 Immediate cut-up and shipment by 580 0.042 rail / truck - no decontamination Immediate cut-up and shipment by rail / truck 270 0.010 - with chemical decontaminatioa a30 to 40 years. Since the steam generator will be sealed before it is removed from containment, no release of radioactive material is expected during the shipment or storage. cEstimates for short-term storage followed by intact shipment would be only slightly larger than this, perhaps 5 person-rem. \\l l l s 5-7

Table 5.4 COST OF ALTERNATE DISPOSAL METHODS (FPL)2 Exposure,a Method Cost, dollars Person-rem a. Cut-up and disposal near term $5,320,000 1500-3050 with no decontamination b. Cut-up and disposal near term $4,930,000 800-1650 with solidification agent c. Cut-up and disposal near term $5,540,000 250-1150 with decontamination d. Long-term storage with deferred $3,470,000 70-90 cut up and disposal e. Near-term barge shipment $2,620,000 53-56 f. Long-term storage with disposi-tion during decommissioning $3,000,000 51-53 aNote that these doses are for six lower assemblies. The estimates in Table 5.3 are for one lower assembly. 5-8

6.0 CONCLUSION

S We have reviewed the proposed steam generator repair action and have reached the following conclusions. (1) The proposed replacement of the lower assemblies of the steam generators is the best available option, from both the radiological and economic ~ standpoints, for eliminating the tube degradation problem. (2) We have reviewed the dose reduction measures to be used by the licensee and conclude that the doses would be ALARA. We have also considered the health effects resulting from such exposure and concluded that these are not significant. (3) The new steam generator design incorporates features which will eliminate the potential for the various forms of tube degradation observed to date. (4) The restoration would restore the generators to the condition evaluated in the FES.9 (5) Offsite doses resulting from the steam generator repair will be less than those from recent plant operations, comparable to doses presented in the FES,8 and small compared to the annual doses from natural background radiation. Therefore, the offsite doses will not be significant. On the basis of the foregoing analysis, the staff concludes that the proposed steam generator repair will not significantly affect the quality of the human environment. Even if it had been concluded that the absolute occupational dose of the proposed repair program would be of significant impact to the human environment, this impact is outweighed by the decrease in the long-term radiclogical exposure compared to what would be incurred if the facility were to operate without the proposed repairs and by the economic advantage of the proposed repair program, so we would nonetheless conclude that the prcposed repair program should be implemented. '\\ l 1 t 6-1 i t

7. 0 REFERENCES 1.

" Steam Generator Repair Report - Turkey PMnt Units 3 and 4," Florida Power and Light Co., September 20, 1977 a..a Revisions 1 through 7, dated 3 December 20,1977, March 7, April 25, June 20, August 4,1978, January 26, 1979 and March 28, 1980, respectively. 2. " Radiological Assessment of Steam Generator Removal and Replacement," G. R. Hoenes, D. A. Waite, and W. D. McCormack, Pacific Northwest Laboratories, NUREG/CR-0199, September 1978. Revised as NUREG/CR-1595, October 1980. 3. Safety Evaluation by the Office of Nuclear Reactor Regulation, License Nos. OPR-31 and DPR-41, Florida Power and Light Co., Turkey Point Plant Units 3 and 4, Docket Nos. 50-250 and 50-251, U.S. Nuclear Regulatory Commission, May 14, 1979, updated December 17,1980 (NUREG-0756). 4. Steam Generator Repair Program, Surry Power Station Unit Nos.1 and 2, Virginia Electric and Power Co., August 17, 1977 and revisions dated December 2,1977, April 21, June 2, June 13, June 30, September 1, October 25, and November 10, 1978. 5. Safety Evaluation Report by the Office of Nuclear Reactor Reculation, License Nos. OPR-32 and DPR-37, Virginia Electric and Power Cempany, Surry Fower Station Units 1 and 2, Docket Nos. 50-280 and 50-281, U.S. Nuclear Regulation Commission, December 15, 1978. 6. Environmental Impact Appraisal by the Office of Nuclear Reactor Regulation, License Nos. OPR-32 and DPR-37, Virginia Electric and Power Co., Surry Power Stations, Units 1 and 2, oc:.~ + Nos. 50-280, 50-281, U.S. Nuclear Regulatory Commission, January 20, _979. A Draft Environmental Statment (NUREG-0663) was issued on March 20, 1980 and a Final Environmental Statement (NUREG-0692) was issued in July 1980 (see also reference 24). 7. "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be As low As is Reasonably Achievable," U.S. Nuclear Regulatory Commission, Regulatory Guide 8.8, Revision 3, June 1978. 3 8. " Radioactive Material Released from Nuclear Power Plants (1976)," T. R. Decker, U.S. Nuclear Regulatory Commission, NUREG-0367, March 1978 9. Final Environmental Statement related to the Operation of te Turkey Point Plant, U.S. Atomic Energy Commission, July 1972. 10. " Radioactive Material Released from Nuclear Power Plants (1977)," T. R. Decker, U.S. Nuclear Regulatory Commission, NUREG-0521, January 1979. 11. " Steam Generator Retubing and Refurbishment," Westinghouse Electric Corp., WCAP-9398, January 1979. 7-1 l

