Text

Forwards Draft Responses & Environ Rept - OL Stage Page Changes,Per Requests for Addl Info E450.1,E450.2,E450.3 & E450.4,in Response to NRC Request.Responses Will Be Incorporated Into Environ Rept - OL by End of Apr 1983
	ML20072M690
Person / Time
Site:	Limerick
Issue date:	03/28/1983
From:	Bradley E; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Schwencer A; Office of Nuclear Reactor Regulation
References
	NUDOCS 8304010347
	Download: ML20072M690 (65)
	v • d • e

,

s PHILADELPHIA ELECTRIC COMPANY 2301 M ARKET GTREET P.O. BOX 8699 EDWARD o. BAUER JR.

viCE PREssoENT AND GENERAL COUNSEL (2 j $) 84 j .40QQ EUGENE J. DRADLEY ASSOCi ATE GENERAL COUNGEL DON ALD BLANKEN RUDOLPH A. CHILLEMI March 28, 1983 E C. KIRK H ALL T. H. M AHEN CORNELL ASSIS ANT C LM RAL COUNSEL FDWARD J. CULLEN. JR.

JOHN F. KENNEDY. JR.

assistant COUNSEL Mr. A. Schwencer, Chief Docket Nos. 50-352 Licensing Branch No. 2 50-353 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Limerick Generating Station Units 1 and 2 Revisions to ER0L Section 7.1 and Responses to RAI's E450.1, E450.2, E450.3, and E450.4

Reference:

Letter, A. Schwencer to E. G. Bauer, Jr.,

dated August 11, 1982,

Subject:

RAI -

Limerick EROL

Dear Mr. Schwencer:

Transmitted herewith are draft responses and EROL page changes related to the subject RAI's. This material is provided in draft form at the request of R. Martin, NRC Project Manager for Limerick, as an aid to the staff review. We plan to formally incorporate these responses and page changes into the ER0L at the end of April.

Very truly yours, SHG/dg/E/8 Enclosures Copy to: See attached service list gQ 8304010347 830328 PDR ADOCK 05000352

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Judge Richard F. Cole (w/o enclosure)

Judge Peter A. Morris (w/o enclosure)

Troy B. Conner, Jr. , Esq. (w/o eG]ihre)

Ann P. Hodgdon ~~.

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Mr. Frank R. Btrnano y (w/o enclosure) .'

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Mr. Robert L. Anthony Mr. Marvin I. Lewis (w/o enclosure)

Judith A. Dorsey, Esq. (w/o enclosure) ..

Charles W. Elliott, Esq. (w/o enclosure) 1i Mr. Alan J. Nogee (w/o enclosure) ,,

Pobert W. Mler, Esq. (w/o enclosure)

Mr. Thmas Gerusky (w/o enclosure)

T Director, Pennsylvania Emergency Managment Agency (w/o enclosure)

Mr. Steven P. Hershey (w/o enclosure) ~

James M. Neill, Esq. (w/o enclosse)

Donald S. Bronstein, Esq. (w/o enclosure) .,

Mr. Joseph H. hhite, III (w/o enclosure)

Dr. Judith H. Johnsrud ,(w/o enclo'sure)

Walter W. Cohen, Esq. (w/o enclosure)

Robert J. Sugarman, Esq. (w/o"enh$mre)

' Rodney D. Johnson (w/o enclocure)

Atmic Safety and Licensing Appeal Board (w/o enclosure) "

Atmic Safety and Licensing Board Panel 7 (w/o enclosure)

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DRAFT g LGS EROL s CHAPTER 7 ENVIRONMENTAL EFFECTS OF ACCIDENTS TABLE OF CONTENTS Section Title 7.1 STATION ACCIDENTS INVOLVING RADIOACTIVITY 7.1.1 ' Approach to the Analysis of Class 1-8 Accidents 7.1.2 Models'and Data Used to Evaluate the Eravironmental

, Consequences of Class 1-8 Accidents 7.1.2.1 Radiation Dose Models and Data for Class 1-8 Accidents

7.1.2.2 Source Term Models and Data for Class 1-8 Accidents

' l 7.1.2.3 Atmospheric Diffusion Estimates for Class 1-8 Accidents 7.1.3 Class 1-8 Accident Analysis 7.1.3.1 Class 1 - Trivial Incident Inside Primary Containment 7.1.3.2 Class 2 - Small Releases Outside Primary t

~

Containment

! 7.1.3.3-s Class 3 - Radwaste System Failure

! 7.1.3.3.1.

Equipment Leakage oWPMalfunction 7.1.'3.3.1 Offgas Treatment System Failure 7.1.3.3.3 Release of Waste Sludge Tank Contents 7.1.3.4 Class 4 - Fission Products to Primary System (BWR) 7.1.3.4.1 Fuel Cladding Defects 7.1.3.4.2 Off-Design Transients that Induce Fuel Failure i

7.1.3.5 Class 5 - Fission Products to Primary and Secondary Systems (PWR)

. 7.1.3.6 Class 6 - Refueling Accidents 7.1.3.6.1. ' Fuel Assembly Drop 7.1.3.6.2 Heavy Object Drop onto Fuel in Core 7.1.3.7 -

Class 7 - Spent Fuel Handling Accidents 7.1.3.7.1 Fuel' Assembly Drop in Fuel Storage Pool 7.1.3.7.2 Heavy Object Drop onto Fuel Racks 7.1.3.7.3 Fuel Cask Drop 7.1.3.8 Class 8 - Accident Initiation Events Considered Design Basis Evaluation in Safety Analysis Report 7,1.3.8.1 Loss of Coolant Accidents (LOCA) 7.1.3.8.2 Control Rod Accidents 7.1.3.8.3 Steam Line Break Accidents 7.1.3.9 Summary of Environmental Consequences and Public l

Risk of Class 1-8 Accidents 7.1.4 Approach to the Analysis of Severe Accidents 7.1.4.1 Models and Data 7.1.4.1.1 Source Term Description and Associated Frequencies 7.1.4.1.2 Consequence Model I

7-i Rev. 12, 04/83 m

l-

DRAFT LGS EROL CHAPTER 7 TABLE OF CONTENTS (Cont'd)

Section Title 7.1.4.1.3 Uncertainty 7.1.4.2 Analysis 7.1.4.3 Results 7.1.4.3.1 CCDFs 7.1.4.3.2 Risk Considerations Conclusions 7.1.4.4 7.1.5 References 7.2 TRANSPORTATION ACCIDF.NTS INVOLVING RADIOACTIVITY 7.3 OTHER ACCIDENTS 7.3.1 Storage and Use of Oil 7.3.2 Storage of Condensate and Refueling Water 7.3.3 Storage and Use of Acid and Caustic 7.3.4 Storage and Use of Chlorine 7.3.5 Storage and Use of Compressed Gases l 7.3.6 Summary I

l l

l 7-ii Rev. 12, 04/83

DRAFT LGS EROL CHAPTER 7 TABLES No. Title 7.1-1 Classification of Postulated Incidents 7.1-2 Physical Data for Radiation Dose Models

