ML17266A505

Requests Review of Encl Revised Preoperational & Operational Biological Monitoring Program.Location of Intake & Discharge Pipes in High energy-low Impact Area Requires Only Minimal Monitoring W/Emphasis on Sea Turtle Studies

ML17266A505
Person / Time
Site:	Saint Lucie
Issue date:	09/10/1981
From:	Zeller H ENVIRONMENTAL PROTECTION AGENCY
To:	Ballard R Office of Nuclear Reactor Regulation
Shared Package
ML17266A506	List: ML17266A505 ML17266A507
References
4SA-EB, NUDOCS 8110010369
Download: ML17266A505 (72)

Text

<< )~I+

~ 'I I <<lJ 2 I A <<3<<%:'L P\ t W 0 s

/ QAa g'L +(I<<

I I ( I Qj I ~l

~ ~

~

~ I l A,(M

~ g%

~

++<<6 g~

<< I'M. l I <N

~

it<<I ~

=~ 3l~ ~3 I V

-:SEP 1 O I":"=1 REF s 4SA-EB Dry Ronald L. Ballard, Chief Environmental Engineering Branch Division of Engineering U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Dr. Ballard:

\~

We appreciate NRC participation in the'eview of the St. Lucie Nuclear Plant preoperational and operational biological monitor-ing proposal. Your comments with those from NMFS, FWS, and the Florida DER were carefully considered in our deliberations.

Concurrence of general changes to the proposal were received from Mr. Masnik of your staff. The revised monitoring plan, in lieu of a comprehensive study, should adequately warn of sub-stantial damage to biota. The location of the, intake and dis-charge pipes are in what EPA considers a high energy-low 'mpact area; thus, we are inclined to require a minimal monitoring effort with ma)or emphasis upon studies of sea turtles. Your contribution to the sea turtle monitoring plans will be most appreciated. I understand Mr. Mike Masnik is assessing potential impacts of the nuclear Plant on sea turtles and will be contact-ing EPA about the evaluation.

The enclosed revision of the St. Lucie preoperational and opera-tional biological monitoring program is for your information/

review. We believe will meet with your approval.

it addresses your concerns, and hopefully it Sincerely yours, H ward D. Z&l r Enforcemen D'ision Enclosure ccrc Mike Masnik NRC 811001036'P 810910 PDR ADQCK 05000335 P eon C-47

AB-358 STLU1 MON ITOR1-55 PROPOSED ST. LVCIE PLANT PREOPERATIONAL AND OPERATIONAL BIOLOGICAL MONITORING PROGRAM AUGUST 1981 810910 8110010374 0500033S PDR ADQCK PDR P

PROPOSEO ST. LUCIE PLANT PREOPERATIONAL ANO OPERATIONAL BIOLOGICAL MONITORING PROGRAM E. GENERAL The ecological baseline study of Florida Power & Light Company's (FPL) St. Lucie Unit No. I was designed and implemented by the staff of the Florida Oepartment of Natural Resources Marine Research Laboratory.

Five offshore sampling stations were established (Figure 1) and sampling was conducted from July 1971 to August 1974. These results have been reported as St. Lucie Plant- baseline data pr pared by the Florida Oepartment of Natural Resources (References 4-12). The last portions of the data 'analyses and report preparation for this baseline study are presently being completed. Folloyiig the sampling for the baseline study, the Environmental TechnicaT. Specifications (ETS) for the opera-tional monitoring program, contained in the operating license for St.

Lucie Unit No. 1 issued by the Nuclear Regulato"y Commission (NRC), were written. These specifications delineated the biotic communities to be studied and stated that sampling was to be conducted at the same five stations established for the baseline study. The objective of the opera-tional monitoring study was to gather data for comparison with data obtained during the baseline study.

In March 1976, sampling for the operational monitoring program was begun by Applied Biology, Inc. (ABI). In addition to the five stations established for the baseline study, a nearshore site south of the plant was selected -as a control station. This control station was located distant from the plant and therefore away from possible influence from warmwater discharges. In accordance with the ETS, collections were made 1

C-49

YARDS KILOMETERS 2000 1000 0 1 SCALE v'

'-i. Q0

~ l 1~

""Iy

~

c

r

r'

~

( ~,

)

r~

~0 Qs

~

~

~ ~

41

~

D

~ ~

0

, 'L o nl

~ ~ ~

4 v'

- .": '. Offshore Discharge j ); 1 St. LUCI Plagt, ffshare 1

" '? .... " ""..'ntake Qg

~ ~

I ~ ~ ~

\ h Figure 1, Location of the five offshore sampling stations ( 1-5) established for baseline study and the control ( C ) station designated for the operational monitoring study.

2 C-50

to assess benthic organisms, plankton, nekton, macrophytes, water quality and migratory sea turtles. The results and analyses of these collections have been reported annually (Ref. 1, 2, 3, 16).

The five offshore stations were established by the Florida Oepartment of Natural Resources (FONR) before a comprehensive evaluation of the offshore currents was available. More recently, water current data (Ref. 14) have been obtained that indicates that if the stations were relocated they could better evaluate the biological communities in areas of potential plume impact. As shown in Figure 2, the predominant surface currents, and subsequent plume orientation from the point of discharge (Station I), are to the north. Based on water current eval-uation and the results of the biolpgical monitoring program to date, FPL believes that certain revisions to the program prescribed in the ETS and/or NPOES Permit are appropriate. The program described herein reflects these re'visions and would be used by both St. Lucie Unit No. 1 (operational monitoring) and St. Lucie Unit No. 2 (preoperational and operational monitoring). It is -proposed that the program continue for 2 years after St. Lucie Unit No. 2 is operational.

In the regulatory scheme established by the FederaI Water Pollution Control Act of 1972 (FNPCA), 33 USGA 5 5 1251 et ~se ., the Environmental Protection Agency .(EPA) was given jurisdiction over all water quality matters relating to non-radiological liquid ef,luents. in its Ye11ow Creek decision (ALAB-515), the NRC's Atomic Safety and Li.censing Appeal Board held that the NRC may not specify water quality restrictions in

~

3

g l

0 /

'V (Q

S.

~ ~ ~

0 p ','L A t!

~ I' c ~$

. 2 'e;

~ e jK~

~W

<O'-'

I v>> I

~

e

~ V%

're

' 1 ~;

~ ~ '.

~ <<~ 'I r 0 Q/0

~

r ~e f'

~ er

~

~ ~ ~ ~

bo)o 1

~

~ r ~eI ~

~b e

~ ~

0

%i

~ /

rC ".

~ ~

4~

I

~

/~

ohio

~ C

~,

~ ~

~ I O'.i, "I $

C3I Ji, I

~

~ ~ ~

e<<

\

, O lucie',.

Ve (~ lan

~ ~

~ ~ 'e 1 'q. ~ .Cb e1e Ve O A't.

0 ~ <<

~ I~

I'r.

S ~e

~ ~

~

e>>e1 I

<< ~e e ohio

~ ee V~ ~

~I r>>I')

~

6.5'I (Pe e

Figure 2. Frequency distribution of sur face current direction in relation to operational monitoring sampling stations.

C-52

excess of those imposed by the EPA. On the basis of ALAB-515 and the water quality effluent limitations and monitoring requirements contained in the National Pollutant Discharge Elimination System (NPOES) permit issued by EPA pursuant to FWPCA for St. Lucie Unit Ho. 1, FPL has peti-tioned the NRC for the deletion of thermal'nd chemical monitoring requirements contained in the ETS for St. Lucie Unit 1. However, this request to the NRC did not address the aquatic biological monitoring requirements also contained in the St. Lucie Unit Ho. 1 ETS. To remove this state of implicit dual. regulation, FPL proposes to incorporate appropriate aquatic biological monitoring requirements into the NPDES permit for St. Lucie. Units 1 and 2 and to request their deletion from the Unit 1 ETS. (The NRC operating license and accompanying ETS for St.

Lucie Unit No. 2 have not yet beerr issued.) The program described below is herewith submitted to the EPA for that purpose.

I I. PROPOSED BIOLOGICAL MONITORING PROGRAM

~0b'ective - To monitor the popuiations of sea turties, nektonic and benthic organisms of the Atlantic Ocean near the plant to determine the extent that plant operations may be influencing the nearshore ecosystem.

~Elf' -Tl Hagi I dpi 111b4 li terms of abundance and composition of the marine biotic community and 2) in terms of the relationship between physical, properties of the waters

, and the abundance and composition of the biological community.

Communities described below are to be evaluated to, determine potential alterations due to plant operation.

C-53

A. Benthic Or anisms Benthic organisms will be collected quarterly and inventoried as to kind and 'abundance.

B. Nektonic Or anisms Samples will be collected by gill netting once per month dur ing April through September and twice per month during October through March.

Kind and abundance of organisms present will be determined.

C. MIIM1i Analysis will be made at the surface at the same time as the nekton sample collections and near the bottom at the same time as the benthic sample collections. Parameters measured will be temperature,'alinity, disso'ived oxygen and turbidity.

D. ~fli S T Sea turtle nesting surveys will be conducted biannually on the FPL shore'line property ,and along selected control beaches. Sea turtles entering the intake 'will be removed, tagged and released ,back into the ocean on' continual basis.

E. 'e ortin Re uirements Results of the aquatic biological monitoring program shall be reported in an Annual Non-Radiological Environmental Monitoring Report to be submitted to the EPA.

C-54

III. IMPLEMENTATION OF PROPOSED BIOLOGICAL MONITORING PROGRAM A. Ihtroduction The monitoring program study design originated and was implemented in 1971 by the Florida Department of Na'tural Resources Marine Research Laboratory. The sampling regime was based on the ecological information available at the time. Sample locations were selected in relation to predicted plume direction, predicted plume areal extent (Ref. 4) 'and the major macrohabitats known to exist off Hutchinson Island. Stations 1, 2 and 3 were located in the prgdicted thermal plume area, while 4 and 5 were established as north and south controls located in the same macrohabitat as Station 2 (Ref. 5).

Since'972, extensive data on the biological communities near the St. Lucie Plant have been obtained (Refs. 1-3, 6-12, 16). Additional physical data have been gathered on winds (Ref. 13), currents (Ref. 12) and the thermal plume (Ref. 15).

These biological and physical studies indicate that effects of the St. Luci e discharge are 1 imi ted to surface areas near the point of discharge. The proposed study is therefore designed to evaluate the biological conditions in the near-field area of potential plume impact.

B.. Benthic Or anisms To assess the potential that there are thermal effects on the benthic community, quarterly samples will be taken at control Station BC, .

Station Bl, and at a station (B2) to be located just north of the thermal C-55

1 plume's warmest spot (Figure 3). Four, or more replicates will be taken.

Station 2 of the current program will be retained as Station 'C1 to help integrate the modified program with the existing data. Station 5 of the current program will be retained as Station B3 for at least,= one or two after Unit 2 goes on-line, to document the probability that there 'ears is no effect of cdmbined Units 1 and 2 discharge 't this location.

Benthic sampling at other offshore stations (3 and 4) will be terminated.

C. Nekt0n The sampling program will consist of nearshore gill netting. Two sampling stations will be established near the intake structure and three in the discharge area (Figure 4). The discharge station samples will pr ovide data, on near, intermediath and distant effects of the plume on fish distribution. Stations will be located in the thermal plume's warm-est spot and approximately 200 meters and 450 meters from this warmest spot. These stations will be sampled as follows: once per month during April through September when the commercially important migratory species are. generally not present offshore the St. Lucie Plant and twice per month during October through March when these species are present.

Station 2 (Cl) will be retained to help integrate the data from the modified program with the exisiting data.

D. Mi rator Sea Turtles Sea turtle nesting surveys will be conducted biannually during odd-numbered .years to monitor species, numbers and nesting characteristics.