12. " Final Environmental Statement Related to the Construction of St. Lucie Plant Unit 2," Section 8.1, U.S. Atomic Energy Commission, May 1974. 13. Letter from R. E. Uhrig, FFL, to V. Stello, NRC, dated July 20, 1977, transmitting " Report on Systems Disturbance, May 16, 1977," Florida Power and Light Co., June 29, 1977. 14. Letter from S. Varga, NRC, to R. E. Uhrig, FPL, dated June 12, 1980 transmitting Amendments 58 and 51 to the Turkey Point Licenses permitting continued operation of Unit 4 for six months and operation of Units 3 and 4 with 25% of the steam generator tubes plugged. 15. Letter from S. Varga, NRC, to R. E. Uhrig, FPL, dated October 30, 1980 transmitting Amendment No. 60 to the Turkey Point Unit 3 license per-mitting six months of operation. 16. Letter from R. E. Uhrig, FPL, to D. G. Eisenhut, NRC, dated June 30, 1980. 17. " Occupational Radiation Exposure at Light Water Cooled Power Reactors, 1978," B. G. Broahr, U.S. Nuclear Regulatory Commission, NUREG-0594, September 1979. 18. "The Effect on Populations of Exposure to Low Levels of Ionizing Radiation," Report of the Advisory Committee on the Biological Effects of Ionizing Radiation, National Academy of Sciences, 1972. 19. " Reactor Safety Study, an Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," U.S. Nuclear Regulatory Commission, WASH-1400, NUREG-75/014, 1975. 20. " National Radiation Exposure in the United States," D. T. Oakley, U.S. Environment:J Protection Agency, ORP/SID 72-1, June 1972. 21. " Radioactive Materials Released from Nuclear Power Plants (1975)," U.S. Nuclecr Regulatory Commission, NUREG-0218, March 1977. 22. Letter from R. E. Uhrig to S. Varga, dated November 4, 1970. 23. " Final Environmental Statement Related to Steam Generator Repair of Surry Power Station, Unit No.1, Virginia Electric and Power Company," U.S. Nuclear Regulatory Commission, NUREG-0692, July 1980 (see also ref 6). 24. Busnel, Rene-Guy, 1978. Introduction, p. 7-22, J. L. Fletcher and R. G. Busnel [ed.], Effects of Noise on Wildlife, Academic Press, New York. 25. Uhrig, Robert E., 1980, Letter to Steven A. Varga, U.S. Nuclear Regulatory Commission, Washington, D.C., from Robert E. Uhrig, Florida Power and Light Co., Miami, FL, dated September 24, 1980. 26. Uhrig, Robert E., 1980 Letter to Steven A. Varga, U.S. Nuclear Regulatory Commission, Washington, D.C., from Robert E. Uhrig, Florida Power and Light Co., Miami, FL, dated July 22, 1980. 7-2

27. Uhrig, Robert E., 1980, Letter to Steven A. Varga, U.S. Nuclear Regulatory Commission, Washington, D.C., from Robert E. Uhrig Florida Power and Light Co., Miami, FL, dated November 4,1980. 28. Uhrig, Robert E., 1980, Letter to Albert Schwencer, U.S. Nuclear Regulatory S Commission, Washington, D.C., from Robert E. Uhrig, Florida Power and Light Co., Miami, FL, dated March 18, 1980.

29. McKee, J. E. aid H. W. Wolf.

1963. Water Quality Criteria, 2nd Ed. o Publication No. 3-A. The Resources Agency of California, State Water Resources Control Board. 30. United States Environmental Protection Agency, 1976. Quality criteria for water. EPA-440/9-76-073. U.S. Environmental Protection Agency, Washington, D.C. N 31. Kunin, R., 1979. The Disposal of Spent Powdex ion exchange materials. In: Letter to A. Schwencer, USNRC, from R. E. Uhrig, FPL, dated March 18, 1980. FPL Co., Miami, FL. 32. Letter S. Varga, NRC, to K. Black, FWS, June 19, 1980 re endangered species at the Turkey Point Plant. 33. Letter D. J. Hankla, FWS, to S. Varga, NRC, July 14, 1980, regarding assessment of endangered species in the area. 34. Letter J. C. Oberhan, FWS, to S. Varga, NRC, July 25, 1980, supplementing the July 14, 1980 letter. 35. Letter S. Varga, NRC, to D. J. Hankle, FWS, December 12, 1980, trans-mitting the Assessment of the Impacts of the Steam Generator Repair Program at the Turkey Point Stations on Threatened or Endangered Species. 36. " Occupational Risks of Ontario Hydro's Antomic Radiation Workers in Perspective," R. Wilson and E. S. Koehl, Health and Safety Division of Ontario Hydro, Presented at Nuclear Radiation Risks. Vol. 1, Occupational Radiation Standards, A Utility Medical Dialogue, International Institute of Safety and Health, September 22-23, 1980. l 7-3