.7.1-3 Fission Product Inventories.in the Fuel 7.1-4 Equilibrium Primary Coolant Radicactivity 7.1-5 Effective Probability Levels for Fifty Percentile X/O 7.1-6 Fifty Percentile Atmospheric Diffusion Factors -

X/O (sec/m2) 7.1-7 Class 3.1 Accident - Radioactivity Released as a Result of Equipment Leakage or Malfunction 7.1-8 Class 3.2 Accident - Radioactivity Released as a Result of First Charcoal Bed Failure in the Offgas Treatment System 7.1-9 Class 3.3 Accident - Radioactivity Released as a Result of Gross Equipment Failure 7.1-10 Class 4.2 Accident - Radioactivity Released as a Result of an Off-Design Transient Accident 7.1-11 Class 6.1 Accident - Radioactivity Released as a Result of,a Fuel Assembly Drop 7.1-12 Class 6.2 Accident - Radioactivity Released as a Result of a Heavy Object Dropped onto Fuel ir. Core 7,.1-13 Class 7.2 Accident - Radioactivity Released as a Result of Heavy Object Dropped onto Fuel Rack 7.1-14 Class 8.1 Accident - Radioactivity Released as a Result of Loss of Coolant - Small Pipe Break 7.1-15 Class 8.1 Accident - Radioactivity Released as a Result of Loss of Coolant - Large Pipe Break 7.1-16 Class 8.1(a) Accident - Radioactivity Released as a Result of a Primary Systems Instrument Line Break 7-iii Rev. 12, 04/83

DRAFT LGS EROL CHAPTER 7 TABLES (Cont'd)

No. Title 7.1-17 Class 8.2(b) Accident - Radioactivity Released as a Result of a Rod Drop Accident 7.1-18 Class 8.3(b) Accident - Radioactivity Released as a Result of Steam Line Break - Small Pipe 7.1-19 Class B.3(b) Accident - Radioactivity Released as a Result of a Steam Line Break,- Large Pipe 7.1-20 Summary of Maximum Exclusion Area Boundary ED Doses Resulting From Accidents 7.1-21 Summary of Population Doses Resulting From Accidents 7.1-22 Source Term Characteristics - Point Estimate l 7.1-23 Frequencies of Table 7.1-22 Source Terms l 7.1-24 Activity in the Limerick Reactor Core at 3293 MWt l 7.1-25 Permanent Resident Population for the Limerick Site l 7.1-26 Average Values of Environmental Risks Due to Accidents Per Reactor-Year l

l 7.2-1 Environmental Impact of Transportation of Fuel and l

Waste l

l l

l 7-iv Rev. 12, 04/83

DRAFT LGS EROL CHAPTER 7 FIGURES No. Title 7.1-1 Schematic Outline of Consequence Model l 7.1-2 Median CCDF of Bone Marrow Dose Greater than 200 Rem 7.1-3 Median CCDF of Population Exposure l 7.1-4 Median CCCF of Actue Fatalities l 7.1-5 Median CCDF of Latent Cancer Fatalities l 7.1-6 Median CCDF of Ex-Plant Costs l 7.1-7 Median Individual Risk of Early Fatality as a Function of Distance 7-v Rev. 12, 04/83

LGS EROL CHAPTER 7 DRAFT ENVIRONMENTAL EFFECTS OF ACCIDENTS 7.1 STATION ACCIDENTS INVOLVING RADIOACTIVITY The purpose of this section is to consider the potential radiological effects on the environment of accidental events and to compare these potential effects with those of r.ormal station operation and natural background radiation. Radiological effects that result from normal station operation are discussed in '

Section 5.2, and natural background radiation is discussed in Section 6.4.

A detailed accident and safety analysis is a normal part of the design and licensing of each power station. The results of this analysis are presented to the NRC in the form of safety analysis reports (SARs). These reports contain detailed descriptions of the facility and station site, as well as a highly conservative l analysis of the effects of normal and abnormal plant conditions.

t In addition to the analysis presented in the SAR, further examination of the environmental effects of normal and abnormal l station conditions, based upon realistic parameters, is required to be presented in this Environmental Report. An assessment of the risks associated with the Limerick plant from accidents more severe than included in the design bases for the station was undertaken and is required to be presented in Section 7.1.4.

There are two main aspects of station safety: prevention of station accidents, and containment of radioactivity in the event of an accident. Prevention of station accidents begins with conservative design of the reactor and its control system, and conservative engineering of the reactor installation. Starting with this base, the designer seeks to anticipate the possible sources of malfunction, and to make provisions for mitigating their effects in the design. A strict quality assurance program ensures high component and system reliability.

Radioactive materials produced in the core of the reactor are contained within the station by a number of successive barriers that are incorporated in the station design. These barriers are the fuel material, zircaloy fuel cladding, the steel wall of the i

reactor vessel, and the primary and secondary containment systems. Containment of radioactivity in the event of an accident also invclves the incorporation of engineered safety 7.1-1 Rev. 12, 04/83 I

LGS EROL DRAFT l

features (ESF) in the station design, such as radiation shields, emergency cooling systems, and air filtration systems.

l In considering the environmental effects of postulated station ,

accidents, several important distinctions must be made from other !

station environmental effects. The estimated effects are i potential ratner than certain. As a result of measures taken, or prevention of accident through design, manufacture, and operation, occurrences of accidental events in operating nuclear power plants have been rare. The improbability of accidental events in operating nuclear plants has been maintained at this low level through design review, operating limits, and quality assurance procedures. Therefore, the environmental effects of these potential events must be considered in conjunction with their probability of occurrence.

7.1.1 APPROACH TO THE ANALYSIS OF CLASS 1-8 ACCIDENTS l In the Federal Register of June 13, 1980 (45FR 40101), the Nuclear Regulatory Commission published a statement of interim policy regarding accident considerations. This statement withdrew the proposed annex to Appendix D of 10CFR50 and suspended the rulemaking procedures associated with it. It also put forward the Commission's interim policy that

...Envir'onmental Impact Statements shall include consideration I of the site-specific environmental impacts attributable to l accident sequences that can result in inadequate cooling of the I reactor fuel and in melting of the reactor core. In this regard, attention shall be given both to the probability of occurrence of such releases and to the environmental consequences of such releases."

Accordingly, Section 7.1.4 describes an analysis of the public risk associated with these severe accidents.

Although, as is described above, the proposed annex was subsequently withdrawn, the information for accidents formerly l designated as Class 1-8 is given in Sections 7.1.1 to '7.1.3. The l

public risk associated with these accidents is summarized in Section 7.1.3.9.