4 The nesting surveys will be conducted during the summer nesting season on C-56

~ ~

'\ VAROS KILOMETERS 0

~

~ ) 2000 'l000 1

~ ~

SCALE

~ ~ ~

I~

I~

N

~ OO r~ ~ 0

~S ~

\

~

~"

0

~ a 0 s3 0

~~ tf t Q)

Vo I ~ la ~ :m . ~

.-'C)..

p ~ ' . I11

~ t

~ O I11 W ~ ~~

);. Ql g~

V ~

n

'el

.I ~ ~

G sa St. Luci Plant.: ~~

~.

I~

\

1

~ ~+ t j I

'jQ

'. 0 BC (CONTROL STATION)

Y

~ 0 I

~

C'igure

3. Location of benthic sampling stations.

C-57

I li P

0>>

~ II I

/

IHERMAL PLUME NOTE: These stations vill be oriented

.:I I as needed to maintain their N positions relative to the plume.

I2ISCHARGE CANAL IR

~

~~

s

~Unit 2 Discharge (submerged multiport diffuser)

ST. LUCIE Unit 1 Discharge (submerged 2-port Y diffuser)

POWER PLANT 2I

T P~

~,

0 OF INTAKE CANAL Q Submerged Intake Structure

'J':

Fl INDiAN RIVER 0 .skm

- -AI L

Figure 4. Location of gill net stations.

the FPL shoreline property and along selected control beaches. Specifics of the nesting surveys, such 'as sampl ing frequency and'he amount of beach sampled, vary between study years and are established following "input from the appropriate state and federal agencies.

Sea turtle removal from the intake canal is conducted on a continual basis. The turtles are removed with nets from the canal, measUred and weighed, tagged and released back into the ocean. The utmost care is taken so as not to injure the animals.

E. ~0 Samples for water quality analysis will be collected concurrently with the biological samples.

IV. SIGNIFICANT CHANGES FROM THE ETS MONITORING PROGRAM The ETS contain a provision for modification of the program based upon the data accumulated after two years of operation. The program pro-posed in Section II,above differs significantly from that prescribed in the St. Lucie Unit No. 1 ETS in several respects. These changes and their bases are described below.

A. Plankton - The monthly collection of phytoplankton and zooplankton has been deleted.

C-59

J usti fi cati on Interstation comparisons t have 'hown that concentrations of zooplankton, phytoplankton and chlorophyll a generally have been higher in the area of the Station 1 discharge than at the other stations, suggesting some enhancement of plankton concentrations due to the thermal input. "

It .is 'unlikely that differences in plankton concentrations are significant in the nigh. energy, nearshore location under consideration.

Continued plankton monitoring does not appear to be justified.

B.'ektonic Or anisms - Collecting allocations of samples by trawling and seining has been deleted and gill net station have been revised.

Justification The ETS allowed collection 5f samples by "trawling, seining, or other suitable method". Trawling and beach seining are sampling tech-niques that are highly seIective for bottom dwelling and surf zone dwelling forms. During operational monitoring, neither of these com-munities appeared to be influenced by .the ,thermal discharge (Ref. 3).

Gi 11 netting obtains 'samples . in the .water column and is an effective method for collecting sport and commercial fish species. The proposed schedule emphasizes collections during the period of the year when migra-tory species such as bluefish, Spanish mackerel and king mackerel are near the St. Lucie Plant. Stations moved to the immediate plume area will better assess the influence of the thermal discharge on the move-ments of fishes in the area.

12 C-60

C. Macroohvtes- - The quarterly collection of macroscopic aquatic vege-tation has been deleted.

Justification The highest diversity of algae, 88 species, was collected during the third year of the study. "The number of species collected was lowest in early spring and highest in summer and early fall. This seasonal pattern was typical for subtropical marine vegetation. Diversity was higher near shore because Grift (unattached) algae were the predominate form:; and these were carried inshore by the prevailing winds and currents (Ref. 3).

Vegetation distribution and gr~th at all

~

nearshore stations sur-seems to be limited by a lack 'eyed of appropriate substrate for vegeta-ti6n attachment. Well-developed macrophyte communities may occur on isolated rock outcroopings, but the chances of the collecting dredge encountering one of these outcroppings is remote. Because the attached macrophyte community is'o limited, it is not considered an important food source or habitat for organisms living in 'the St. Lucie area.

Because of the abov'e, the samoling provides little useful data and there is no need for further monitoring of macrophytes.

13 C-61

Justification Oata from the control station, located distant from the St. Lucie Plant, were compared with results from station-specific water parameter analyses. Oata from the literature for marine waters of nearshore coastal environments adjacent to the plant were also compared with the present study. Oats comparisons (Ref. 3) indicated:

a. Nearly all parameters measured varied significantly dur ing dif-ferent months of the year; and

b. There were no significant differences in parameters among sta-tions or at different depths.

These results show that the operation of the St. Lucie Plant has no I

significant effect on the selected ~trients in this study. Continued nutrient analyses does,not appear warranted.

E. Mi rator Sea Turtles - Various requirements relating to the effects of the di scharge thermal plume and temperature stress, hatching and rearing factors for migratory sea turtles have been deleted.

j Justification The requirements of the ETS have been satisfied.' report was pre-pared (Ref. 2) and submitted to the HRC by FPL letter No. L-78-109, dated 30 March 1978, that described studies performed to determine the effects of the discharge thermal plume on turtle nesting patterns and turtle hatchling swimming. Additionally, contro1 studies on temperature stress, hatching and rearing factors conducted using turtle eggs from displaced nests were reported. The results of the studies of turtle hatchlings 14 C-62

show no evidence that potential nearshore surface temperatures from the plant will~ cause permanent impairment or mortality (Re'f. 2).

Entrainment of A uatic Or anisms ETS 4.1 >> Various requirements r elating to assessment of the effects on planktonic organisms of passage through the plant condensers have been deleted.

Justification The results of the ichthyoplankton and zooplankton sampling have been presented in the Annual Non-Radiological Environment'al Monitoring Reports for 1976, l977, 1978 and 1979 (Ref. 1, 2, 3, 16).

These studies show that 4he i~shore ocean waters near the St. Lucie Plant are not typical of a productive fish nursery area. Physical characteristics needed in a nursery area ar low or 'luctuating salinities, silt-sand-mud bottom, and extensive, beds of rooted aquatic vegetation. ChemicaIly, the waters in the St. Lucie Plant area are homo-geneous with little seasonal variations. Physically, the nearshore areas are characterized by the presence of relatively constant salinities, shell-hash sediments and the absence of significant macrophytic grassbeds.

important migratory sport and commercial fishes were not found to be spawning in the area of the St. Lucie Plant. In general, low con-centrations ,.of fish eggs and larvae have been recorded in ,the intake canal, which confirms that entrainment is not significant. Zooplankton losses through entrainment are not significant.

15 C-63

Based on the above, the required Entrainment Studies need not be included in the operational monitoring program.

16 C-64

LITERATURE C ITEO

1. ABI. 1977. Ecological monitoring at the Florida Power & Light Company, St. Lucie Plant, annual report, 1976. Report to Florida Power & Light Company, Miami, Fla.

2. . 1978. Ecological monitoring at the Florida Power & Light Company, St. Lucie Plant, annual report, 1977. Report to Florida Power & Light Company, Miami, Fla.

3. . 1979. Florida Power & Light Company, St. Lucie Plant annual non-radiological environmental monitoring report, 1978. Vol. II and III. Biotic monitoring. Report to Florida Power & Light Company, Miami, Fla.

4. Florida Power & Light Co.. 1971. Hutchinson Island plant unit No. 1 environmental report Docket No. 50-335. 20 May 1971. Florida Power & Light Company, Miami, Fla.

5. Florida Department of Natural Resources. 1972. Preliminary environmental studies of coastal waters near Hutchinson Island, Florida. Progress report to Florida Power & Light Company, Miami, Fla.

6. Gallagher, R.M. 1977a.. Nehrshore marine ecology at Hutchinson Island, Florida: 1971-1974. I. Rationale and methods. Fla.

Mar. Res. Publ. No. 23:1-5.

7. 1977b. Nearshore marine ecology at Hutchinson Island, Florida: 1971-1974. II. Sediments. Fla. Mar. Res.

Publ. No. 23:6-24.

8. Morth, D.F., and M.L. Hollinger. 1977. Nearshore marine ecology at Hutchinson Island, Florida: 1971-1974. III. Physical and chemical environment. Fla. Mar. Res. Publ. No. 23:25-85.

9. Futch, C.R., and S.E. Owinell. 1977. Nearshore marine ecology at Hutchinson Island, Florida: 1971-1974. IV. Lancel ets and fishes. Fla. Mar. Res. Publ. No. 24:1-23.

10. Camp, O.K., N.H. Whiting, and R.E. Martin. 1977. Nearshore marine ecology at Hutchinson Isl and, Fl orida: 1971-1974. V.

Arthropods. Fla. Mar. Res. Publ. No. 25:1-63.

11. Gallagher, R.M., M.L. Hollinger, R.M. Ingle, and C.R. Futch. 1972.

Marine turtle nesting on Hutchinson Island in 1971. Fla. Dept.

Nat. Resour., Mar. Res. Lab. Spec. Sci. Rept. No. 37; 1-11.

12. North, O.F., and J .B. Smith. 1976. Marine turtle nesting on Hutchinson Island in 1973. Fla. Mar. Res. Publ. No. 18: 1-17.

13. Dames & Moore. 1977. Graphical and tabular wind roses. St. Lucie, Hutchinson Island, Florida, 1973. Report to Florida Power &

Light Company, Miami, Fla..

17 C-65

LITERATURE CITEO (continued)

14. Envirosphere Co. 1976. St. Lucie Plant site ocean current analysis.

Report to Florida Power & Light Company, Miami, Fla.

15. 1977. Thermal evaluation study. St. Lucie Unit 1 ocean diffuser. Report to Florida Power & Light Company, Miami, Fl a.

16. ABI. 1980. Florida Power & Light Company, St. Lucie Plant annual non-radiological environmental monitoring report, 1979. Vol. II and III. Biotic monitoring. Report to Florida Power & Light Company, Miami, Fla.

18 C-66

ST. LUCIE UNIT NO- 2 BIOLOGICAL MONITORING PROGRAM - OPERATIONAL PHASE ADDITIONS The following additions to the Biological Monitoring Program sub-mitted to EPA on 3 April 1980 are recommended for the program to serve St. Lucie Unit No. 2 in the operational mode.

d. ~BIT I . ~gifi I - T ddfgf I BII g tions wil,l be added near the Unit No. 2 discharge. These stations will be in close proximity to the discharge pipe with one north and one so'uth of the pipe. Stations will be sampled quarterly with four or more replicates collected to assess the taxonomic composition and abundanace.

Justification The Unit No. 2 discharge pipe will extend 1875 feet further offshore than the Unit No. I pipe. There is a habitat and sedi-ment change from beach terrace gray sand near shore (e.g. Unit 1 discharge area) to a shell hash substrate in the area of Unit 2 discharge. The ongoing monitoring program has shown these habitats to support somewhat different communities. These different communities may react differently to a heated discharge.

B. Bk . ~gifi I -T ddf I I ffk g111 tions will be established. One station will be in the middle of the Unit No. 2 thermal plume's warmest area and the other, the control, about 200 meters upcurrent from this warmest spot. The stations will be sampled once per month during April through 20 C-67

September when the commercially important migratory species are generally not present offshore the St. Lucie Plant and twice per month during October through March when these species are present.

Justification The adult fish community in the discharge plume from Unit No. 2 should be examined to determine if attraction I or exclusion is occurring. The St. Lucie No. 2 discharge pipe will extend about 1875 feet past the Unit No. 1 point of discharge and the discharged water may influence fish movement 'in the area.