8.0 FEDERAL, STATE, AND LOCAL AGENCIES TO WHOM THIS ENVIRONMENTAL STATEMENT WAS SENT This Draft Environmental Statement was sent to the following: Advisory Council on Historic Preservation Department of Agriculture Department of the Army, Corps of Engineers Department of Commerce Department of Energy Department of Health, Education and Welfare Department of Housing and Urban Development Department of the Interior Department of Transportation Environmental Protection Agency State of Florida Dade County In addition, a copy was sent to Mark P. Oncavage. 'l 8-1

APPENDIX A g ENVIRONMENTAL EVALUATION OF OFFSITE SHIPMENT OF THE TURKEY POINT STEAM GENERATOR LOWER ASSEMBLIES

1. 0 Introduction Six steam generator lower assemblies (SGLAs), three from Unit 3 and three from Unit 4, at the Turkey Point Nuclear Power Station are to be changed.

One of several alternatives discussed by the applicant 1 for the disposal of the SGLAs is to ship each SGLA to the nuclear waste burial ground at Barnwell, S.C. The following evaluation presents a general description of the plans and estimated environmental effects from this alternative. Further relevant details are contained in the Steam Generator Rept.ir Report.1 The steam generators used at Turkey Point have a common design called a vertical U-tube heat exchanger. Heated primary water from the reactor enters near the bottom, passes through about three thousand U-shaped tubes, first in the upward direction and then downward, finally leaving near the same elevation at which it entered. U-shaped tubes and their supports fill the shell of the lower assembly. Heat flows through the tube walls from the primary water to secondary water introduced outside the tubes within the lower assembly. The secondary water boils; the resultant steam flows upward into an assembly of moisture separation equipment to remove the small quantities of liquid water droplets invariably present in steam. The steam is ultimately used to drive a turbine to produce electricity. The Turkey Point SGLAs are snaped like a cylinder topped by a truncated cone. They stand 34 feet tall. The bottom section is 10.6 feet in diameter and 25.5 feet long, and the truncated cone is 10.6 feet in diameter at bottom, 13.1 feet in diameter at top, and 8.5 feet long. The Turkey Point SGLAs have steel shells two to three inches thick. Each SGLA weighs-about 173 tons. l Each SGLA will contain at the time of disposal about 250 curies of radioactive i metallic corrosion products deposited as a tightly bound layer on the inside tube surfaces where the primary water flowed through the tubes. 2.0 General Description of Offsite Shipment Before each SGLA is removed from its containment building and shipped offsite, l the SGLA will be completely sealed by using two-or three-inch thick steel plates and inserts and 1-1/2-inch fillet welds to prevent release of radio-active material within its packaging during shipment. 1 1 l l 1" Steam Generator Repair Report, Rev. 7, U.S. Nuclear Regulatory Commission j Docket Nos. 50-250, 50-251 (March 1980). l l A-1 l l i- -

f After removal from the conttinment buiiding, the first SGLA will be lifted by a crane mounted on a gantry t?w?r and laid horizontally in the lower half of a steel shipping cask attached to a special transport vehicle. The SGLA will then be secured and the upper half of the shipping cash will be bolted to the lower half, completely encasing the SGLA. The steel shipping cask will be about 42 feet long and 20 feet in diameter. Its cylindrical steel wall will be 2-1/2 inches thick and the steel ends will be 2-3/4 inches thick. Its tare weight will be 132 tons. The transport vehicle will then be moved to the barge slip at Turkey Point, driven onto a barge, snd secured. The barge will proceed through Biscayne Bay to the Atlantic Ocean, along the coastline, and up the Savannah River, a distance of 600 miles. The transport vehicle will be taken off the barge at Johnson's Landing, S.C., and driven by highway to the burial ground, a distance of 40 miles. At the burial ground, the upper half of the shipping cask will be removed; the SGLA will be lifted out of the cask and placed in a burial trench. The trans-port vehicle and steel shipping cask will then be returned to Turkey Point for shipment of another SGLA. While the first SGLA is being transported to the burial site, the second and third SGLAs will be removed from the containment building and temporarily stored on the laydown area to await transportation. Although the SGLAs are operated in a vertical position, they will rest in a horizontal position in this laydown area of the plant site. Because the SGLAs are only mildly radioactive (<100 mc/hr at the surface) and because the laydown area is remote from both the puclic and plant operations, the SGLAs are assumed not to be shielded while in the laydown area. The second SGLA will.1st at the laydown area for approximately three weeks before loading, and the third SGLA will rest approximately six weeks before loading. The process described above will be repeated until each of the six SGLAs reaches its destination at the buria; site.