The occurrence of abnormal station conditions and accidental events must be considered in design, licensing, and operation of '

nuclear power plants. In technical terms, an accident is an unexpected chain of events (i.e., a process rather than a single event). In SARs, the basic events involved in various possible Rev. 12, 04/83 7.1-2

LGS EROL DRAFT ;

station accidents are identified and studied with regard to the adequacy of the performance of the engineered safety features (ESF). In addition, the potential radiological effects of station accidents are analyzed by the evaluation of physical factors involved in each chain of events that might result in radiation exposures to humans. These factors include the meteorological conditions existing at the time of the accident, radionuclide uptake rates, and exposure times and distances, as well as the many factors that depend upon station design and the mode of operation. In these analyses, the factors affecting the consequences of each accident are identified and evaluated, and uncertainties in their values are discusced. Because some degree of uncertainty always exists in the prediction of these factors, it has become general practice in SARs to assume conservative values in making calculated estimates of radiation doses.

As a result of the highly conservative analysis, the radiation exposure levels calculated in SARs are not actually expected to be reached, even if the event initiating the accident occurs. In fact, the calculated exposures resulting from a DBA are generally far in excess of what would be expected, and do not provide a realistic means of assessing the radiological effects of postulated station accidents. In the analyses presented here, the radiation exposures associated with station accidents have been analyzed on a more realistic basis, as specified in the proposed annex to Appendix D of 10 CFR Part 50, which is

referenced by NRC Regulatory Guide 4.2, Rev. 2 (Ref 7.1-1). In many cases, the assumptions are still conservative in that the

( most probable assumptions would result in even lower radiation exposure.

The effectiveness of measures that have been taken for accident prevention is judged by the frequency at which the accident occurs; that is, the accident probability. The effectiveness of the measures taken in containment of radioactivity can be judged by the calculated values of the radiological exposures associated with each accident. As discussed in the Federal Register (36 FR 22851) for the proposed annex to Appendix D of 10 CFR Part 50, the determination of the environmental impact of potential accidents requires the consideration of both the potential l

exposures, and the probabilities of receiving these exposures.

The environmental impact of the postulated accidents is evaluated for eight accident classes identified in Table 7.1-1. These classes are defined in the proposed anner to Appendix D of 10 CFR Part 50.

i 7.1-3 Rev. 12, 04/83

DRAFT LGS EROL 7.1.2 MODELS AND DATA USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF CLASS 1-8 ACCIDENTS l Maximum individual dose estimates are based upon a receptor located at the exclusion area boundary. Man-rem dose estimates are based upon the year 2000 population projections. The population distribution as a function of distance and sector for the year 2000 has been estimated, and presented in Section 2.1.

The total population dose was determined by taking the product of

. the dose and the number'of people receiving that dose in an area segment defined by a 22.50 sector, at a particular distance from the station, and summing the product of each 22.50 sector for a distance out to 50 miles from the station. l 7.1.2.1 Radiation Dose Models and Data for Class 1-8 Accidents l The'models used are based upon NRC Regulatory Guides 1.3 (Ref 7.1-2) and 1.25 (Ref 7.1-3). The following assumptions are basic to both the model for the whole-body dose due to immersion in a j cloud of radioactivity, and the model for the thyroid dose due to inhalation of radioactivity:

a. Direct radiation from the station is negligible compared to whole-body radiation due to immersion in the cloud of radioactivity.

b. All radioactive releases are treated as ground level releases, regardless of the point of discharge.

c. Continuous release atmospheric dispersion factors are applicable, and cloud depletion due to ground deposition is assumed to be insignificant.

d. The dose receptor is a standard man, as defin9d by the International Commission on Radiological Protection

. (ICRP) (Ref 7.1-4).

For all distances and time periods, the semi-infinite cloud model

! is used to calculate the whole-body dose. The procedure results in population exposures that are conservative.

The semi-infinite, whole-body gamma dose is given by the following equation from TID-24190 (Ref 7.1-5):

Rev. 12, 04/83 7.1-4

DRAFT LGS EROL N

rDoo = (0.25) (X/Q) I (Qi)(Ei) (7.1-1) l i=1 where:

rDoo = gamma dose from semi-infinite cloud (rad)

X/O = atmospheric dilution factor (sec/ meter 3)

N = number of isotopes Oi = source strength for isotope 1 (curies)

Ei = average gamma energy for isotope 1 (MeV/ dis)

The thyroid dose for a given time period is obtained from the following equation:

N D= (X/0)(BR) E (01)(DCFi) (7.1-2) l i=1 ,

where:

D = thyroid inhalation dose (rem)

X/O = atmospheric dilution factor (sec/ meter 3)

BR = breathing rate (meter 3/sec)

N = number of isotopes 01 = total activity of iodine isotope i released (curies)

DCFi = dose conve sion factor for iodine isotope i (rem / curies inhaled)

Table 7.1-2 lists the physical data for the radiation dose models. The half-life values were taken from the Meek and Rider Report (Ref 7.1-6), and are in general agreement with those in

! TID-14844 (Ref 7.1-7) and ORNL-2127 (Ref 7.1-8). The values for the gamma energies are those given in the Table of Isotopes (Ref 7.1-9). The thyroid dose conversion factors are taken from the ICRP Committee II Report (Ref 7.1-10), and the breathing rates used in the calculations of inhalation doses are based upon the average daily breathing rates assumed in the ICRP Report, which are also used in the NRC Regulatory Guide 1.3 (Ref 7.1-2). l 7.1-5 Rev. 12, 04/83

DRAFT' LGS EROL 7.1.2.2 Source Term Models and Data for Class 1-8 Accidents l It is the purpose of this section to provide the general information used for accident evaluations.

The inventories of radioactive materials in the fuel pellets and fuel rod gap spaces in the reactor core depend upon the following:

a. Core power

b. Plant capacity factor

c. Temperature distribution in the pellets

d. Length of operating time prior to the accident.or shutdown

e. Diffusion rates of radioisotopes through the fuel pellet materials.

Fission product inventories for the core and gap are based upon operation at 3458 MWt for 1000 days. Activity inventories for the total core, total gap, and gap of one fuel rod are given in t Table 7.1-3. Reactor coolant concentrations are given in Table 7.1-4. These coolant concentrations were calculated using the methodology of NUREG-0016 (Ref 7.1-11).

7.1.2.3 Atmospheric Diffusion Estimates for Class 1-8 Accidents l Estimates of atmospheric diffusion (X/Q) have been made at the exclusion area boundary, the outer boundary of the low population zone (LPZ), and at 0.5, 1.5, 2.5, 3.5, 4.5, 7.5, 15, 25, 35, and 45 miles for each sector. These estimates have been made for periods of 2, 8, and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , and 3 and 26 days following a postulated accident. The sector-dependent hiodel in Draft RegulatoryGuide1.Wg,(Ref7.1-12) has been used.

/p[

The calculation procedure used to determi e X/O for the appropriate time periods following a pos ulated accident is described in Draft Regulatory Guide 1. . The diffusion model presented in this guide is used to determine X/O values for the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following the accident. X/Q values for longer time periods are determined by logarithmic interpolation between the 2-hour accident value and the annual X/O at each receptor point.