C. ~1i . Physical parameters will be stations frequency as the biological

~if'easured at the same. and samples.

Justification Mater quality determinations are made to support the biological program and should be taken concurrently with biological sampling.

This program will enable an evaluation of the impact of the Unit No.

2. discharge to be made. The addition of these stations and sampling regimes takes into consideration the option of directing the plant discharge through the St. Lucie Unit No.. 2 diffuser pipe 'if one unit is down.

C-68

APPENDIX D HISTORICAL AND ARCHEOLOGICAL SITES Letter from George M, Percy, Deputy State Historic Preservation Officer, Division Archives,'Historic and Records Management, Florida Department of State to

'f B. J. Youngblood, U,S. Nuclear Regulatory Commission, dated April 13, 1981.

~ ~

St. Lucie 2 D)S

g <Hc srsp s

FLORIDA DEPARTMENT OF STATE George Firestone Secreta of State DIVISION OF ARCHIVES. HISTORY AND RECORDS MANAGEMENT I L. Ross Morfell, Director April 13, 1981 (904) 488 l 480 In reply refer to:

Ms. Rowan Comex-Tesar Project Archaeologist (904) 487-2333 Mr. B. J. Youngblood, Chief Licensing Branch No.l U. S. Nuclear Regulatory Commission Washington, D.C. 20555

.Re: Your Letter of April 6, 1981 Cultural Resource Assessment Request St. Lucie Power Plant Unit No. 2, St. Lucie County, Florida Deax Mr. Youngblood:

In accordance with the procedures contained in 36 C.F.R.,

Part 800 (" Procedures for the Protection of Historic and Cultural Properties" ), we have reviewed the above referenced pro)ect for possible impact to archaeological and historical sites or properties listed, or eligible for listing, in the National Receister of Historic Places. Tha authorities for these procedures are the National Historic Preservation Act of 1966 (Public Law 89-665) as amended by P.L.91-243, P.L.

93-54, P.L.94-422, 94-458, and P.L.96-515 and Pxesiden-tial Executive Order P.L.11593 (" Protection and Enhancement of the Cultural Environment" ).

A review of the Florida Master Site File indicates that no archaeological or historical sites are listed for the pro-ject area. Furthermore, because of the location of the project, it is considered highly unlikely that, any significant, unrecorded sites exist in the vicinity. Therefore, it is the opinion of this office that, the proposed project will have no effect on any sites listed, or eligible for listing, in the National Register of Historic Places, or otherwise of national, state, or local significance.

0-2 FLORIDA-State of the Arts The Capitol Tallahassee, Florida 32301 (904) 488-3680

Nr. B. J. Youngblood April 13, 1981 Page Two On behalf of the Secretary of State, George Firestone, and the staff of the Bureau of Historic Sites and Properties, I would like to thank you for your interest and cooperation in the protec-tion of Florida's irreplaceable historic resources.

Since ely, George . Percy Deputy State Historic Preservation Officer GWP:Ceh

APPENDIX E NEPA POPULATION-DOSE ASSESSMENT

I II

E. 1 Introduction Population-dose commitments are calculated for all individuals living within 80 km (50 mi) of the St. Lucie Plant, employing the same models used for individual doses (see Regulatory Guide l. 109, Rev. 1) , for the purpose of meeting the "as low as. r'easonably achievable" (ALARA) requirements of 10 CFR Part 50, Appendix I. In addition, dose commitments to the population residing beyond the 80-km region, associated with the export of food crops produced within the 80-km region and with the atmospheric and hydrospheric transport of the more mobile effluent species, such as noble gases, tritium, and carbon-14, .are taken into consideration for the purpose of meeting the requirements of, the National Environmental Policy Act of 1969 (NEPA). This appendix describes the methods used to make these NEPA population dose estimates.

E. 2 Iodines and Particulates Released to the Atmos here Effluent nuclides in this category deposit onto the ground as the effluent moves downwind, thus the concentration of these nuclides remaining in the plume is continuously being reduced. Within 80 km (50 mi) of the site, the deposition model in Regulatory Guide 1. 111, Rev. 1, is used in conjunction with the dose models in Regulatory Guide 1. 109, Rev. 1.~ Site specific data concerning pro-duction and consumption of foods within 80 km (50 mi) of the reactor are used.

For estimates of population doses beyond 80 km (50 mi) it is assumed that excess food not consumed within the 80-km region will be consumed by the population beyond 80 km (50 mi). It is further assumed that none, or very few, of the particulates released from the facility will be transported beyond the 80-km region; thus they will make no significant contribution to the population dose outside the 80-km region except by export of food crops. This assumption was tested and found to be reasonable for the St. Lucie Plant.

E. 3 Noble Gases Carbon-14 and Tritium Released to the Atmos here For locations within 80 km (50 mi) of the reactor facility, exposures to these effluents are calculated with a constant mean wind-direction model according to the guidance provided in Regulatory Guide l. 111, Rev. 1, and the dose models described in Regulatory Guide 1. 109, Rev. 1. For estimating tPe dose commitment from these radionuclides to the U.S. population residing beyond the 80-km region, two dispersion regimes are considered. These are referred to as the first-pass dispersion regime and the world-wide dispersion regime. The model for the first-pass dispersion regime estimates the dose commitment to the population from the radioactive plume as it leaves the site and drifts across the continental United States toward the northeastern corner of the U.S. The model for the world-wide dispersion regime estimates the dose commitment to the U.S. population after the released radionuclides mix uniformly in the world's atmosphere or oceans.

E.3. 1 First-Pass Dis ersion For estimating the dose commitment to the U.S. population residing beyond the 80-km region due to the first pass of radioactive pollutants, it is assumed that the pollutants disperse in the lateral and vertical directions along the plume path. The direction of movement of the plume is assumed to be from the site toward the northeast corner of the U.S, The extent of vertical dispersion is assumed to be limited by the ground plane and the stable atmospheric layer St. Lucie 2 OES E-1

aloft, th'e height of which determines the mixing depth. The shape of such a plume geometry can be visualized as a right cylindrical wedge whose height is equal to the mixing depth. Under the assumption of constant population density, the 'population dose associated with such a plume geometry is independent of the extent of lateral dispersion, and is only dependent upon the mixing depth and other nongeometrical related factors.4 The mixing depth is estimated to be 1000m, and a uniform population density of 62 persons/km~ is assumed along the plume path, with an average plume transport velocity of 2 m/s.

The -total-body population dose commitment from'he first-pass of radioactive effluents is due principally to external exposure from gamma-emitting noble gases, and to internal exposure from inhalation of air containing tritium and from ingestion of food containing carbon-14 and tritium.

E.3.2 World-Wide Dis ersion a For estimating the dose commitment to the U.S. population after the first-pass, world-wide dispersion is assumed. Nondepositing radionuclides with half-lives greater than one year are considered. Noble gases and carbon-14 are assumed to mix uniformly in the world's atmosphere (3.8 x 10' ), and radioactive decay is taken into consideration. The world-wide dispersion model estimates the activity of each nuclide at the end of a 15-year release period (midpoint of reactor life) and estimates the annual population dose commitment at that point in time, taking into consideration radioactive decay and physical removal mechanisms (e.g., C-14 is gradually removed to the world's oceans). The total-body population dose commitment from the noble gases is due mainly to external exposure from gamma-emitting nuclides, while from carbon-14 it is due mainly to internal exposure from ingestion of food containing carbon-l4.

The population dose commitment due to tritium releases is estimated in a manner similar to that for carbon-14, except that after the first-pass, all of the tritium is assumed to be immediately distributed in the world's circulating water volume (2.7 x 10 m ) including the top 75m of the seas and oceans, as well as the rivers and atmospheric moisture. The concentration of tritium in the world's circulating water is estimated at the point in time after 15 years of releases have occurred, taking into consideration radioactive decay; the population dose commitment estimates are based on the incremental concentration at that point in time. The total-body population dose commitment from tritium is due mainly to internal exposure from the consumption of food.

E.4 ~Li id Eff1 Population dose commitments due to effluents in the receiving water within 80 km (50 mi) of the facility are calculated as described in Regulatory Guide 1.109.

It is assumed that no depletion by sedimentation of the nuclides present in the receiving water occurs within 80 km (50 mi). It also is assumed that aquatic biota concentrate radioactivity in the same manner as was assumed for the ALARA maximally exposed individual evaluation. However, food consumption values appropriate for the average, rather than the maximum, individual are used. It is further assumed that all the sport and commercial fish and shellfish caught within the 80-km region are eaten by the U.S. population.

St. Lucie 2 DES E-2

Beyond 80 km (50 mi), it is assumed that all the liquid-effluent nuclides except tritium have deposited on the sediments so that they make no further contri-bution to population exposures. The tritium is assumed to mix uniformly in the world's circulating water volume and to result in an exposure to the U.S.

population in the same manner as discussed for tritium in gaseous effluents.

E.5 References U.S. Nuclear Regulatory Commission, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I." Revision 1, Regulatory Guide 1.109, October 1977.

2. Title 10 Code of Federal Regulations Part 50, "Domestic Licensing of Production and Utilization Facilities." January 1981.

3. U:S. Nuclear Regulatory Commission, Regulatory Guide l. 111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Mater-Reactors." Revision 1, July 1977.

4. U.S. Nuclear Regulatory Commission, "User's Guide to GASPAR Code."

NUREG-0597, June 1980.

St. Lucie 2 DES E-3

APPENDIX F EXAMPLES OF SITE-SPECIFIC DOSE ASSESSMENT CALCULATIONS l

St. Lucre 2 DES F-$

F. 1 Cal cul ational A roach As mentioned in the text the quantities of radioactive material that may be released annually from the St. Lucie Plant are estimated on the basis of the description of the radwaste systems in the applicant's ER-OL and FSAR and by using the calculational model and parameters developed by the staff.'hese estimated effluent release values along with the applicant's site and environ-mental data in'he ER-OL and in subsequent answers to staff questions are used in the calculation of radiation doses and dose commitments.

The models and considerations for environmental pathways that lead to estimates of radiation doses and dose commitments to individual members of the public near the Plant and of cumulative doses and dose commitments to the entire population within 80km (50 mi) of the site as a result of Plant operations are discussed in detail in Regulatory Guide 1. 109.~ Use of these models with additional assumptions for environmental pathways that lead to exposure to the general population outside the 80-km region are described in Appendix E of this Draft Environmental Statement.

The calculations performed by the staff for the releases,to the atmosphere and hydrosphere provide total integrated dose commitments to the entire population within 80 km (50 mi) of the Plant based on the projected population distribution in the year 2000. The dose commitments represent the total dose that would be received over a 50-yr period, following the intake of radioactivity for 1 yr under the conditions existing 15 years after the St. Lucie 2 begins operation (i.e., the mid-point of St. Lucie 2 operation). For younger persons, changes in organ mass and metabolic parameters with age after the initial intake of radioactivity are accounted for.

F.2 Dose Commitments from Radioactive Effluent Releases Radioactive effluents released to the atmosphere and to the hydrosphere from the Plant will result in very small radiation dose commitments to individual members of the public and to the general population. The staff estimates of the expected gaseous and particulate releases (listed in Table F-1) and the expected liquid releases (listed in Table F-8) along with the site meteorological and hydrological considerations (summarized in Tables F-2 and F-9 respectively) were used to estimate radiation doses and dose commitments.