3. 0 Environmental Effects of Offsite Shipment 3.1 Nonradiological Effects Shipping the SGLAs offsite may produce two principal nonradiological effects:

use of resources and undesirable consequences. Resources to be used include fuel, transportation vehicles, packaging materials, such as for the shipping cask, and burial land volume. Undesirable consequences include noise, pollution, accidental injuries and deaths, property damage, such as possible breakup of the highway from the heavy weight of the shipment, and denial of some resources, such as road space at the times the shipping cask is transported from barge landing to burial ground. The quantities of fuel required to move the SGLAs from the Turkey Point site to the Barnwell burial ground are negligibly small compared to the quantities of fuel used in daily commerce and thus may be discounted as an environmental effect. The transportation vehicles used to ship the SGLAs can be used again for hauling other heavy loads, so this resource use does not represent an A-2

environmental effect. The shipping cask for the SGLAs is a special package because of its size, about 20 feet in diameter and about 42 feet long, but the materials in its construction do not represent a significant fraction of the ' S materials used in daily commerce and the cask can be used again for shipment of other SGLAs. Hence the use of resources for the packaging may be discounted as an environmental effect. Burial land volume must be committed for the SGLAs. The upper limit for the volume of waste admitted to the Barnwell burial ground is being decreased from a present value of 144,000 ft per month 3 to a value after September 1981 of 100,000 fta per month. The total volume of 3 the six SGLAs is about 20,000 ft, so use of this resource represents about one week of burial capacity after September 1981. Noise and pollution that may be produced by the sh.pments of SGLAs are small compared to the noise and pollution produced by shipments in daily commerce. Consequently, these items are not counted as significant environmental effects. Transporting a large heavy object like an SGLA or the empty cask will take time and will disrupt normal traffic. The risk of an accident for the truck portion of the trip thus will be smaller than that for normal two way traffic at highway speeds. The risk of accidental death and injury is thus overestimated by the calculation for normal traffic. This calculation is discussed as follows. The probability of truck accidents in the United States was estimated for 1969 as 1.7 per million truck-miles.2 Abcut one third of the accidents for that year resulted in injury, but only about three percent resulted in death. Since 1969, the fatality rate on United States highways has decreased from 4.7 per vehicle mile in 1970 to 3.2 per vehicle mile in 1977.3 The death rate dropped markedly in 1974, probably because states imposed a 55 mile per hour speed limit then. Assuming the fatality rate decreases because of a similar decrease in accident rate, we infer that the current accident rate is about 1.1 accidents per million vehicle miles. We will use this number as a bound for estimating risks. Transporting all six SGLAs offsite to the Barnwell burial ground and transporting the empty (except for possible interior contamination) cask back again will require about 480 truck miles and 7200 barge miles of movement. The expected number of truck accidents (normal traffic) for this much movement 3 is 5 x 10 4 The accident rate for barge transportation is estimated to be 10 accidents per million barge-miles.4 The expected number of barge accidents is then 7 x 10 2 If many similar shipping campaigns were conducted, we would expect a barge accident every 140 campaigns and one truck accident every "" Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants," WASH-1238 (December 1972), Appendix C. 3 Data supplied by National Highway Traffic Safety Administration. 4" Final Environmental Statement on the Transportation of Radioactive Material by Air and Other Modes, "NUREG-0170 (December 1977), Table 5-7. A-3

i t 200 campaigns for normal traffic conditions. For the actual concitions of slow movement and disrupted traffic, we would expect one truck accident in a more than 200 campaigns. We conclude that a transportation accident involving the shipping cask carrying one of the SGLAs is not likely to happen. Injury or death from such an accident is even less likely. t State highway weight restrictions limit the gross weight of trucks for routine shipments so that the gross weight of the cargo is limited to about 25 tons. Shipments of cargoes weighing up to about 35 tons may be allowed in most states under a special overweight permit. The states often prescribe special routing for overweignt shipments and in some cases restrict the period during wnich 7 trucks can travel. Repetitive shipments of overweight loads may cause breakup i of the roadway. For shipment of each OGLA, the gross weight is 305 tons. While the SGLA is being moved over land, the truck will require use of the wnole hignway. A single lane of highway customarily accommodates 8 feet wide vehicles, but the SGLA shipping cask will be on the orcer of 20 feet wide. Consequently, traffic will have to be stopped or diverted onto another highway wnile the SGLA is being moved. The shipper or carrier will have to get specific authority from the state to make the shipment. 3.2 Radiological Effects of Routine Shioment Before removal from the containment building, the exterior surfaces of the SGLAs will be decontaminated as necessary to reduce the level of removable contamination to a value below that considered significant in U.S. Department of Transportation regulations (49 CFR 173.397). The NRC requires its licensees to observe these regulations (10 CFR 71.5). For example, beta or gamma radiation from radionuclides on the surfaces shall not exceed 220 disintegrations per minute per cm2 of surface area. To determine this extent of decontamination, the licensee will, following the regulation, measure the radioactivity on absorbent material wiped over about 300 cm2 of surface. If the measured activity per square cm coes not exceed ten percent of 220 dis / min /cm, then he 2 may conclude that any contamination residing on the surface is not significant. i As mentioned earlier, each SGLA contains about 250-Ci of radionuclides. This radioactivity represents a few grams of material plated more or less evenly over the miles of tubing. Because the SGLAs are sealed, however, none of this material can be released from an SGLA during routine transportation. Radiological effects during routine transportation may still arise, though, because of radiation emitted from the materials inside the SGLA. Dose rates outside the cask from this activity will be limited by regulation to less than 10 mrem /hr at six feet from the cask surface. The applicant declares that radiation rirveys will be taken periodically before, during, and after transportation to confirm that radiation levels are within this limit,s although only surveys befnre and after transportation are required by the