Rev. 12, 04/83 7.1-6

/16I The annual X/O values have been calculated u ing the model described in Regulatory Guide 1.111 (Ref 7. 13). The Limerick emission has been classified as a low-leve release, according to the criteria of Draft Regulatory Guide 1. . This requires that the source be treated as ground level. This assumption has also been made in the annual X/O calculations.

Meteorological data from Limerick Weather Station No. 1, from January 1972 through December 1974, have been used in the diffusion calculations. Lapse rate wind distributions have been computed using wind speed and direction from the 30-foot level, and temperature difference from the 266-26 foot hei-ght interval.

The lapse rate, wind speed, and wind direction categories are consistent with the recommendations of Regulatory Guide 1.23 (Ref 7.1-14). The wind distribution used to calculate the 2-hour accident X/O values has been normalized by di,5 p tional sector, in accordance with Draft Regulatory Guide 1.MNK N. This distribution is shcwn in Table 2.3.2-2. In each sector, the total frequency of wind speed and stability categories equals 100%. The stability classes designated as 1 through 7 in this distribution refer to the Pasquill classes A through G. A wind distribution computed in the standard manner is shown in Table 2.3.2-42. This distribution was used to calculate the annual X/O values used in the logarithmic interpolation scheme.

The dispersion parameters developed by Pasquill (Ref 7.1-15) and Gifford (Rsf 7.1-16) have been used in the accident calculations.

Analytical approximations to these curves, developed by Eimutis and Konicek (Ref 7.1-17), have been used for sigma-y. The approximations of Busse and Zimmerman (Ref 7.1-18) have been used for sigma-z. A building wake correction of 2298ma was used.

This is equal to one-half the minimum cross-sectional area of the reactor turbine enclosure complex.

The effective probability level is an adjustment necessary to equate the directionally dependent approach of Draft Regulatory Guide 1.XXX with the 50th percentile criterion previously employed by the NRC in the directionally independent model. This parameter is calculated as follows:

Pe = P(N/n) (7.1-3) l s

where:

Pe = effective probability level-7.1-7 Rev. 12, 04/83

LGS EROL P = desired probability level (50%)

N = total number of hours having valid wind and stability data in the period of record n =

total number of hours having valid wind and stability data in the directional sector of interest S =

total number of directional sectors (16)

The effective probability levels calculated for each sector at the Limerick Generating Station are listed in Table 7.1-5.

Cumulative frequency distributions of X/O for the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a postulated accident were computed for distances of interest in each sector. These distributiens were then plotted on a log probability scale. In each plot, the data points were enveloped by a fitting function, as described by Markee and Levine (Ref 7.1-19). The accident X/Q values in each directional sector were then obtained from the intersection of this function and the effective probability level.

Accident X/O values for. periods of 8 and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months and 3 and 26 days following the accident have been determined by logarithmic interpolation between the maximum 2-hour and the maximum annual X/O at each distance. A complete summation of the estimated X/0 values for the entire duration of the postulated accident is ,

given in Table 7.1-6 for distances up to 50 miles for each sector.

7.1.3 CLASS 1-8 ACCIDENT ANALYSIS. l

! In the following subsections, postulated accidents are identified I and analyzed, and their radiological consequences are estimated.

j 7.1.3.1 Class 1 - Trivial Accidents Inside Primary Containment Class 1 accidents are postulated as the release of small quantities of radioactive material inside the primary containment. The various mechanisms by which this may occur include small spills and small leaks from equipment and valve t

packing. A low level of continuous leakage from components such l

as valve packing stems, pump seals, and flanges, etc, is expected. Radioactivity release events of this class are considered as part of normal operating conditions, and analyzed along with radioactivity releases due to normal operation in Sections 3.5 and 5.2.

j Rev. 12, 04/83 7.1-8

DRAFT LGS EROL 7.1.3.2 Class 2 - Small Releases Outside Primary Containment Class 2 events are postulated as the release of small quantities of radioactive material outside the primary containment. These include small spills and leaks from equipment outside the primary containment. A low level of continuous leakage from components such as valve packing stems, pump seals, and flanges, etc, is expected. Radioactivity release events of this class are considered to be minor perturbations of normal operating conditions, and analyzed as " miscellaneous leakages," along with radioactivity releases due to normal operation in Sections 3.5 and 5.2.

l The events in Classes 1 and 2 represent occurrences that are

anticipated during station operation. Their consequences, which i are small, are considered within the framework of routine effluents from the station.

7.1.3.3 Class 3 - Radwaste System Failure l

l Class 3 accidents are postulated to involve the release of I

radioactivity to the environment through a failure, or malfunction, in the radwaste systems.

l The most serious radiological consequences will be caused by a j release from the waste sludge tank in the solid radwaste system, l or from the charcoal delay tank in the offgas treatment system.

l A number of combinations of inadvertent operator errors and equipment malfunctions, or failures, could be identified that might result in a release of some or all of the radioactivity stored in the waste sludge tank and the offgas treatment system charcoal delay tank. Iodine isotopes in the liquid tank are l assumed to become partially airborne after its failure. In general, the amounts of radioactivity that could be released by any such combinat.vn of events are limited in the following ways:

Stat' ion Feature Function Limits on reactor coolant Restricts total curies present activity in radwaste system tanks .

Radiation monitors Allow early detection of radioactivity releases, 7.81-9 Rev. 12, 04/83 l __ _ .- .. . _ , .__. ..

s DRAFT LGS EROL allowing operator action to terminate release Limits on tank size Restricts total curies present in any one tank Tsolation valves Allow operator to terminate radioactivity releases Interlock procedures Reduce probability of inadvertent releases Charcoal filters Delay tanks are' continuously vented to limit the accumulation of gases ,

Three releases of different types have been analyzed to cover the range of postulated events.

7.1.3.3.1 Class 3.1 - Equipment Leskage or Malfunction The accident postulated is a failure of equipment in the liquid radwaste system that would cause the sudden release to the radwaste enclosure of 25% of the average inventory contained in the waste sludge tank. This tank is considered because its failure would result in the largest amount of radioactivity (iodine) released from the radwaste enclosure by the failure of any one tank. The radioactivity of the liquid released is based on the normal accumulation of liquid radwaste over a 6-day l period.

The parameters and assumptions used in this analysis are as follows:

a. Twenty-five percent of the average inventory of' l accumulated liquid waste will be spilled.

I

b. An iodine partition factor of 0.01 is used for analysis,

c. Noble gas release as a result of the accident is negligible.

Rev. 12, 04/83 7.1-10

DRAFT LGS EROL

d. There is no liquid released to the environment.

e. Meteorology for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is used because the release from this accident is expected to last for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The radioactivity released to the environment is given in Table 7.1-7.

7.1.3.3.2 Class 3.2 - Offgas Treatment System Failure The c#fgas treatment system has been incorporated in the station design to reduce the gaseous radwaste release from the station.