Annual average relative concentration (X/g) and relative deposition (D/g) values are calculated using the straight-line Guassian model described in Regulatory Guide 1. 111.3 A 2-year period of record (January 1977-December 1978) of meteorological data was used. Previous data collected at the site could not be incorporated because of changes in instrument heights and sensor specifica-tions. Mind speed and direction data were based on measurements at the 10 m level and atmospheric stability was defined by the vertical temperature gradient measured between the 10 m and 57.9 m levels. All releases were considered to be at ground level with mixing, in the turbulent wake of plant structures. The results of the straight-line model were adjusted to consider spatial and temporal variations in airflow. These adjustments were based on a comparison performed by the applicant of the results of a variable trajectory model with the results of the straight-line model. This comparison indicated that the straight-line model may underpredict X/g values by up to a factor of 1.7. The results of the St. Lucie 2 DES F-2

straight-line model were not adjusted by factors of less than 1.0 pending a reevaluation of such adjustments for distances greater than about 8 km (5 mi).

The influence of the Indian River was not considered for calculations of annual average 2/g values because the river woud likely act as both a stabilizing fac-tor and a de-stabilizing factor. depending on time of day and season of the year.

The stabilizing and de-stabilizing influences would likely counter-balance over the course of a year.

F. 2. 1 Radiation Dose Commitments to Individual Members of the Public As explained in the text, calculations are made for a hypothetical individual member of the public (i. e., the maximally exposed individual) .who would be expected to receive the highest radiation dose from all appropriate pathways.

This method tends to overestimate the doses since assumptions are made that would be difficult for a real individual to fulfill.

Individual receptor locations and pathway locations considered for the maxi-mally exposed individual are listed in Table F-3. The estimated dose commitments to the individual who is subject to maximum exposure at s'elected offsite locations from airborne releases of radioiodine and particulates, and waterborne releases are listed in Tables F-4, F-5, and F-6. The maximum annual total body and skin dose to a hypothetical individual and the maximum beta and gamma air dose, at the site boundary, are presented in Tables F-4, F-5, and F-6.

The maximally exposed individual is assumed to consume well above average quantities of the potentially affected foods and to spend more time at poten-tially affected locations than the average person as indicated in Tables E-4 and E-5 of Regulatory Guide l. 109.~

F.2.2 Cumulative Dose Commitments to the General Po ulation Annual radiation dose commitments from airborne and waterborne radioactive releases from St. Lucie 2 are estimated for two populatjons in the year 2000:

(1) all members of the general public within 80 km (50 mi) of the station (Table F-5) and (2) the entire U.S. population (Table F-7). Dose commitments beyond 80 km (50 mi) are based on the assumptions discussed in Appendix E.

For perspective, annual background radiation doses are given in the Tables for both populations.

F.3 References

l. U.S. Nuclear Regulatory Commission, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Mater Reactors (PWR-GALE Code)." NUREG-0017, U.S. Nuclear Regulatory Commission,

. April 1976.

2. U.S. Nuclear Regulatory Commission, Regulatory Guide 1. 109: "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I."

Revision 1, Commission, October 1977.

3. U.S. Nuclear Regulatory Commission, Regulatory Guide l. ill: "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Mater Reactors." Revision 1, July 1977.

St. Lucie 2 DES F-3

Table F-l.a Calculated Releases of Radioactiv'e Materials In Gaseous Effluents from St. Lucie Plant Unit 1 (Ci/yr)

Nuclide Plant Stack Vent Turbine buil'ding I

Continuous Intermittent Ar-41 a 25 a Kr - 83m a a a Kr - 85m .2 1 1 Kr - 85 240 49 1 Kr - 87 1 a a Kr - 88 4 2 2 Kr - 89 a a a Xe - 131m 97 37 1 Xe - 133m 9 32 3 Xe - 133 5400 4700 200 Xe - 135m a a a Xe - 135 7 9 4 Xe - 137 a a a Xe - 138 a a a Total Noble Gases 11,000 Mn - 54 . 005 b b Fe - 59 . 002 b b Co - 58 . 016 b b Co - 60 . 007 b b Sr - 89 .00034 b b Sr' 90 .000062 b b Cs - 134 .005 b b Cs - 137 .008 b b Total Particulates .04

" 131 .061 . 0039 .04

- 133 .074 . 0031 .05

- 14 7 1

- 3 610 Less than 1.0 Ci/yr per reactor for noble gases arid carbon-14; less than 10-4 Ci/yr per reactor for iodine.

Less than 1X of total for this nuclide.

St. Lucie 2 DES F-4

Table F-l.b Calculated Releases of Radioactive Materials In Gaseous Effluents from St. Lucie Plant Unit 2 (Ci/yr)

Nuclide Plant Stack Vent Turbine building Continuous Intermittent Ar-41 b 25 a Kr - 83m a 1 1 Kr - 85m 2 14 a Kr - 85 210 35 a Kr - 87 1 3 a Kr - 88 18 2 Kr - 89 a a a Xe - 131m 102 42 a Xe - 133m 10 88 2 Xe - 133 6200 7500 180 Xe - 135m a a a Xe - 135 7 74 4 Xe - 137 a a a Xe - 138 a a a Total Noble Gases 15,000 Mn - 54 0. 0047 0.00022 Fe - 59 0. 0016 0.000075 Co - 58 0. 02 0.00074 Co - 60 0. 007 0.00034 Sr - 89 0. 00034 0.000017 Sr - 90 0.000062 0.000003 Cs - 134 0.003 0. 0004 Cs - 137 0.008 0. 00038 Total Particulates .05 I - 131 0. 058 0. 022 0. 04 I " 133 0. 071 0. 018 0. 048 C

- 14 7 1 H-3 510 b Less than 1.0 Ci/yr per reactor for noble gases and carbon-14; less than 10-4 Ci/yr per reactor for iodine.

b Less than 1X of total for this nuclide.

St. Lucie 2 DES F-5

Table F-2 Summary of Atmospheric Dispersion Factors (X/g) and Relative Deposition Values for Maximum Site Boundary and Receptor Locations Near the St. Lucie Plant Relative Location X/g (sec/ms) Deposition (m-~)

Site boundary (NW, 0.97 mi) 1.3 x 10-e 5.7 x 10-e Nearest residence (SE, 1.4 mi) 4.6 x 10-~ 2.2 x 10-s Nearest garden (M, 1.9 mi) 3.0 x 10-~ 1.4 x 10-~

The values presented in this table are corrected for radioactive decay and cloud depletion from deposition, where appropriate, in 'accordance with Regulatory Guide l. 111, Rev. 1, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routin'e Releases from Light Mater Reactors," July 1977.

"Nearest" refers to that type of location where the highest radiation dose is expected to occur from all appropriate pathways.

Table F-3 Nearest Pathway Loc'ations Used for Maximum Individual Dose Commitments for the St. Lucie Plant Location Sector Distance (mi)

Site boundary NW 0. 97 Residence SE 1.4 Garden W 1.9 Beta and gamma air doses, total body doses, and skin doses from noble gases are determined at site boundaries in the sector where the maximum potential value is likely to occur.

Dose pathways including inhalation of atmospheric radioactivity, exposure to deposited radionuclides, and submersion in gaseous radioactivity are evaluated at residences.

St. Lucie 2 DES

Table F-4.a Annual Dose Commitments to a Maximally Exposed Individual Near the St. Lucie Plant Unit 1 Location Pathway Doses (mrem/yr per unit)

Noble Gases in Gaseous Effluents Total Body Skin Gamma Air Dose Beta Air Dose (mrad/yr per (mrad/yr per unit) unit)

Nearest ~ite Direct radiation boundary fr'om plume 0. 25 0. 70 0. 42 1.2 (NW, 0.97 mi)

Iodine and Particulates in Gaseous Effluents Total Body Maximum Organs Thyroid Bone Nearest Site Ground deposition o.o51 (T) 0.051 (C) o.o51(c)

Boundary. Inhalation o.'o31 (T) O'.12 (C) o.'ooo9 (c)

(NW 0.97 mi)

Nearest Ground deposition o.o2 (c) 0.02 (C) o.o2 (c) resident Inhalation o.o1 (c) o.o47 (c) 0.00033 (C)

(SE, 1.4 mi) Vegetable consumption o.o8 (c) O.28 (C) 0.39 (C)

Nearest Ground deposition 0.012 (C) o.o12 (c) 0.39 (C) garden Inhalation 0.0066 (C) 0.028 (C) o'.ooo21 (c)

(W, 1.9 ml) Vegetable consumption 0.052 (C) 0.017 (C) O.'26 (C)

Li uid Effluents (Adults)

Total Body Organ Discharge Fish point Consumption 0. 0058 0.040 (Thyroid)

Invertebrate 0. 0018 0.47 (Thyroid)

Shoreline 0.00054 0.00054 (Thyroid)

"Nearest" refers to that site boundary location where the highest radiation doses as a result of gaseous effluents have been estimated to occur.

Doses are for the age group and organ that results in the highest cumulative dose for the location: T=teen, C=child, I=infant. Calculations were made for these age groups and for the following organs: GI-tract, bone, liver, kidney, thyroid, lung, and skin.

"Nearest" refers to the location where the highest radiation dose to an individual from all applicable pathways has been estimated.

St. Lucie 2 DES F-7

Table F-4.b Annual Dose Commitments to a Maximally Exposed Individual Near the St. Lucie Plant Unit 2 Location Pathway Doses (mrem/yr per unit)

Noble Gases in Gaseous Effluents Total Body Skin Gamma Air Dose Beta Air Dose (mrad/yr per (mrad/yr per Nearest ~ite Direct radiation boundary from plume 0. 40 0. 67 1.8 (NW, 0.97 mi)

Iodine and Particulates in Gaseous Effluents Maximum Organs Total Body Thyroid Bone Nearest site Ground deposition 0.062 (T) 0.062 (C) 0. 062 (C) boundary Inhalation 0.026 (T) 0.18 (C) (c)'.oo12 (NW, 0.97 mi)

Nearest Ground deposition 0.026 (C) 0.026 (C) 0.026 (C) resident Inhalation o.oo85 (c) 0.074 (C) 0.00049 (C)

(SE,.1.4 mi) Vegetable consumption 0.082 (C) 0.47 (C) 0.40 (C)

Nearest Ground deposition o.o15 (c) 0.015 (C) o.o15 (c) garden Inhalation o.oo56 (c) 0.042 (C) 0.00028 (C)

(W, 1.9 mi) Vegetable consumption 0.053 (C) 0.26 (C) 0.26 (C)

I Liquid Effluents (Adults)

Total Body 'rgan Discharge Fish point Consumption 0. 0064 0.059 (Thyroid)

Invertebrate 0.0023 0. 071 (Thyroid)

Shoreline 0.00058 0.00058 (Thyroid)

"Nearest" refers to that site boundary location where the highest radiation doses as a result of gaseous effluents have been estimated to occur.

Doses are for the age group. and organ that results in the highest cumulative dose for the location: T=teen, C=chi ld, I=infant. Calculations were made for these age groups and for the following organs: GI-tract, bone, liver, kidney, thyroid, lung, and skin.

This location includes the assumption that a garden could be located here.

"Nearest" refers to the location where the highest radiation dose to an individual from all applicable pathways has been estimated.