• Letter from R. E. Uhrig, Florida Power & Light Co. to S. A. Varga, U.S. Nuclear Regulatory Commission responding to requested information (October 31, 1980),

Regulatory Information Distribution System, Accession Number 3011050436, p. 4. A-4 .._y. -r-4

regulations. If the surveys show excessive radiation, the regulations require that they be reported and that the licensee follow his previously developed g procedures in remedying the situation (10 CFR 20.205). Other limits will also apply (49 CFR 173.393(1) and (j)). The radiological effect on the public from shipment is estimated by the calculated population dose. The population may absorb dose from several operations: temporary onsite storage, loading, offloading, and transportation of all six SGLAs. Let us consider the dose to the public from each of these operations. Temporary onsite storage will cause only a small population dose, if any, since the public is removed from the site. The applicant asserts that the measured values of the radiation field at various points on the surface of an SGLA are on the order of 100 mr/hr.6 If the source distribution causing this field is modeled as a line source, the field would be expected to vary inversely with the distance from the source.7 Thus at 127 inches from the sources (the outside diameter of the SGLA) the field is taken on the order of 100 mr/hr. At six feet from the surface, a presumed distance for a person standing by the SGLA in the laydown area, the field would be (127 in) + (127 in + 72 in) = 0.64 as great, or 64 mr/hr. At larger distances than about 50 feet, the SGLA is modeled as a point source, for which the field varies inversely vith the square of the distance from the source. The laydown area is located on the south edge of the Turkey Point station site between Loch Rosetta and Lake Warren.8 A research activity and a military a tivity are set up at a distance of about one thousand feet from the laydown area. The dose rate at this distance from an SGLA temporarily stored on the laydown area is estimated to be 1 x 10 2 mr/hr. The population dose to 100 people involved in these activities would be about 0.7 person-rem, assuming each l person worked 8 hours a day, 5 days a week, for the three weeks of possible exposure to the second SGLAs from each unit and for the six weeks of possible exposure to the third SGLAs from each unit. One would not expect many other people who did not have business at the site I to ordinarily be this close to the SGLAs. At half a mile, an offsite member of the public might be exposed to a field of about 2x10 3 mr/hr. Assuming for the sake of argument that ten people were at that distance for as much as one day (for example, they may be fishing or camping) during the six week period 3 estimated for an SGLA to be on the laydown area, the population dose would be about 5 x 10 4 person-rem. This estimate assumes the SGLA on the laydown area is not shielded. This dose calculation represents an upper limit since the measured valuess of the radiation field at the surface of an SGLA do not exceed the 100 mr/hr used. This population dose estimate could be doubled, however, since during the first-three weeks of the six week period, two SGLAs are expected to be on the laydown area. j 60p cit, Steam Generator Repair Report, Fig. 3.3-7. 7 Ibid., Table 2.1-1. 8 Ibid., Fig. 3.3-7. A-5

I The summed population dose to the research, military, and fishing groups from onsite storage is thus estimated to be about 0.7 person-rem. The dose to the public from loading, offloading, and transportation will be affected by the shielding afforded by the shipping cask. The cask wall represents a 2-1/2-inch thick carbon steel shield, which reduces the dose rate just inside the cask by a factor of 5 just outside the cask. Thus one would expect no more than 20 mr/hr at the outer cask surface and about 60 percent of that dose rate, or 12 mr/hr, at six feet from the cask. ' Extra shielding may be required in places to limit the field at six feet to the regulatory value of 10 mr/hr. If a person were to stand in such a field for an hour, he would receive a dose of ten millirem, or about ten percent of his annual background dose. Normally only persons working on the shipment would be so exposed. The loading and offloading operations will take place onsite away from the public, and will radiologically affect the public in the same way as onsite storage of an SGLA on the laydown area, except that for at least part of the operations, the public will be shielded from the SGLA by the cask. Assuming the loading and unloading operations each take a day, that 10 persons are exposed offsite a half mile away for the day, and that the SGLA is within its cask for half the day, the population dose for all six SGLAs is about 1 x 10 4 person-rem. The SGLAs are to be transported by barge and heavy duty truck in remote areas. The barge trip will consist of a segment at sea and a segment on a river. No population exposure to the public will be expected from the sea portion of the trip because the barge will be remote from the public. On the river segment, however, people who live along the river could reasonably be expected to be exposed. The situation is similar to that of a wide highway. This part of the population dose will be considered later on. On land, the closest approach of members of the public to the cask would probably not be less than 30 feet. For example, if the truck loaded with the SGLA were to pass by a store or restaurant, which could conceivably be located only 30 feet from the highway, then merchants and patrons could be exposed at this distance. The truck will be traveling very slowly, so an individual at this distance (30 feet) would be exposed the same as if the truck were stopped. In either case, an individual at this distance would be exposed to a dose rate of about 4 mr/hr. If 100 patrons and employees were so exposed for an hour, the population exposure for this situation would be 0.4 person-rem. The applicant plans to use a highway from Johnson's Landing, South Carolina, to a burial facility near Barnwell, South Carolina, a distance of about 40 miles. The highway traverses a rural area for which we estimate the population density to be no greater than 250 people /mi2 The total number of people living within a half mile on either side of the highway should be no larger than 10,000. The population dose to them from one SGLA passing by at one mile oer A-6