It is assumed the.t, within this system, the first charcoal delay tank failure would result in the most significant whole-body dose. The analysis of this event is based on the following assumptions:

a. Source term: an offgas release rate of 60,000 microcuries/sec after 30 minutes decay, and maximum accumulated activity in the first charcoal delay tank based on 22.5 days buildup time for xenon and 0.98 days buildup time for krypton.

b. Releast of 100% of the noble gas activity contained in the first cnarcoal delay tank. The iodine releases are negligible.

c. Meteorology for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is used because the release from this accident is expected to last for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

l The radioactivity released to the environment is given in Table l 7.1-8.

7.1.3.3.3 Class 3.3 - Release of Waste Sludge Tank Contents This accident is defined to be the sudden release of 100% of the average inventory contained in the waste sludge tank. Other assumptions used in evaluating the consequences of this accident are identical to those used in the Class 3.1 accident. The 7.1-11 Rev. 12, 04/83 1

DRAFT LGS EROL radioactivity released to the environment is given in Table 7.1-9.

In making an assessment of the probability of releases of this type, it is not possible to establish precise numerical values.

Events in Class 3 are not anticipated during station operation.

7.1.3.4 Class 4 - Fission Products to Primary System (BWR)

Class 4 accidents are postulated as those events that release radioactivity from the fuel into the primary system.

To demonstrate the potential environmental consequences of these events, two situations are postulated and evaluated:

a. Fuel cladding defects

b. Off-design transients that' induce fuel failures above those expected (such as flow blockage and flux maldistributions).

l 7.1.3.4.1 Class 4.1 - Fuel Cladding Defects i

l Releases from these events are included and evaluated under routine releases in accordance with 10 CFR Part 50, Appendix I, and included in the routine radioactive discharge discussed in Section 5.2.

l 7.1.3.4.2 Class 4.2 - O.ff-Design Transients That Induce Fuel Failures Above Those Expected (Such as Flow Blockage and Flux Maldistributions) i This accident is assumed to induce fuel failures to the core l

above those normally expected. The following assumptions are postulated for an off-design transient:

a. A release into the reactor coolant of 0.02% of the core inventory of noble gases and 0.02% of the core inventory l of halogens.

l l

Rev. 12, 04/83 7.1-12

DRAFT LGS,EROL

b. One percent of the halegens and 100% of the noble gases in the reactor coolant are released into the steam.

. c. If the radioactivity release from the core is high, the radiation monitors in the main steam line will initiate :

MSIV closure. MSIV closure will result in the release of radioactivity to the turbine enclosure by condenser leakage, and then to the atmosphere. If the j radioactivity level is not high enough to trip the MSL monitor, the inventory released from the core will be processed through the offgas treatment system, from ,

which the eventual release of radioactivity yields a lower exclusion area boundary (EAB) whole-body gamma

! dose than that from condenser leakage. The more -

conservative case (radioactivity released through condenser leakage) is used in this accident analysis.

d. Radioactivity is carried over to the condenser, where 10% of the halogens and 100% of the noble gases are available for leakage from the condenser to the j '

environment at 0.5% per day of condenser volume for the i course of the accident (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

l

e. Meteorology used is for a 24-hour accident.

The radioactivity released to the environs-for the duration of the accident is given in Table 7.1-10.

7.1.3.5 Class 5 - Fission Products to Primary and Secondary Systems (PWR)

Analysis of a Class 5 accident is not applicable because the reactor is a BWR.

7.1.3.6 Class 6 - Refueling Accidents Class 6 accidents are postulated to include refueling accidents inside the refueling area. Following the accident, radioactive material is released to the environs from the refueling area via the standby gas treatment system. It should be noted that the refueling area will be automatically isolated on detection of high radiation levels in the ventilation exhaust air from the refueling area.

7.1-13 Rev. 12, 04/83

. -_ _ a . - - - - --. . . . - - ._ - - - - -

DRAFT LGS EROL To demonstrate the potential environmental consequences of this type of accident, two refueling accidents are postulated and evaluated:

a. Fuel assembly drop

b. Heavy object drop onto fuel in core.

7.1.3.6.1 Class 6.1 - Fuel Assembly Drop A fuel assembly drop is postulated to occur as a result of the mishandling of a spent fuel assembly. The accident is assumed to result in damage to one row of fuel rods in the assembly. The subsequent release of radioactivity from the damaged fuel assembly will bubble through the water covering the assembly, where most of the radioactive iodine will be entrained. The following assumptions are postulated for a fuel assembly drop accident:

a. The gap activity (noble gases and halogens) in one row of fuel rods is released into the water. (Gap activity is 1% of total activity in a rod.)

i b. There is a one-week decay time before the accident occurs.

c. Iodine decontamination factor in water is 500. Noble gases are not retained by water.

I

d. Fission products released to the refueling area atmosphere are mixed by the reactor enclosure recirculation system. Part of the recirculated flow is exhausted to the environment via the standby gas treatment system.

l

e. The filter efficiency for iodines of the standby gas i

treatment system is 99%, that of the reactor enclosure

! recirculation and filt;ation system is'95%.

f. Meteorology for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is used because the release from this accident is expected to last for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Rev. 12, 04/83 7.1-14

DRAFT LGS EROL The activity contained in the fuel rod gap, and that released I

from the refueling area is a result of this accident, is given in Table 7.1-11.

7.1.3.6.2 Class 6.2 - Heavy Object Drop onto Fuel in Core This accident is assumed to result in damage to an average fuel assembly. The same assumptions as used in the fuel assembly drop accident apply, except that 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of decay time is assumed before the object drop cccurs. The radioactivity released to the pool water, and from the refueling area as a result of this 4 accident, is given Table 7.1-12.

J l

7.1.3.7 Class 7 - Spent Fuel Handlina Accidents i

i Class 7 accidents are postulated to include spent fuel handling accidents in the refueling area. Following accidents in the refueling area, evacuation and isolation of the area will be initiated by high radiation alarms. The normal HVAC system in the area will be automatically isolated. The refueling area atmosphere will then be treated by the reactor enclosure recirculation system and the standby gas treatment system before release to the env, irons. To demonstrate the potential environmental consequences'of this type of accident, three spent i fuel handling accidents are postulated and evaluated:

a. Fuel assembly drop in fuel storage pool j b. Heavy object drop onto fuel rack

c. Fuel cask drop.

7.1.3.7.1 Class 7.1 - Fuel Assembly Drop in Fuel Storage Pool This accident is defined as the mishandling of a spent fuel assembly and assumes the same radioactivity release as postulated for a Class 6.1 accident. The assumptions used in evaluating this accident, as well as the resultant offsite doses, are identical to those in Class 6.1 (Section 7.1.3.6.1).