St. Lucie 2 DES F-8

Table F-4.c Combined Ahnual Dose Commitments to a Maximally Exposed Indiv'idual from St. Lucie Plant Location Pathway Doses (mrem/yr per site)

Noble Gases in Gaseous Effluents Total Body Skin Gamma Air Dose Beta Air Dose (mrad/yr per (mrad/yr per Nearest ~ite Direct radiation site) site) boundary from plume 0. 65 1.8 1.1 3.0 (NW, 0.97 mi)

Iodine and Particulates in Gaseous Effluents Maximum Organ Total Body Thyroid Bone Nearest site Ground deposition O.113 (T) o.113 (c) 0.113 (c) boundary Inhalation 0.057 (T) o'.3o (c) o'.oo2 (c)

(NW, 0.97 mi)

Nearest Ground deposition o.o5 (c) o.o5 (c) o.o5 (c)

.resident Inhalation o.'o2 (c) o.12 (c) 0.00082 (C)

(SE, 1.4 mi) Vegetable consumption o'. 16 (c) 0.75 (C) 0.79 (C)

Nearest Ground deposition 0.03 (C) o.o3 (c) o.o3 (c) garden Inhalation 0.01 (C) 0.07 (C) o.'ooo49 (c)

(W, 1.9 mi) .Vegetable consumption 0.'11 (c) 0.43 (C) 0.52 (C)

Liquid Effluents (Adults)

Total Body Organ Discharge Fish point Consumption 0. 012 0.099 (Thyroid)

Invertebrate 0. 0041 0.12 (Thyroid)

Shoreline 0. 0011 0. 0011 (Thyroid)

"Nearest" refers to that site boundary location where the highest radiation doses as a result of gaseous effluents have been estimated to occur.

Doses are for the age group and organ that results in the highest cumulative dose for the location: T=teen, C=child, I=infant. Calculations were, made for these age gr'oups and for the following organs: GI-tract, bone, liver, kidney, thyroid, lung, and skin.

"Nearest" refers to the location where the highest radiation dose to an individual from all applicable pathways has been estimated.

St. Lucie 2 DES F-9

Table F-5 Calculated Appendix I Dose, Commitments to a Maximally Exposed Individual and to the Population from Operation of St. Lucie Plant Unit 2 Annual Dose per Reactor Unit Individual Appendix I Calculat'ed Design Objectives Doses Liquid effluents Dose to total body from all pathways 3 mrem 0.01 mrem Dose to any organ from all pathways 10 mrem 0.13 mrem Noble-gas effluents (at site boundary)

Gamma dose in air 10 mrad 0.67 mrad in air

'0 Beta dose mrad 1.8 mrad Dose to total body of an individual 5 mrem 0.40 mrem Dose to'skin of an individual 15 mrem 1.1 mrem Radioiodines and particulates 4 Dose to any organ from all pathways 15 mrem 0.57 mrem Population Within 80 km Total Body Thyroid 7

(person-rem)

Natural-background radiation 100,000 Liquid effluents '.'4

15. 0 Noble-gas effluents 0.'5 0. 5 Radioiodine and particulates 0.57 2.7 Design Objectives, from Sections II.A, II~ B, II.C, and II.D of Appendix I, 10 CFR Part 50 consider doses to maximally exposed individual and population per reactor unit.

Numeric values in this column are from summing appropriate values in Table F-4.b.

Locations resulting in the maximum dose are presented here.

Carbon-14 and tritium have been added to this category.

"Natural Radiation Exposure in the United States," U.S. Environmental Protection Agency, ORP-SID-72-1, June 1972; using. the average background dose for Florida of 83 mrem/yr, and year-2000 projected population of 1,233,051.

St. Lucie 2 DES F-10

Table F-6 Calculated RM-50-2 Dose Commitments t~ a Maximally Exposed Individual from Operation of the St. Lucie Plant Annual Dose per Site RM-50-2 b

Calculated Design Objectives Doses Liquid effluents Dose to total body or any organ from all pathways 5 mrem 0.22 mrem Activity-release estimate, excluding tritium (Ci) 10 0. 45 Noble-gas effluents (at site boundary)

Gamma dose in air 10 mrad 1.1 mrad Beta dose in air 20 mrad 3.0 mrad Dose to total body of ari individual 5 mrem 0.65 mrem Radioiodine and particulates Dose to any organ from all pathways 15 mrem 0.57 mrem I-131 activity release (Ci) 2 '0. 23 An optional method of demonstrating compliance with the cost-benefit Section (II.D) of Appendix I to 10 CFR Part 50.

b Annex to Appendix I to 10 CFR Part 50.

Carbon-14 and tritium have been added to this category.

St. Lucie 2 DES F-ll

Table F-7 Annual Total-Body Population Dose Commitments, Year 2000 from St. Lucie Plant U.S. Population Category Dose Commitment, person-rem/yr Natural background radiation 26,000,000 St. Lucie Plant Units 1 and 2 (combined) operation Plant workers 880 General public:

Liquid effluents 5.7 Gaseous effluents 79 Transportation of fuel and waste 14 Using the average U.S. background dose (100 mrem/yr) and year 2000 projected U.S. population from "Population Estimates and Projections," Series II, U.S. Department of Commerce, Bureau, of the, Census, Series P-25, No. 541, February 1975.

80-km (50-mile) population dose St. Lucie 2 DES F-12

Table F-8.a Calculated Releases of Radioactive Materials in Liquid Effluents from St. Lucie Plant Unit 1 Nuclide Ci/yr Nuclide Ci/yr Corrosion and Activation Products Fission products (continued)

Cr-51 0. 00009 Te-129 0. 00011 Mn-54 0. 001 I-130 0.00033 Fe-55 0. 00009 Te-131m 0.00006 Fe-59 0. 00006 Te-131 0.00007 Co-58 0. 0049 I-131 0.1 Co-60 0.0088 Te-132 0.00081 Np-239 0.00003 I-132 0. 016 I"133 0. 078 Fission Products I-134 0. 00082 Cs-'134 0. 027 Br-83 0. 00021 I-135 0. 02 Br-84 0.00003 Cs-136 0. 0042 Rb-86 0.00003 Cs-137 0. 034 Rb-88 0. 0011 Ba-137m 0. 0061 Sr-89 0. 00002 La-140 0. 00001 Mo-99 '.003 All Others 0.00006 Tc-99m 0.0044 Total except '.

Te-127m 0. 00001 Tritium 32 Te-127 0. 00003 Tritium Te-129m 0. 0000? Release 470 Nuclides whose release rates are less than 10-s Ci/yr are not listed individually but are included in the category "All Others."

St. Lucie 2 DES F-13

Table,F-8. b Calculated Release of Radioactive Materials in Liquid Effluents from St. Lucie Plant Unit 2.

Nuclide Ci/yr- Nuclide Ci/yr Corrosion and Activation Products. Fission products (continued)

Cr-51 .00013 Te 132 .0013 Mn-54 .001 I-132 .023 Fe-55 .00011 I-133 ,.12 Fe-59 .00007 I"134 .0012 Co-58 .0051 Cs-134 .031 Co-60 .0088 I"135 ,.028

'r-95

.0014 Cs"136 .0072 Nb-95 .003 Cs-137 .036 NP-239 .00005 Ba-137m .0064 Ba-140 .00001 Fission Products La-140 .00002 Ce-144 .0052 Br-83 .00031 Br-84 .00004 All others .00005 Rb-86 .00005 Total Rb-88 .0016 (except H-3) .45 Sr-89 .00003 H-3 510 Sr-91 .00001 Mo-99 .0046 Tc-99m .0064 a=nuclides whose release rates Ru-103 .00014 are less than 10- Ci/yr are Rb"106 .0024 not listed individually but Ag-110m .00044 are included in the category Te-127m .00002 "all others."

Te-127 .00004 Te-129m .00009 Te-129 .00014 I-130 .00048 Te-131m .00009 Te-131 .0001 I-131 .15 St. Lucie 2 DES F-14 .,

Table F-9 Summary of Hydrologic Transport and Dispersion for Liquid Releases from the St. Lucie Plant Unit 2 Transit Time Dilution Location (hours) Factor ALARA Calculations I

Sport fishing (discharge canal) 0.5 1.0 Population Dose Calculations I

Commercial fishing (discharge canal) 0.5 1.0 See Regulatory Guide 1. 113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

St. Lucie 2 DES F-15 '

'l i

i I

1

'1

APPENDIX G REBASELINING OF THE REACTOR SAFETY STUDY RESULTS FOR PMRs St. Lupine 2 DES G-1

The results of the Reactor Safety Study (RSS) have been updated. The update was done largely to incorporate results of research and development conducted after the October 1975 publication of the RSS and to provide a baseline against which the risk associated with various LWRs could be consistently compared.

Primarily, the rebaselined RSS results reflect use of advanced modeling of the processes involved in meltdown accidents, i.e., the MARCH computer code model-ing for transient and LOCA initiated sequences of and the CORRAL code used for calculating magnitudes of release accompanying various accident sequences.

These codes* have led to a capability to predict the transient and small LOCA initiated sequences that is considerably advanced beyond what existed at the time the Reactor Safety Study was completed. The advanced accident process models (MARCH and CORRAL) produced some changes in our estimates of the release magnitudes for various accident sequences in MASH-1400. These changes primarily involved release magnitudes for the iodine, cesium, and tellurium fami lies of isotopes. In general, a decrease in the iodines was predicted for many of the dominant accident sequences while some increases in the release magnitudes for the cesium and tellurium isotopes were predicted.

Entailed in this rebaselining effort was the evaluation of individual dominant accident sequences as we understand them to evolve rather than the technique of grouping large numbers of accident sequences into encompassing, but synthetic release categories as was done in WASH-1400. The rebaselining of the RSS also eliminated the "smoothing technique" that was criticized in the report by the Risk Assessment Review Group (sometimes known as the Lewis Report, NUREG/CR-0400).

In both of the RSS designs (PMR and BMR), the likelihood of an accident'sequence leading to the occurrence of a steam explosion (a) in the reactor vessel was decreased. This was done to reflect both experimental and calculational indica-tions that such explosions are unlikely to occur in those sequences involving small size LOCAs and transients because of the high pressures and temperatures expected .to exist within the reactor coolant system during these scenarios.

Furthermore, if such an explosion were to occur, there are indications that would be unlikely to produce as much energy and the massive missile-caused breach it of containment as was postulated in WASH-1400.

For rebaselining of the RSS-PWR design, the release magnitudes for the risk dominating sequences, e.g., Event tj', TMLB; 6, y, and SyC-6 (described later) were explicitly calculated and used in the consequence 'modeling rather than being lumped into release categories as was done in MASH-1400. The rebaselin-ing led to a small decrease in the predicted risk to an individual of early fatality or latent cancer fatality relative to the original RSS-PWR predictions.

This result is believed to be largely attributable to the decreased likelihood of occurrence for sequences involving severe steam explosions that breached containment. In WASH-1400, the sequences involving severe steam explosions were artificially elevated in their risk significance (i.e., made more likely) by use of the "smoothing technique."

t It should be noted that the MARCH code was used on a number of scenarios in connection with the TMI-2 recovery efforts and for post-TMI-2 investigations to explore possible alternative scenarios that TMI-2 could have experienced.

St. Lucie 2 DES G-2

In summary, the rebaselining of the RSB results led to small overall differences from the predictions in WASH-1400. It should be recognized that these small differences due to the rebaselining efforts are likely to be far outweighed by the uncertainties associated with such analyses.

The accident sequences which are expected to dominate risk from the RSS-PWR design are described below.- These sequences are assumed to represent the approximate accident risks from the St. Lucie PWR design. Accident sequences are designated by strings of identification symbols in the same manner as in the RSS. A key to these symbols is at the end of this Appendix. Each of the characters represents a failure in one or more of the important plant systems or features that ultimately would result in melting of the reactor core and a significant release of radioactive materials from containment."

Event Y (Interfacin S stem LOCA)

During the Reactor Safety Study a potentially large risk contributor was identi-fied due to the configuration of the multiple check valve bar riers used to separate the high pressure reactor coolant system from the low design pressure portions of the ECCS (i.e., the low pressure injection subsystem--LPIS). If these valve bariers were to fail in various modes, such as leak-rupture or rupture-rupture, and suddenly exposed the LPIS to high overpressures and dynamic loadings, the RSS judged that a high probability of LPIS rupture would exist.