~ hour is estimated to be 3 x 10 4 person-rem and frgm all six SGLAs is 2 x 10 3 person-rem.9 This dose represents 2 x 10 4 percent of the annual background dose to these people. The river portion of the barge trip will probably cover about twenty miles. The river is modeled as a broad highway such that the ratio of the dose to the public, compared for rivers to highways, for minimum distance to the shipment centerline, is equivalent to the similar ratio comparing highways to city streets.10 This ratio is found to be 0.25. For the same population density as for the highway portion of the trip, the_ population dosg for the river portion of the trip is (0.5) (0.25) (2 x 10 3 person-rem) s 2 x 10 4 person-rem. In an environmental statement on transportation of radioactive materials, radiological effects from traffic jams, following traffic, and oncoming traffic are normally considered. Since the SGLAs will be traveling slowly and normal traffic will be disrupted, however, such events will not happen. In summary, the routine population exposure from offsite shipment of all six SGLAs is tabulated as follows. Approximate Population Dose Operation (person-rem) Onsite Storage

0. 7

~ Loading, offloading 1 x 10 4 ~ Transportation by barge 2 x 10 4 ~ Transportation by truck 2 x 10 3 Stop by store or restaurant 0.4 (perhaps a one-time event) Occupational exposure associated with effsite shipment of all six SGLAs has been evaluated as about 30 person-rem. " This population dose is roughly equivalent to the estimated occupational exposure which would be incurred from the alternative of storing all six SGLAs onsite and considerably less than that fcr other alternative ways to manage the SGLAs. 3 i '36ased on Eq. 0-7 in " Final Environmental Statemeat on the Transportation of Radioactive Material by Air and Other Modes, "NUREG-0170 (December 1977). In this equation, f

• f

= 0, f = 1 because of the rural environment. s u r The dose rate factor K for the SGLA is evaluated from K % 100 mr/hr - (127 in)2 2 . I ft /144 in2 = 88 mrem /hr ft2 10 Ibid., p. 0-5. 'Op cit, Steam Generator Repair Report, p. 3-20. A-7

Exposures to individual workers will be maintained within the limits specified in 10 CFR Part 20. Additionally, a training program will be conducted for all personnel participating in shipping activities whose duties will require them to enter radiation areas or to handle radioactive material. This training program, as well as other administrative controls, will help ensure that occupational exposures will be maintained as low as reasonably achievable. 3.3 _ Radiological Effects from Postulated Accidents 3.3.1 Introduction Shipment of new steam generators and other heavy equipment by barge is not a novel or unusual task. Steam generators have previously been shipped by barge, including six original and six replacement steam generators for both Turkey Point and Virginia Electric and Power Company's Surry Plant. Addition-ally, one contaminated steam generator lower assembly has been shipped by barge from Surry, Virginia, through the Panama Canal, to Hanford, Washington. Experience and accident statistics indicate that the occurrence of an accident during shipment is extremely unlikely. 3.3.2 Prevention and Mitigation of Accidents Several procedures will be used to prevent accidents.12 The gantry and cranes will be inspected and tested before the lower assemblies are loaded or offloaded. The roads along the truck route have been selected to minimize grades and sharp curves. Rural roads will be used. Roads will be inspected before shipment to ensure their ability to withstand the expected load. The speed of the truck will be controlled to reduce the probability of an accident.

Finally, an escort to control traffic will be provided for the truck.