7.1-15 Rev. 12, 04/83

DRAFT LGS EROL 7.1.3.7.2 Class 7.2 - Heavy Object Drop Onto Fuel Racks This accident assumes a release of radioactivity from a damaged

' fuel assembly, similar to that postulated for the Class 6.2 accident, except that a 30-day decay period before the accident j

~

occurs is assumed. Other assumptions used are identical to those in Class 6.2 (Section 7.1.3.6.2). Table 7.1-13 lists the activity release from the fuel assembly to the spent fuel pool and the activity released to the environment.

7.1.3.7.3 Class 7.3 - Fuel Cask Drop i

i The spent fuel cask will be equipped with redundant sets of lifting lugs and yokes compatible with the reactor enclosure crane main hook, thus preventing a cask drop due to a single failure. Therefore, the spent fuel cask drop is not considered to be a credible accident, and no analysis was performed. FSAR Section 9.1.5 describes the reactor enclosure crane and the interlocks that prevent moving the spent fuel cask over the fuel pool.

During fuel handling operations in the reactor enclosure, there

, exists the remote possibility that one or more fuel assemblies will sustain some mechanical damage. There exists an even more remote possibility that this damage will be severe enough to breach the cladding and release some of the radioactive fission products contained therein. Accidents in Classes 6 and 7 are of similar or lower probability than accidents in Classes 3 and 4, but are still possible.

7.1.3.8 Class 8 - Accident Initiation Events Considered for l Desian Basis Evaluation in the Safety Analysis Report

, Class 8 accidents include the loss-of-coolant accident (small and l large pipe breaks), reactivity excursion accident, and steam line break accident. -

7.1.3.8.1 Class 8.1 - Loss-of-Coolant Accidents (LOCA) l l -

l A LOCA is defined as a loss of reactor coolant due to a sudden circumferential rupture of a reactor coolant system pipe, or any line connected to that system, inside containment.

i i

Rev. 12, 04/83 7.1-16

DRAFT LGS EROL To demonstrate the potential environmental consequences of this type of accident, two LOCAs are postulated and evaluated:

a. Small pipe break (6 inches or less)

b. Large pipe break.

7.1.3.8.1.1 Small Pipe Break (6* inches or less)

The following assumptions and parameters are postulated for

evaluating the environmental consequences of a LOCA for a small pipe, break (6 inches or less)

a. Source term: The average radioactivity inventory in the m

pri'ary coolant is released to the primary containment.

i b. A reduction factor of 0.2 is used in the source term for

! the effects of plateout and the decontamination factor in the pool.

c. The effects of radiological decay during holdup in the containment are taken into account.

d. The free iodine and noble gases leak from the primary containment to the reactor enclosure at a rate of 0.5%

of the contained volume per day.

e. Fifty percent mixing in the reactor enclosure.

i f. Negative pressure in the reactor enclosure is maintained for the duration of the accident, and whatever is leaked from the enclosure is released through the SGTS.

g. The SGTS exhausts a portion of the air from the reactor

, enclosure recirculation and filtration system. Charcoal l

filter efficiency for the standby gas treatment filters is 99% for iodines, and that for the reactor enclosure filtration system is 95%.

i i

7.1-17 Rev. 12, 04/83

LGS EROL DRAFT

h. The breathing rate for persons offsite is 3.47 x 10-*

meters 3/see for the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . From 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the accident, the breathing rate is 1.75 x 10-4 meters 3/sec. Thereafter, the rate is 2.32 x 10-* meters 3/sec.

i. Meteorology for both short time (<8 hours) and longer time (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 30 davs) releases is used for this accident.

.The release as a function of time from this accident is given in Table 7.1-14.

7.1.3.8.1.2 Large Pipe Break The large pipe break LOCA is assumed to be a sudden circumferential break of a recirculation line, permitting the discharge of coolant into the primary containment from both sides of the break. The assumptions and parameters postulated for evaluating the environmental conseugences of this accident are identical to those assumed for the LOCA small p'pe break, with the following exceptions:

a. Source Term: The average radioactivity inventory in the reactor coolant is released to the containment, plus a release into the coolant of 0.2% of the core inventory of halogens and noble gases.

~

b. Fission product inventories in the core are calculated at the end of core life (1000 days), assuming fuel power l

operation at 3458 MWt.

The release as a function of time is given in Table 7.1-15.

7.1.3.8.1.3 Class 8.1(a), Break in Instrument Line From Primary System That Penetrates the Containment This accident is postulated to involve lines outside the primary containment that are not provided with isolation capacity inside the primary containment.

Rev. 12, 04/83 7.1-18

DRAFT I

LGS EROL l

i The following assumptions are used for a primary system instrument line break accident:

a. The average radioactivity inventory in the primary coolant is based on an offgas release rate of 60,000 microcuries/sec after 30 miautes delay.

b. Total mass release through the failed line is 25,000 lb.

c. The charcoal filter efficiency for the SGTS is 99% for iodine, and that for the reactor enclosure filtration system is 95%.

d. A reduc: ion factor of 0.1 in the source term is assumed from combined plateout and building . nixing. .

e. Meteorology for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is used for this accident.

The activity releases from this accident are given in Table 7.1-16.

7.1.3.8.2 Class 8.2 - Control Rod Accidents 7.1.3.8.2.1 Class 8.2(a), Rod Ejection Accident (PWR)

This class cf accident is not applicable for this analysis.

i 7.1.3.8.2.2 Class 8.2(b), Rod Drop Accident (BWR)

A rod drop accident is defined as the complete (but not necessarily sudden) rupture, breakage, or disconnection of a random fully-inserted control rod drive from its cruciform control blade, at or near the coupling, in such a way that the blade becomes stuch at its location (fully inserted). This assumption sets up a condition where, if the drive were withdrawn, the stuck blade could later fall from the core, causing a reactivity excursion accident. The following assumpticns are postulated for a rod drop accident:

7.1-19 Rev. 12, 04/83

a

_- .-. ..,-.m

, . - - - - - -.-n-- - . - - - , _ _ , , -

-.a LGS EROL

a. There is a release into the coolant of 0.025% of the core inventor *j of noble gases and 0.025% of the core inventory of halogens.

b. One perecnt of the halogens and 100% of the noble gases in the reactor coolant are released into the condenser.

c. A high radiation signal in the main steam lines will automatically close the MSIVs and trip and mechanical vacuum pump. Activity 4.n the turbine-condenser offgas

! systems will leak to the turbine enclosure, and then to the atmosphere.

i d. Radioactivity is carried over to the condenser, where 10% of the halogens and 100% of the nobic gases are available for leakage from the condenser at 0.5% of the condenser. volume per day for the course of the accident (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). -

e. Meteorology used is for a 24-hour accident.

The activity released to the environs, as a function of time for the duration of the rod drop accident, is given in Table 7.1-17.

f 7.1.3.8.3 Class 8.3 - Steam Line Break Accidents 7.1.3.8.3.1 Class 8.3(a), Steam Line Breaks (PWR) l This class of accident is not applicable for this analysis.

, 7.1.3.8.3.2 Class 8.3(b), Steam Line Breaks (BWR) l A steam line break accident is a circumferential break of a main steam line outside primary containment.