Since the LPIS is largely located outside of containment, the Event V scenario would be LOCA that bypassed containment and those mitigating features (e.g.,

sprays) within containment. The RSS assumed that if the rupture of LPIS did not entirely fail the LPIS makeup function (which would ultimately be needed to prevent core damage), the LOCA environment (flooding, steam) would. Predic-tions of, the release magnitude and consequences associated with Event V have indicated that this scenario represents one of the largest risk contributors from the RSS-PWR design. The NRC has recognized this RSS finding, and has taken steps to reduce the probability of occurrence of Event V scenarios in both existing and future LWR designs by requiring periodic surveillance testing of the interfacing valves to assure that these valves are properly functioning as pressure boundary isolation barriers during plant operations. Accordingly, Event V predictions for the RSS-PWR are likely to be conservative relative to the design and operation of the St. Lucie 2 PWR.

~INLB '5 This sequence essentially considers the loss and nonrestoration of all AC power sources available to the plant along with an independent failure of the steam turbine-driven auxiliary feedwater train which would be required to operate to remove shutdown heat from the reactor core. The transient event is initiated by loss of offsite AC power sources which would result in plant trip (scram) and the loss of the normal way that the plant removes heat from the reactor core (i.e., via the power conversion system consisting of the turbine, condenser, the condenser cooling system, and the main feedwater and condensate delivery system that supplies water to the steam generators. This initiating event would then demand operation of the standby onsite emergency AC power supplies For additional information detail see Appendix V of "Reactor Safety Study,"

WASH-1400, NUREG-75/014, October 1975.

St. Lucie 2 DES G-3

(2 diesel generators) and the standby auxiliary feedwater sytem, two trains of which are electrically driven by either onsite or offsite AC power. With failure and nonrestoration of AC and the failure of the steam turbine-driven auxiliary feedwater train to remove shutdown heat, the core would ultimately uncover and melt. If restoration of AC was not successful during (or following) melt, the containment heat removal and fission produce mitigating systems would not be operational to prevent the ultimate overpressure (6, y) failure of containment and a rather large, energetic release of activity from the containment. Next to the Event V sequence, TMLB'6, y is predicted to dominate the overall accident risks in the RSS-PWR design.

SyC-5 (PWR 3)

In the RSS the SqC-6 sequence was placed into PWR release Category 3 and it actually dominated all other sequences in Category 3 in t'erms of pr'obability and release magnitudes. The rebaselining entailed explicit calculations of the consequences from SqC-6 and the results indicated that it was next in overall risk importance following Event V and TMLB'6, y.

The SyC-6 sequence included a rather complex series of dependencies and inter-actions that are believed to be-somewhat unique to the containment systems (subatmospheric) employed in the RSS-PWR design.

In essence, the SyC-6 sequence included a small LOCA occurring in a of the plant (reactor vessel cavity); failure of the recirculating specific'egion containment heat removal systems (CSRS-F) because of a dependence on water draining to the recirculation sump from the LOCA and a resulting depe'ndence imposed on the quench spray injection system (CSIS-C) to'provide water to the sump. The failure of. the CSIS(C) resulted in eventual overpressure fail'ure of containment (6) due to the loss of CSRS(F). Given the overpressure fai,lure of containment the RSS assumed that the ECCS functions would be lost due either to the cavitation of ECCS pumps or from the rather severe mechanical loads that could result from the overpressure failure of containment. The core was then assumed to melt in a breached containment leading to a significant release of radioactive materials.

Approximately 20K of the iodines and 20K of the alkali metals present in the core at the time of release would be released to the atmosphere. Most. of the release would occur over a period of about 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The release of radio-active material from containment would be caused by the sweeping action'f gases generated by the reaction of the molten fuel with concrete. Since these gases would be initially heated by contact wth the melt, the rate of sensible energy release to the atmosphere would be moderately high.

PWR 7 This is the same as the PWR release Category 7 of the or~iginal RSS which was made up of several sequences such as SqD-c (the dominant'contributor to the risk in this category), SqDc, S 'H-s, SqH-c, AD-c, AH-c,TML-c, and TKg-c. All of these sequences involved a containment basemat melg-through as the:containment failure mode. With exception of. TML-c, and TKg-c all involve the potential fai lure of the emergency core. cooling system following occurrence of' 'base LOCA with the containment ESFs continuing to operate as designed until the mat St. Lucie 2 DES G-4

was penetrated. Containment sprays would operate to reduce the containment temperature and pressure as well as the amount of airborne radioactivity, The containment barrier would retain its integrity until the molten core proceeded to melt through the concrete containment basemat. The radioactive materials would be released into the ground, with some leakage to the atmosphere occurring upward through the ground. Most of the release would occur continuously over a period of about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The release would include approximately 0.002K of the iodines and 0.002K of alkali metals present in the core at the time of release.

Because leakage from containment to the atmosphere would be low and gases escaping through the ground would be cooled by contact with the soil, the energy release rate would be very low.

St. Lucie 2 OES G-5

KEY TO PMR ACCIDENT SE UENCE SYMBOLS,',

A - Intermediate to large LOCA.

B Failure of electric power to ESFs.

B' Failure to recover either onsite or offsite electric power within about 1 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> following an intiating transient which is a loss of offsite AC power.

Failure of the containment spray injection system.

Failure of the emergency core cooling injection 'system.

i Failure of the containment spray recirculation system.

G

- Failure of the containment heat removal system.

- Failure of the emergency core cooling recirculation system.

K - Failure of the reactor protection system.

L - Failure of the secondary system steam relief valves and the auxiliary feedwater system.

Failure of the secondary system steam relief valves and the power conversion system.

Failure of the primary system safety relief valves to reclose after opening.

R

- Massive rupture of the reactor vessel.

Sz

- A small LOCA with an equivalent diameter of about 2 to 6 inches.

Sy A small LOCA with an equivalent diameter of about '1/2 to 2 inches.

T - Transient event.

V - LPIS check valve failure.

e - Containment rupture due to a reactor vessel steam explosion.

Containment failure resulting from inadequate i'solation of containment openings and penetrations.

Containment failure due to hydrogen burning.

Containment failure due to overpressure.

s - Containment vessel melt-through.

St. Lucie 2 DES G-6

APPENDIX H EVACUATION MODEL St. Lucip 2 DES H-1

"Evacuation" used in the context of offsite emergency response in the event of substantial amount of radioactivity release to the atmosphere in a reactor accident, denotes an early and expeditious movement of people to avoid exposure to the passing radioactive cloud and/or to acute ground contamination in the wake of the cloud passage. It should be distinguished from ",relocation" which denotes a post-accident response to reduce exposure from long-term ground contamination. The Reactor Safety Study~ (RSS) consequence model contains provision for incorporating radiological consequence reduction benefits of public evacuation. Benefits of a properly planned and expeditiously carried out public evacuation would be well manifested in reduction of, acute health effects associ-ated with early exposure; namely, in the number of cases of acute fatality and acute radiation sickness which would require hospitalization. The evacuation model originally used in the RSS consequence model is decribed in MASH-1400~

as well as in NUREG-0340.~ However, the evacuation model which has been used herein is a modified version~ for the RSS model and is, to a certain extent, site emergency planning oriented. The modified version is briefly outlined below:

The model utilizes a circular area with a specified radius (such as a 16 km (10 mi) plume exposure pathway Emergency Planning Zone (EPZ)), with the reactur at the center. It is assumed that people living within portions of this area would evacuate if an, accident should occur involving imminent or actual release of significant quantities of radioactivity to the atmosphere.

.Significant atmospheric releases of radioactivity would in general be preceded by one or more hours of warning time (postulated as the time-interval between the awareness of impending core-melt and the beginning of the release of radio-activity from the containment building). For the purpose of calculation of radiological exposure, the model assumes that all people who live in a fanshaped area (fanning out from the reactor) within the circular zone, with the down-wind direction as its centerline--i.e., those people who would potentially be under the radioactive cloud that would develop following the release--would leave their residences after a specified amount of delay time* and then evacuate.

The delay time is reckoned from the beginning of the warning time and is the sum of the time required by the reactor operators to notify the responsible authorities; the time required by the authorities to interpret the data, decide to evacuate, and direct the people to evacuate; and the time required for the people to mobilize and get underway.

The model assumes that each evacuee would move radially outward in the downwind direction with an average effective speed~ (obtained by dividing the zone radius by the average time taken to clear the zone after the delay time) over a fixed distance" from the evacuee's starting point. This distance is selected to be 24 km (15 mi)--which is 8 km (5 mi) more than the 16 km (10 mi) plume exposure pathway EPZ radius. After reaching the end of. the travel distance the evacuee is assumed to receive no further radiation exposure. Persons who are outside the evacuation radius are assumed to remain in place for seven days prior to relocating, unless remaining for that long a period of time would produce a dose greater than 200 rem to the whole body from ground exposure. In that case, relocation takes place after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, with a dose appropriate to that time period.

t Assumed to'be a constant value for all evacuees.

St. Lucie 2 DES H-2

The model incorporates a finite length of the radioactive cloud in the downwind direction which would be determined by the product of the duration over which the atmospheric release would take place and the average windspeed during the release. It is assumed that the front and the'back of the cloud formed would move with an equal speed which would be the same as the prevailing windspeed; therefore, its length would remain constant at its initial value. At any time after the release, the concentration of radioactivity is assumed to be. uniform over the length of the cloud. If the delay time were less than the warning time, then all evacuees would have a head-start, i.e., the cl'oud would be trai ling behind the evacuees initially. On the other hand, if the delay time were more than the warning time, then depending on initial locations of the evacuees there are possibilities that (a) an evacuee will still have a head-start, (b) the cloud would be already overhead when an evacuee starts to leave, or (c) an evacuee would be initially trailing behind the cloud. However, this initial picture of cloud-people disposition would change as the evacuees travel, depending on the relative speed and position between the cloud and the people.

The cloud and an evacuee might overtake one another one or more times before the evacuee would reach his or her destination. In the model, the radial posi-tion of an evacuating person, either stationary or in transit, is compared to the front and the back of the cloud as a function of time to determine a realistic period of exposure to airborne radionuclides, The model calculates the time periods during which people are exposed to radionuclides on the ground while they are stationary and while they are evacuating. Because radionuclides would be deposited continually from the cloud as it passed a given location, a person who is under the cloud would be exposed to ground contamination less concentrated than if the cloud had completely passed. To account for this, at least in part, the revised model assumes that persons are (a) exposed to the total ground contamination concentration which is calculated to exist after complete passage of the cloud after they are completely passed by the cloud, (b} exposed to one half the calculated concentration when anywhere under the cloud, and (c) not exposed when they are in front of the cloud. The model provides for use of different values of the shielding protection factors for exposure due to airborne radioactivity and contaminated ground. Breathing rates for stationary and moving evacuees during delay and transit periods are specifically included.

It is realistic to expect that authorities would evacuate persons at distances from the site where exposures above the threshold for causing acute fatality could occur, r'egardless of the EPZ distance. Figure H. 1 illustrates the reduction in acute fatalities that can occur by extending evacuation to larger distances up to 24 km (15 mi) from the St. Lucie 2 site. (The evacuation distance used in the Reactor Safety Study~ was 40 km (25 mi). Also illustrated in Figure H. 1 is a more pessimistic case for which no, early evaluation is assumed. For this case, all persons within 16 km (10 mi) of the plant are assumed to be exposed for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an accident and are then relocated. Compared to the pessimistic scenario, evacuation of a 24 km (15 mi) zone shows a reduction in acute fatalities of a factor of 8 at the 10-7 probability.