If the shipping cask and the SGLA sink in the Savannah River or Biscayne Bay, they will be recovered. Such a recovery could be easily accomplished given the shallow depth of these bodies of water. 3.3.3 Accident during Temporary Onsite Storage The second and third SGLAs, which will be removed from eact, unit, will be temporarily stored on the Turkey Point laydown area prior to transportation to the burial site. The laydown area is not frequented by vehicles or within i falling distances of nearby structures or equipment. Additionally, this storage will not occur during the hurricane season. Thus, no accident scenario that could breach the lower assemblies during temporary onsite storage is considered credible. I 120p cit, Uhrig letter, p. 8. A-8 l

3.3.4 Accidental Drop of an SGLA During Loading or Offloading S During loading and offloading of the shipping cask, it will be necessary to lift the lower assemblies a maximum of twelve feet off the ground. The, probability of dropping an SGLA during such a lift is of the order of 10 8 23 The consequences of a drop would be inconsequential because the SGLA will be blocked and cribbed; in the event of jack failure, the SGLA would drop only a few inches onto the blocking. A possible drop could occur when the SGLA is lifted for placement on the shipping cradles. The SGLA will be lifted a few feet for mounting on a special truck trailer. If the SGLA is accidentally dropped, the shell would fail. However, a shell failure would not necessarily release radioactive material to the environment, since the contamination is integrally bonded to the inner surfaces of the steam generator tubing. 3.3.5 Accidents during Water Transportation The barge could suffer an accident in Biscayne Bay, the Atlantic Ocean, or the Savannah River, which could involve the release of radionuclides in two pathways to man: through the air or through the water. Barge accidents involving releases through the air may be readily dismissed because the cask is well-sealed and because they are so improbable. The accident rate for barges is estimated to be 1 x 10 5 accidents per mile.24 Only two percent of possible accidents are estimated severe enough to cause a breach in either the shipping cask or SGLA shell. If such an accident did occur, both barriers would probably not be breached because of the slow speeds of barge and ship traffic. i Of greater probability are accidents in which the SGLA would sink and subse-quently De breached by water pressure, so that radionuclides would have an t opening to pass into the water. Such an accident could involve a fire, for example from ignited spilled fuel from a ship colliding with the barge. A depth of 80 to 90 feet is deemed sufficient to buckle and crack a two inch thickness of steel. Consequently, sinking accidents on the Bay or River may i involve breaching of one but not both of the containment envelopes provided by the shipping cask and the SGLA shell, cover plates, and junction welds. Even if both envelopes were ruptured, however, no radioactivity would likely seep i ~, outside the envelopes because the radioactivity is fixed to the insides of the i tubes and because it is not readily soluble in water. l In a barge accident involving fire, no radiological consequences are expected, l either by releases into the air or into the water, because the most severe t fire, while heating the cask surface to several hundred degrees of temper-ature, will heat the corrosion products by only a small amount, insyfficient to cause release. Additional description of barge fires is given below. "" Environmental Assessment, Steam Generator Tube Integrity Program, Surry Steam Generator Project," 00E/EA-0102 (March 1980), p. 23. 240p cit, NUREG-0170, Table V-7. A-9 l

If the SGLA is lost at sea and is not recovered, the population dose is deter-mined assuming that the generator is breached and all of the radioactive inventory inside the SGLA is immediately released to the ocean. This assumed e release is very conservative, since many years are required to leach the radioactive material from the tubing. The largest population that may be exposed is estimated to be one million people.,, Releasing all the inventory would provide a population dose of less than 10 a person-rem per year to this population.15 The per capita dose represents 10 s of the annual dose received by an individual from natural background radiation. The most significant pathway to man for the radionuclides released would be through the ingestion of contaminated seafoods. 3.3.6 Accidents during Ground Transportation Various accidents are theoretically possible during ground transportation, including collapse of the road surface, brake failure, tire blowout, failure of a bridge, underground pipe, or culvert, tipping while negotiating a curve, and collision. Accidents such as tire blowouts or the collapse of the road, followed by tipping of the trailer, would not impart sufficient mechanical shock to breach the SGLA. An operator walking alongside the truck will control turning; the truck speed will be kept low enough to avoid overturning of the cargo. The truck will unlikely reach excessive speeds due to a total brake failure, since both the tractor and trafier have independent braking systems, each of which is adequate to stop both tractor and trailer. Finally, the applicant asserts there are no bridges off which the truck could fall, and the roads and crossroads over which the truck will travel will be blocked to other traffic, thus avoiding potential collisions.18 The worst case acc' dent (that producing the greatest mechanical shock to the SGLA) would be a brake failure on an incline followed by an impact of the shipping cask and SGLA onto an unyielding surface. i Under such circumstances, the speed of the trtek and its cargo is not likely l to be large. A runaway truck traveling 100 feet down a two to three percent j grade will be going about 10 miles per hour at the bottom of the grade.17 l Even if the speed were larger, it wduld not likely be large enough to cause a breach in both the shipping cask and SGLA on impact, because the SGLA would probably deform +.he surface. o 3.3.7 Accident involving Fire In a barge accident involving fire, the generator would remain sealed. The pressure induced from expansion of the air would not cause the generator to i l "" Consequences of Postulated Losses of LWR Spent Fuel and Plutonium Shipping Packages at Sea," PNL-2093 (October 1977). 160p cit, Uhrig letter, p. 6. 170p cit, DOE /EA-0102, p. 20. I A-10 l L

fail. No pressure would result from the decomposition of the corrosion products. The layer of corrosion products containing the radioactive material is quite refractory (Ni Fe 0 compounds). The melting point for the layer ranges between 4 x y 2900 and 3270*F (1600 to 1800 C). The most severe fire would raise the temper-ature of the cask surface several hundred degrees, but would not raise the temperature of the tubes inside the SGLA to the melting point of the corrosion products. Consequently, no radioactive material would volatilize.