To demonstrate the potential environmental consequences of this l type of accident, two steam line breck accidents are postulated and evaluated:

a. Small pipe break (of 0.25 fta)

Rev. 12, 04/83 7.1-20

DRAFT LGS EROL

b. Large pipe break.

For these postulated breaks, considering the most probable operating conditions prior to the break and using realistic assumptions, the calculated two-phase mixture level in the reactor pressure vessel does not reach the steam line before isolation is complete. Therefore, only steam will issue from these breaks for the entire transient.

Small Pipe Break (of 0.25 ft2): The following assumptions and parameters are postulated for evaluating the environmental consequences of a main steam line break accident for a small pipe break:

a. The primary coolant activity is based on an offgas release rate of 60,000 microcuries/sec after 30 minutes delay,

b. It is assumed that the main steam line will release coolant for 5 seconds after the isolation signal is l received.

c. The total amount of steam escaping from the break is 2750 lb. This quantity is the sum of a steam loss for two time periods, a 0.5-second duration prior to reactor trip, and a 5-second duration to complete closure of the MSIVs. .

d. Iodine in the fluid released to the atmosphere is at l one-tenth the primary system liquid concentration.

e. Fifty percent of the iodines and 100% of the noble gas in the fluid exiting through the break are assumed to be released to the atmosphere.

l

f. Meteorology for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is used because the l release from this accident is expected to last for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The total activity released to the environs is given in Table 7.1-18.

I 7.1-21 Rev. 12, 04/83 ,

i _

LGS EROL DRAFT 4 .

Larce Pipe Breaks: The assumptions and parameters postulated for evaluating the environmental consequences of a main steam line break accident for a large pipe break are identical to those given for a small pipe break, with the exception that the total amount of steam escaping from the break is 36,000 pounds. This quantity is the sum of a steam loss for two time periods, a 0.5-second duration prior to reactor trip, and a 5-second duration to complete closure of the MSIVs.

.The total activity released to the environs is given in Table 7.1-19.

! In making an assessment of the probability,of the occurrence of typical events considered as DBAs in the FSAR, a firm numerical estimate is not possible because of the extreme rarity of such events. Quality assurance for design, manufacture, and operation, and highly conservative-design considerations combine to produce piping and vessels with an extremely low probability of failure. Therefore, when the consequences are weighted by probabilities, the environmental risk is low.

7.1.3.9 Summary of Environmental Consecuences and Public Risk

of Class 1-8 Accidents In the preceding discussion, a number of postulated accidents have been identified and analyzed. These selected events cover the full range of accident analyses formerly required in the NRC guidelines. The resulting estimates of potential station EAB doses as a result of each postulated accident, along with an assessment of the likelihood of each event, are listed in Table 7.1-20.

In the column giving the general assessment of the likelihood of these events and conditions, several categories have been used.

Those events that could be expected to occur at frequencies of from once per station lifetime to as often as once per year are classified " occasional". Those events or conditions that would be expected to occur at frequencies less than once per station lifetime are classified " rare". Finally, there are a number of events that are considered unlikely, with projected probabilities much less than once per station lifetime. These events have been classified " extremely rare".

Table 7.1-21 shows the estimated integrated exposure from each postulated accident to the population within 50 miles of the station. When considered with the probability of occurrence, the annual potential radiation exposure of the population from all the postulated accidents is a small fraction of the exposure from Rev. 12, 04/83 7.1-22

DRAFT

. LGS EROL _

natural background radiation and, in fact, is well within naturally occurring variations in the natural background.

From the results in the accident analysis, several specific conclusions can be reached concerning offsite doses:

a. The radiation exposures that would result from the '

occurrence of accidents are generally lower than those expected from normal operation, and much lower than that from natural background radiation. _

b. The population exposure'from possible station accidents:

is negligible when compared to the population exposure received from just the variation in natural background radiation, which overshadows the potential population exposure from any accident considered. ,

c. Most of the radiation dose levels are sb low as to be undetectable, even with the most sensitive modern radiation detection instruments. '

d. When these potential exposures are considered in conjunction with their predicated frequencies of occurrence, it is judged that Class 1-8 accidents are small contributors to public risk. This judgment is based on the Reactor Safety Study (Ref. 7.1-20) and a published risk assessment of Class 3-8 accidents (Ref.

7.1-21). The Class 3-8 study estimated risk to the public using methodology that is similar to that used in the RSS. The results of the study showed that Class 3-8 accidents are small contributors to public risk relative j to postulated more severe accidents.

7.1.4 APPROACH TO THE ANALYSIS OF SEVERE ACCIDENTS l l

l l

This analysis is being provided at the request of the NRC staff ,

(EROL Questions E450.1, E450.2, E450.3 and E450.4) to help provide a response to the Statement of Interim Policy on severe '

accident considerations published by the NRC in the Federal Register on June 13, 1980 (45FR40101).

The analysis uses a comprehensive probabilistic risk assessment of the radiological consequences of accidents at the Limerick site. The assessment includes consideration of both internal and 7.1-23 Rev. 12, 04/83

l DRAFT l LGS EROL external initiators and specifically includes contributions from l internal events, earthquakes, and fires. Internal and external l flood, transportation, tornado, and turbine missile initiators were found to be noncontributors to risk. The analysis involves highly improbable sequences of failures that are more severe than those postulated for the design basis for protective systems and engineered safety features. The analysis treats the frequency of occurrence of these events in a systematic fashion and includes an assecsment of uncertainty in the frequencies, the phenomenological analysis, and the consequence' analysis. The

. focus of the presentation in this section is on the median results for the radiological consequences of the postulated events.

The fire analysis consisth of an estimate of the frequencies of fires in various rooms in the plant and modelM the effects of fires o various safety-related systems. The seismic analysis consis of a detailed study of the predicted characteristics of earthquakes at the Limerick site and of the response of structures and systems. The earthquakes J. t ;.; predicted to cause accidents at the Limerick plant,that are significant contributors to public risk are highly improbable and of a severity that has not occurred in the Limerick area in historical times. Given the occurrence of such an earthquake, it is highly likely that the public consequences of the earthquake itself directly on the surrounding area would be considerably more severe than the consequences of a seismically-induced accident at the plant.

Section 7.1.4.1 contains descriptions of the models and data

., employed in the analysis. Section 7.1.4.2 explains how the analysis was performed. The results are presented in Section l 7.1.4.3. Section 7.1.4.4 contains conclusions.

I

,.1.4.1 Models and Data l l Section 7.1.4.1.1 describes the fission product source terms and l their. associated frequencies. Section 7.1.4.1.2 contains a brief l

outline of the consequence model (the CRAC2 code) and the necessary input data. Section 7.1.4.1.3 discusses the l uncertainty analysis.