The model has the same provision for calculation of the economic cost associated with implementation of evacuat'ion as in the original RSS model. For this purpose, the model assumes that for atmospheric releases lasting 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less all people living within a circular area of 8 km (5 mi) radius centered at the reactor plus all people within a 45'. angular sector within the plume exposure pathway EPZ and centered on the downwind direction will be evacuated and temporarily relocated. However, for releases exceeding 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, the cost of evacuation is St. Lucie 2 DES H-3

PROBABILITY DISTRIBUTION

'o 10 10 OF ACUTE 10 LEGEND 10, FATALITIES HP, o

< = NO EVAC. RELOCATION AFTER 1 DAY

~ =EVACUATION TO 15 MILES Al CQ go r >'o 0+

E 'o o M

O o o 3.(f 10 10 10 X=ACUTE FATALITIES Figure H-1 Acute Fatalities.

Note: P1ease See Section 5.10.4.1.4.7 for discussion of uncertainties in risk estimates.

based on the assumption that all people within the plume exposure pathway EPZ would be evacuated and temporarily relocate. For either of these situations, the cost of evacuation and relocation is assumed to be $ 125 (1980 dollars) per person which includes cost of food, and temporary sheltering for a period of one week.

References.

1. U.S. Nuclear Regulatory Commission, "Reactor Safety Study," MASH-1400 (NUREG-75/014), October 1975.

2. U.S. Nuclear Regulatory Commission, "Overview of the Reactor Safety Study Consequences Model," NUREG-0340, October 1977.

3. Sandia Laboratory, "A Model of Public Evacuation for Atmospheric Radiological Releases," SAND 78-0092, June 1978.

St. Lucie 2 DES

\

1

APPENDIX I IMPACT. OF THE URANIUM FUEL CYCLE

1 I

I. 1 Introduction The following assessment of the environmental impacts of the fuel cycle as related to the operation of the proposed project is based on the values given in Table S-3 (Table 5. 11, Section S. 11) and the staff's analysis of the radio-logical impact from radon releases. For the sake of consistency, the analysis of fuel cycle impacts has been cast in terms of a model 1000-NWe light-water-cooled reactor (LMR) operating at an annual capacity factor of 80K. In the following review and evaluation of the environmental impacts of, the fuel cycle, the staff's analysis and conclusions would not be altered if the analysis were to be based on the net electrical power output of the St. Lucie Plant.

I-2 Land Use The total annual land requirement for the fuel cycle supporting a model 1000-NMe LWR is about 460,000 m~ (113 acres). Approximately 53,000 m~ (13 acres) per year are permanently committed land, and 405,000 m~ (100 acres) per year are temporarily committed. (A "temporary" land commitment is a commitment for the life of the specific fuel cycle plant, e. g., mill, enrichment plant, or succeeding plants. On abandonment or decommissioning, such land can be used for any purpose. "Permanent" commitments represent land that may not be released for use after plant shutdown and/or decommissioning.) Of the 405,000 m~ per year of temporarily committed land, 320,000 m~ are undisturbed and 90,000 m~

are disturbed. Considering common classes of land use in the United States,*

fuel cycle land use requirements to support the model 1000-NMe LWR do not represent a significant impact.

I-3 Mater Use The principal water use requirement for the fuel cycle supporting a model 1000;NMe LMR is that required to remove waste heat from the power stations supplying electrical energy to the enrichment step of this cycle. Of the total annual requirement of 43 x 10 m (11.4 x 10~ gal), about 42 x 10 m~ are required for this purpose, assuming that these plants use once-through cooling.

Other water uses involve the discharge to air (e. g., evaporation losses in process coo]ing) of about 0.6 x 10 m~ (16 x 10? gal) per year and water discharged to the ground (e. g., mine drainage) of about 0.5 x 10~ m~ per year.

On a thermal effluent basis, annual discharges from the nuclear fuel cycle are about 4X of the model 1000-NWe LMR using once-through cooling. The consumptive water use of 0.6 x 10~ ms per year is about 2%%u'f the model 1000-NWe LWR using cooling towers. The maximum consumptive water use (assuming that all plants supplying electrical energy to the nuclear fuel cycle used cooling towers) would be about GX of the model 1000-NWe LWR using cooling towers. Under this condition, thermal effluents would be negligible. The staff finds that these combinations .

of thermal loadings and water consumption are acceptable relative to the water use and thermal discharges of the Plant,.

A coal-fired plant of 1000-NWe capacity using strip-mined coal requires the disturbance of about 810,000 m~ (200 acres) per year for fuel alone.

St. Lucie 2 DES

I-4 Fossil Fuel Consum tion Electrical energy and process heat are required during various phases of the fuel cycle process. The electrical energy is usually produced by the combustion of fossil fuel at conventional power plants. Electrical energy associated with the fuel cycle represents about 5X of the annual electrical power production of the model 1000-MWe LMR. Process heat is primarily generated by the combus-tion of natural gas. This gas consumption, if used to generate electricity, would be less than 0.3X of the electrical output from the model plant. The staff finds that the direct and indirect consumptions of electrical energy for fuel cycle operations are small and acceptable relative to the net power produc-tion of the Plant.

I-5 Chemical Effluents The quantities of chemical, gaseous, and particulate effluents associated with fuel cycle processes are given in Table S-3 (Table 5. 11). The principal species are SO NO , and the particulates. Judging from data in a Council on Environmental gualitg ressort,~ the NRC staff finds that these emissions constitute an extremely small additional atmospheric loading in comparison with emissions from the stationery fuel-combustion and transportation sectors in the United States, that is, about 0.02K of the annual national releases for each of these species.

The staff believes such small increases in releases of these pollutants are acceptable.

Liquid chemical effluents produced in fuel cycle processes are related to fuel enrichment, fabrication, and reprocessing operations and may be released to receiving waters. These effluents are usually present in dilute concentrations such that only small amounts of dilution water are required 'to reach levels of concentration that are within established standards. Table S-3 (Table 5. 11) specifies the flow of dilution water required for specific constituents.

Additional]y, all liquid discharges into the navigable waters of the United States from plants associated with the fuel cycle operations will be subject to requirements and limitations set forth in the NPDES permit. (Draft NPDES permit for St. Lucie 2 can be found in Appendix C.)

Tailings solutions and solids are generated during the milling process. These solutions and solids are not released in quantities sufficient to have a signif-icant impact on the environment.

I-6 Radioactive Effluents Radioactive effluents estimated to be released to the environment from reproces-sing and waste management activities and certain other phases of the fuel cycle process are set forth in Table S-3 (Table 5. 11). Using these data, the staff has calculated the 100-year involuntary environmental dose commitment" to the U.S. population.

The environmental dose commitment (EDC) is the integrated population dose for 100 years; that is, it represents the sum of the annual 'population doses for it is not a total of 100 years. The population dose varies with time, and practical to calculate this dose for every year.

St. Lucie 2 DES I-2

It is estimated from these calculations that the overall involuntary total-body gaseous dose commitment to the U. S. population from the fuel cycle (excluding reactor releases and the dose commitment due to radon-222 and technetium-99}

would be approximately 400 person-rems for each year of operation of the model 1000-MWe LWR (reference reactor year, or, RRY). Based on Table S-3 (Table 5. 11) values, the additional involuntary total body-dose commitments to the U.S.

population from radioactive liquid effluents (excluding technetium-99) due to all fuel cycle operations other than reactor operation would be approximately 100 person-rels per year of operation. Thus the estimated involuntary 100-year environmental dose commitment to the U.S. population from radioactive gaseous and liquid releases due to these portions of the fuel cycle is approximately 500 person-rems (whole-body) per RRY.

At this time Table S-3 (Table 5. 11) does not address the radiological impacts associated with radon-222 and technetium-99 releases. Principal radon releases occur during mining and milling operations and as emissions from mill tailings, whereas principal technetium-99 releases occur from gaseous diffusion enrichment facilities. The staff. has determined that radon-222 releases from these opera-tions for each year of operation of the model 1000-MMe LWR are as given in Table I-1. The staff has calculated population dose commitments for these sources of radon-222 using the RABGAD computer code described in Appendix A of Vol. 3, Chap. IV, Sec. J, of NUREG-0002.~ The results of these calculations for mining and milling activities prior to tailings stabilization are listed in Table I-2.

When added to the 500 person-rems total-body dose commitment for the balance of the fuel cycle, the overall estimated total-body involuntary 100-year environ-mental dose commitment to the U.S. population from the fuel cycle for the model 1000-MWe LWR is approximately 640 person-rems. Over this period of time, this dose is equivalent to 0.00002X of the natural-background total body dose of about 3 billion person-rems to the U.S. population."

The staff has considered the health effects associated with the releases of radon-222, including both the short-term effects of mining and milling, and active tailings, and the potential long-term effects from unreclaimed open-pit mines and stabilized tailings. 'he staff has assumed that after completion of active mining underground mines wi 11 be sealed, returning releases of radon-222 to background levels. For purposes of providing an upper-bound impact assess-ment, the staff has assumed that open-pit mines will be unreclaimed and has calculated that if all ore were produced from open pit mines, releases from them would be 110 Ci per RRY. However, because the distribution of uranium ore reserves available by conventional mining methods is 66X underground and 34K open pit,~ the staff has further assumed that uranium to fuel LMRs will be produced by conventional mining methods in these proportions. This means that long-term releases from unreclaimed open-pit mines will be 0.34 x .110 or 37 Ci per year per RRY.

Based on an annual average natural background individual dose commitment of 100 mi llirems and a stabilized U. S. population of 300 million.

St. Lucie 2 DES I-3

r Table I-1 Radon releases from mining and milling operations and mill tailings for each year of operation of, the model 1000-MWe LWR" Radon Source guantity Released Mining 4060 Ci Milling and tailings (during active mining) 780 Ci Inactive tailings (prior to stabilization) 350 Ci Stabilized tailings (several hundred years) 1 to 10 Ci%ear 1

Stabilized tailings (after I several hundred years) 110 Ci/year R. Wilde, U.S. Nuclear Regulatory Commission trans'cript of direct testimony given "In the Matter of Duke Power Company Company (Perkins Nuclear Station)," Docket No. 50-488, April 17, 1978.

P. Magno, U.S. Nuclear Regulatory Commission transcript of direct testimony given "In the Matter of Duke Power Company I

(Perkins Nuclear Station)," Docket No. 50-488, April 17, 1978.

'"After three days of hearings before the Atomic Safety and Licensing Appeal Board (ASLAB) using the Perkins record in a "lead case" approach, the ASLAB issued a decrsron Nay 13, 1981 (ALAS-640) on the radon-222 release source term for the Uranium Fuel Cycle. The decision, among other matters, produced new source term numbers based on the record developed at the hearings. These new numbers did not differ significantly from those in the Perkins record which are the values set forth in this Table. Any health effects relative to radon-222 are still under consideration before the ASLAB. Since the source term numbers in ALAB-640 do not differ significantly from those in the Perkins record, the staff continues to conclude that "both the dose commitments and health effects of the uranium fuel cycle are insignificant when compared to dose commitments and potential health effects to the U. S. population resulting from all natural background sources." (see page 1-7)

St. Lucie 2 DES

Table I-2 Estimated 100-year environmental dose commitment per year of operation of the model 1000-NMe LMR Radon-222 Dosa e ( erson-rems Radon Source Releases (Ci) otal Body Bone Lung Bronc ia epithelium)

Mining 4100 110 2800 2300 Milling and active tailin s 1100 29 750 620 Total 140 3600 2900 Based on the above, the radon released from unreclaimed open-pit mines over 100- and 1000-year per iods would be about 3700 Ci and 37,000 Ci per RRY respectively. The total dose commitments for a 100 to 1000-year period would be as follows:

Radon-222 Po ulation dose commitments ( erson-rems)

Time span (years) Releases (Ci) Total Bone Lung (bronchial body epithelium) 100 3,700 96 2,500 20000 500 19,000 480 13,000 11,000 1,000 37,000 960 25,000 20,000 environmental dose commitment to the U.S. population from the fuel cycle for the model 1000-NMe LMR is approximately 640 person-rems. Over this period of time, this dose is equivalent to 0.00002K of the natural-background total body dose of about 3 billion person-rems to the U. S. population.