Instead, l

any increase in temperature would promote diffusion between the Ni Fe 0 crud 4 x y layer and the Incont.1 tube wall. The diffusion reactini would make the crud layer more adherent to the tube walls 18 No radioactise material would be l released to the environment. 4.0 Disposal Impacts The volume of each steam genericor lower assembly (SGLA) is assumed to be 3200 ft3 3 (90m ). This volume is based on the maximum dimensions of the SGLA (lenght 34 feet, diameter 10.6 ft at the tube sheet end 13.1 at the transitiors cone end, figure 3.2 (feet). The total volume for the six SGLA's is 20,000 ft3 l (560m ). The total curie content of the six SGLA's is 1500 Ci, assuming each a SGLA has 250 Ci. i The remaining burial capacity at the Barnwell, SC, disposal site at the end of 1979 was 35 million ft3 (1 million m ). Thus, disposal of all six SGLA's would 3 use up approximately 0.06 percent of the Barnewell site capacity. This volume represents a small portion of the remaining disposal site capacity. I In 1979 Barnwell accepted wastes containing 315,000 Ci of byproduct material. The SGLA disposal represents 0.5 percent of the 1979 curie inventory accepted for disposal. This percentage represents a small fraction of the total curie inventory accepted annually for disposal. l The Turkey Point plant has shipped an average of 34,660 ft3 (980m ) of material 3 containing an averg e of 430 Ci annually from 1973 through 1979. The SGLA I 1 disposal is, therefore, approximately equivalent to the annual average waste l volume and three times the annual average Ci content of waste shipped for disposal from the Turkey Point station. These values do not represent a signi-ficant additional impact than exists from the routine shipment of radwaste by 9 the licensee. The State of South Carolina has implemented a license condition at the Barnwell disposal site which reduces the allowable monthly volume to be accepted for disposal. The allowable waste volumes for the disposal site from January 1981 are presented in Table 1. " Ibid., pp.27, 8. 2. A-11

Table 1 BARNWELL WASTE VOLUME LIMITATIONS Month / Year Volume Limit (ft ) 3 January 1981 133,500 (3,780m ) 3 February 1981 133,500 (3,780m ) 3 March 1981 133,500 (3,780m ) 3 April 1981 122,500 (3,470m ) 3 May 1981 122,500 (3,470m ) 3 June 1981 122,500 (3,470m ) 3 July 1981 111,000 (3,150m ) 3 August 1981 111,000 (3,150m ) 3 September 1981 111,000 (3,150m ) 3 October 1981 100,000 (283m ) 3 each month thereafter 100,000 (283m ) 3 The disposal of six SGLA's over an assumed 1 year period could represent as much as 2 percent of the volume limits allowed to be received at Barnwell over that period. In order to assure that Barnwell site will not exceed the volume limits placed on it by the State of South Carolina, each of the Barnwell customers has received individual volume limitations. The volume limitations for Turkey Point are given in Table 2. Allocations beyond February 1981 are not available at this time. However, for those months after February the allocations have been estimated by scaling the February value down by the same factor used in the l overall Barnwell volume limitations. l The SGLA disposal represents up to 83 percent of the Turkey Point allocation for an assumed 1 year period over which waste would be shipped to Barnwell. Unless a specific exemption for SGLA disposal is obtained from the State of South Carolina, the licensee would use up his allocation and would, therefore, be required to store on-site routinely generated radwaste until a new disposal allocation is authorized. Disposal Site Personnel Exposures The occupational exposures to disposal site personnel for monitoring, off-loading l and disposing of a single SGLA are estimated to be 2.4 person-rem. This value is based on an exposure rate of 0.03 r/hr and an 80 person-hour effort required at the disposal site. The total occupational exposure for disposal of the six l j SGLA's are, therefore, estimated to be approximately 15 person-rem. i 1 l{ 12

-.. _ = - - _ 2 l 1 I d "I Table 2 i TURKEY POINT VOLUME ALLOCATIONS FOR THE BARNWELL DISPOSAL SITE

• l l

Month / Year Volume Limet (ft ) 3 January 1981 2,650 (75m ) 3 February 1981 2,650 (75m ) 3 March 1981 2,650 (75m ) a April 1981 2,430 (69m ) 3 1 May 1981 2,430 (69m ) 3 2 June 1981 2,430 (69m ) a July 1981 2,200 (62m3) August 1961 2,200 (62m3) i September 1981 2,200 (62m3) October 1981 2,000 (57m3) i 1 i i c J ^ Values from April through OctoLer are estimated b) sealing according to the Barnwell reductions in Table 1. 9 7 N j A-13 ~ - - -. - -. - -- - -}}