• The staff has considered the health effects associated with the releases of radon-222, including both the short-term effects of mining and milling, and active tailings, and the potential long-term effects from unreclaimed open-pit mines and stabilized tailings. The staff has assumed that after completion of active mining underground mines will be sealed, returning releases of radon-222 to background levels. For purposes of providing an upper-bound impact assess-ment, the staff has assumed that open-pit mines will be unreclaimed and has calculated that if all ore were produced from open pit mines, releases from them would be 110 Ci per RRY. However, because the distribution of uranium ore reserves available by conventional mining methods is 66K underground and 34X open pit,~ the staff has fur ther assumed that uranium to fuel LMRs will be produced by conventional mining methods in these proportions. This means that long-term releases from unreclaimed open-pit mines will be 0.34 x 110 or 37 Ci per year per RRY.

Based on an annual average natural background individual dose commitment of 100 millirems and a stabilized U.S. population of 300 million.

St. Lucie 2 DES I-5

Based on the above, the radon released from unreclaimed open-pit mines over 100- and 1000-year periods would be about 3700 Ci and 37,000 Ci per RRY respectively. The total dose commitments for a 100 to 1000-year period would be as follows:

Radon-222 Po ulation dose commitments ( erson-rems)

Time span (years) Releases (Ci) Total -Bone Lung (bronchial body epithelium) 100 3,700 96 2,500,'3,000 2,000 500 19,000 480 11,000 1,000 37,000 960 '5,000 20,000 The above dose commitments represent a worst-case situation in that no mitigating circumstances are assumed. However, Federal and State laws currently require reclamation of strip and open-pit coal mines, and it is very probable that similar reclamation will'be required for uranium open-pit mines. If so, long-term releases from such mines should approach background levels.

For long-term radon releases from stabilized tailings piles, the staff has assumed that these tailings would emit, per RRY, 1 Ci per year for 100 years, 10 Ci per year for the next 400 years and 100 Ci per year for periods beyond 500 years. With these assumptions, the cumulative radon-222 release from stabilized tailings piles per RRY would be 100 Ci in 100 years and 4090 Ci in 500 years and 53,800 Ci in 1000 years.4 The total-body, bone, and bronchial epithelium dose commitments for these periods are as follows:

Radon-222 Po ulation dose commitments ( erson-rems Time span (years) Releases (Ci) Total Bone Lung (bronchial body epithelium) 100 100 2.6 68 56 500 4,090 110 2,800 2,300 1,000 53,800 1,400 37,000 30,000 If risk estimators of 135, 6.9, and 22 cancer deaths per million person-rems for total-,body, bone, and:lung exposures, respectively, are used, the estimated risk of cancer mortality resulting from mining, milling, and active tai lings emissions of radon-222 is about 0. 11 cancer fatalities per RRY. When this risk from radon-222 emissions from stabilized tailings over a 100-year release period is added, the estimated risk of cancer mort'ality over a 100-year period is unchanged. Similarly, a risk of about 1.2'cancer fatalitie" is estimated over a 1000-year release period per RRY. When potential radon releases from reclaimed and unreclaimed open-'pit mines are included, the overall risks of radon induced cancer fatalities per RRY range as follows: 0.11 to 0. 19 fatalities for a 100-year period, 0. 19 .to 0.57 fatalities for a 500-year period, and 1."2 to 2.0 fatalities for a 1000-year period.

To illustrate: A single-model 1000-NWe LWR operating at an 80K capacity factor for 30 years would be predicted to induce between 3.3 and 5.7 cancer St. Lucie 2 DES I-6

fatalities in 100 yr, 5.7 and 17 in 500 yr, and 36 and 60 in 1000 yr as a result of releases of radon-222.

These doses and predicted health effects have been compared with those that can be expected from natural-background emissions of radon-222. Calculated using data from the National Council on Radiation Protection (NCRP)s the average radon-222 concentration in air in the contiguous United States is about 150 pCi/ms, which the NCRP estimates will result in an annual dose to the bronchial epithelium of 450 millirems. For a stabilized future U.S. population of,300 million, this represents a total lung dose commitment of 135 million person-rems per year.

If(the same risk estimator of 22 lung cancer fatalities per million person-lung-rems used to predict cancer fatalities for the model 1000 NWe LWR is used, estimated lung cancer fatalities alone from background radon-222 in the air can be calculated to be about 3000 per year, or 300,000 to 3,000,000 lung cancer deaths over periods of 100 to 1000 years respectively.

The staff is currently in the process of formulating a specific model for analyzing the potential impact and health effects from the release of technetium-99 during the fuel cycle. However, for the interim period until the model is completed, the staff has calculated that the potential 100-year environmental dose commitment to the U.S. population from the release of Tc-99 should not exceed 100 person-rems per RRY. These calculations are based on the gaseous and the hydrological pathway model systems described in NUREG-0002, Vol. 3, Chap. IV, Sec. J, Appendix A.~ When added to the 640 person-rem total-body dose commitment for the balance of the fuel cycle, including radon-222, the overall estimated total-body involuntary 100-year environmental dose commit-ment to the U.S. population from the fuel cycle for the model 1000-NWe LWR is about 740 person-rems. Over this period of time, this dose is equivalent to 0.00002%%uo of the natural-background total-body dose of about three billion person-rems to the U.S. population."

The staff also considered the potential health effects associated with this release of technetium-99. Using the modeling systems described in NUREG-0002, the major risks from Tc-99 are from exposure, of the GI tract and kidney, although there is a small risk from total-body exposure. Using organ specific risk estimators, these individual organ risks can be converted to total-body risk equivalent doses. Then, by using the total-body risk estimator of 135 cancer deaths per million person-rem, the estimated risk of cancer mortality due to technetium-99 releases from the nuclear fuel cycle is about 0.01 cancer fatality per RRY over the subsequent 100 to 1000 years.

In addition to the radon and technetium related potential health effects from the fuel cycle, other nuclides produced in the cycle, such as carbon-l4, will contribute to population exposures. It is estimated that 0. 08 to 0. 12 addi-tional cancer deaths may occur per RRY (assuming that no cure or prevention of, cancer is ever developed) over the next 100 to 1000 years, respectively, from exposures to these other nuclides.

Tlie latter exposures can also be compared with those from naturally occurring terrestrial and cosmic-ray sources. These average about 100 millirems.

Therefore, for a stable future population of 300 million persons, the whole-body Based on an annual average natural-background individual dose commitment of 100 mrems and a stabli lized U. S. population of 300 million.

St. Lucie 2 DES I-7

dose commitment would be about 30 million person-rems per year, or 3 billion person-rems and 30 billion person-rems for periods of 100 and 1000 years respectively. These natural-background dose commitments could produce about 400,000 and 4,000,000 cancer deaths during the same time periods. From the above analysis, the staff concludes that both the dose commitments and health effects of the uranium fuel cycle are insignificant when compared to dose commitments and potential health effects to the U.S. population resulting from all'natural-background sources.

I-7 Radioactive Mastes The quantities of buried radioactive waste material (low-level, high-level, and transuranic wastes) associated with the uranium fuel cycle are specified in Table S-3 (Table 5. 11). For low-level waste disposal at land burial faci-lities, the Commission notes in Table S-3 (Table 5. 11) thatCommission there will be no significant radioactive releases to the environment. The notes that high-level and transuranic wastes are to be buried at a Federal Repository and that no release to the environment is associated with such disposal. NUREG-0116,6 which provides background and context for the high-level and transuranic Table S-3 (Table 5. 11) values established by the Commission, indicates that these high-level and transuranic wastes will be buried and will not be released to the biosphere. No radiological environmental impact is anticipated from such disposal.

I-8 Occu ational Dose The annual occupational dose attributable to all phases of the fuel cycle for the model 1000-NWe LMR is about 200 person-rems. The staff concludes that this occupational dose will not have a significant environmental impact.

99 ~T The transportation dose to workers and the public is specified in Table S-3 (Table 5. 11}. 'his dose is small and is not considered significant in comparison to the natural-background dose.

1-10 ~Fue1 0 cle The staff's analysis of the uranium fuel cycle did not depend on the selected fuel cycle (no recycle or uranium-only recycle), because the data provided in Table S-3 (Table 5.11) include maximum recycle option impact for each element

'of the fuel cycle. Thus the staff's conclusions as to acceptability of the environmental impacts of the fuel cycle are not affected by the specific fuel cycle selected.

I-ll References

1. t Council on Enviornmental guality, "The Seventh Annual Reopr of the Council on Environmental guality." September 1976, Figs. 11-27 and 11-28, pp.

238-239.

9

2. U.S. Nuclear Regulatory Commission, "Final Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel fin Light-Mater-Cooled Reactors." NUREG-0002, August 1976.

St. Lucie 2 DES I-8

3. U.S. Department of Energy, "Statistical Data of. the Uranium Industry."

GJ0-100(8-78), 1 January 1978.

4 R. Gotchy, U.S. Nucle'ar Regulatory Commission, transcript of direct testimony given "In the Matter of Duke Power Company (Perkins Nuclear Station)." Docket No. 50-488, 17 April 1978.

5. National Council on Radiation Protection and Measurements, "Natural Background Radiation in the United States." Publication No. 45, November 1975.

U.S. Nuclear Regulatory Commission, "Environmental Survey of the Reprocessing and Waste Management. Portions of the LWR Fuel Cycle." NUREG-0116 (Supplement 1 to'ASH-1248), October 1976.

St. Lucie 2 DES I-9

I NRC FORM 335 1. REPORT NUMBER /Assi/N)ed by DDCJ 17 77) U.S. NUCLEAR REGULATORY COMMISSION BI B L IOG RAP HI C DATA SHE ET 4, TITLE AND SUBTITLE (Add Volume lVo.,ifappropriate/ 2. (Leave blank J Draft Environmental Statement Related to the Operation of St tucie Plant, Unit No. 2 3. RECIPIENT'S ACCESSION No.

7. AUTHOR IS)

5. DATE REPORT COMPLETED MONTH YEAR October 1981 9.,PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code/ DATE REPORT ISSUED

.~ U.S. Nuclear Regulatory. Commission MONTH YEAR

," Office of Nuclear Reactor Regulation October 1 81 Washington, D.CD 20555 6. /Leave blank/

B. (Leave blankJ

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (/ne/ude Zip Code/

10. PROJECT/TASK/WORK UNIT No.

Same as 9. above 11. CONTRACT No.

)3 TYPE OF REPORT PE RIOD COVE RED /Inclusive dates/

Draft Environmental Statement

)5. SUPPLEMENTARY NOTES 14. /Leave blankJ Pertains to Docket No. 50-389

)6. ABSTRACT (200 words or less/

h The information in this statement is the second assessment of the environmental impact associated with the construction and operation of the St. tucie Plant, Unit No. 2, located in St. tucie County, Florida. The first assessment was the Final Environmental Statement, related to construction, issued in May 197/4. prior to issuance of the St. Lucie 2 construction permit. The St. tucie 2 plant construction is now 80pfo complete and startup is scheduled is the result of the NRC staff review of the for October 1982. The present assessment.

activities associated with the proposed.

operation of the plant.

I 1 KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS

) 71), IDENTIFIERS/OPEN ENDED TERMS

)9 AVAILABII.ITYSTATEMENT 19. SECURITY CLASS /Th)s report/ 21. No. OF PAGES Unclassified Unlimited 20. SECURITY CLASS (7'his page/ 22. PRICE Uncl ssified S N)lC FORM 335 I7 771 de, GOVERNMENT PRINTING OFFICEl 1911 Sdl 1st/IS92 1 2