I 7.1.4.1.1 Source Term Description and Associated Frequencies l The magnitude and frequency of fission product source terms used in this assessment are given in Tables 7.1-22 and 7.1-23, Rev. 12, 04/83 7.1-24

DRAFT LGS EROL respectively. Source term is defined in this section to mean the magnitude of the release of fission products to the atmosphere, together with associated characteristics such as the time of release, warning time, duration of release, and rate of release of heat. These source terms have been selected to characterize the release anticipated from the various events analyzed in this section. These source terms tend to be conservative estimates that, for example, exclude deposition in the primary system and

, in the reactor enclosure. Detailed descriptions and the basis

for selection of these source terms is given in the Limerick Generating Station Severe Accident Risk Assessment (Ref. 7.1-22).

4

a. OXRE -- This source term includes the releases due to oxidation reactions that occur as a result of an in-vessel or ex-vessel steam explosion, or a hydrogen explosion following core melt. Fire is the most important contributor to this source term, contributing 55 percent of the point estimate frequency of 1.3x10-7 per year. -

y b. OPREL -- This source term is dominated by gross rupture of the containment, either as a result of the buildup of noncondensable gases or a hydrogen burn, following loss r

~

of coolant inventory, core melt and vessel repture Again, fires contribute most significantly to the point estimate frequency, given 55 percent of the total of 2.0x10-5 per year.

c. C47 -- This source term is for an ATWS sequence ending in gross rupture of the drywell. Seismic and internal initiators are roughly equal contributors, and the total point estimate frequency is 1.3x10-7 per reactor year,

d. C47' -- This source term is for an ATWS sequence ending in gross rupture of the wetwell, without loss of the suppression pool. Seismic and internal initiators are roughly equal contributors, and the total point estimate l frequency is 1.1x10-7 per reactor year.

e. C4r" -- This source term is for an ATWS sequence ending l in gross rupture of the wetwell, with loss of the l suppression pool. Seismic and internal initators are l roughly equal contributors, and the total point estimate frequency is 1.3x10-s per reactor year.

7.1-25 Rev. 12, 04/83

LGS EROL

f. C123r" -- This source term is for those sequences other than C4r" that result in a gross rupture of the containment in the wetwell with loss of the suppression pool. It has a total point estimate frequency of 1.0x10-* per year, to which fires contribute 58 percent.

g. LEAK 1 -- This source term is for core melt sequences in which the containment leaks relatively slowly without operation of the standby gas treatment system (SGTS).

The leakage sizes are smaller than for the y failure modes and preclude gross rupture. These sequences are small contributors to public risk. The most important initiator is fire, and the total point estimate frequency is 3.2x10-6 per year.

h. LEAK 2 -- This source term is for core melt sequences that are similar to those An LEAK 1 except that the SGTS

, is operating effectively. The most important initiator is fire, and the total point estimate frequency is l 1.8x10-s per reactor year.

i. R5 -- This source term includes the releases that result from the collapse of the reactor enclosure as a result of an earthquake. This leads to failure of the RHR heat exchanger 1steral supports, which is assumed to lead to failure of the attached piping leading from the suppression pool. The pool will drain down to the pipe, leading to an open containment while the core melts.

However, the suppression pool is still available for fission product scrubbing of the melt release of fissjon

~~' ~

.ornducts

j. VR -- This is a source term for the case in which the reactor vessel fails, and the containment fails shortly thereafter.

For internal events, this source term is caused by a spontaneous vessel rupture that can cause immediate containment failure. In this case, VR has a predicted point estimate frequency of 1.4x10-s per reactor year.

1 For earthquakes, this source term is dominated by events in which there is failure of the vessel upper lateral Rev. 12, 04/83 7.1-26

+

=

l l

LGS EROL DRAFT l 1

supports, causing rupture of the four main steam lines

while collapse of the reactor enclosure breaks pipework connected to the suppression pool (as in the case of source term RB). In this seismic case, VR has a predicted point estimate frequency of 3.7x10-7 per reactor year.

i

k. VRH2O -- This source term is also for the case in which the reactor vessel fails, and the containment fails 4 shortly thereafter. The only_ difference between this source term and VR is that, in the case of VRH20, sufficient water is assumed to remain in the bottom of tne vessel so that fission products are driven rapidly out into the atmosphere when molten core falls and causes the generation of steam. In the case of VR, the

vessel is assumed to be completely dry, and it takes a relatively long time to drive the fission products out into the atmosphere. For spontaneous (internal) vessel rupture, VRH2O has a point estimate frequency of 1.4x10-8 per reactor year. In the seismic case, VRH2O has a point estimate frequency of 4.1x10-s per reactor year.

The derivation of the point estimate frequencies is presented in Reference 7.1-22 and a discussion of the methods employed in the uncertainty evaluation of frequency is given in Section 7.1.4.1.3.1.

7.1.4.1.2 Consequence Model l The CRAC2 code was used to generate the complementary cumulative distribution functions (CCDFs) that are the final product of the analysis (Figures 7.1-2 to 7.1-6). The code is discussed in the PRA Procedures Guide (Ref. 7.1-23). A schematic outline of CRAC2 is given in Figure 7.1-1. Reference 7.1-23 should be consulted for discussion of such topics as exposure pathways, dosimetric and health effects models, and protective actions. Those parts of the input data or the coding that were modified to take account of Limerick specific features are discussed below.

7.1.4.1.2.1 Curies of Fission Products and Actinides in the Core at the Initiation of the Accident s

The amounts (curies) of each radionuclide released to the atmosphere for each accident sequence or release category is obtained by multiplying the release fractions specified in the 7.1-27 Rev. 12, 04/83

, DRAFT LGS EROL definition of the source term (Table 7.1-22) by the amounts that would be present in the core at the time of the hypothetical accident. These amounts are shown in Table 7.1-24 for the Limerick reactor.

7.1.4.1.2.2 Meteorological Data l The CRAC2 input data file for Limerick contains five years of consecutive hourly values of wind speed, wind direction, stability class, and precipitation intensity. These were processed from measurements taken at the Limerick site during the years 1972 to 1976.

These five years of data were processed by CRAC2 using the bin sampling technique. This required a minor code modification to enable CRAC2 to sample from the entire five years of data. The sampling techniques used by CRAC2 are described in Reference 7.1-23. The use of five years of data and the improved sampling techniques of CRAC2 yield a more complete and representative sample than has been possible using the " stratified sampling" techniques of CRAC. The data are consistent with those used and presented elsewhere in the EROL.

7.1.4.1.2.3 population Distributions l The population distribution around the site has been assigned to a grid consisting of 16 sectors, the first of which is centered on due north, the second on 22-1/2 degrees east of north, etc.

There are also 34 radial intervals (Table 7.1-24) that contain l the predicted permanent resident population for the year 2000.

l l

The population within 50 miles was taken from Tables 2.1-5 and 2.1-12 and assigned to the finer CRAC2 grid by ratioing by area.

In the 50 to 500 mile range, 1980 U.S. census data were used on a county-by-county basis, and 1981 Canadian census data were used in census tracts, which are comparable in size to U.S. counties. ,

The population within counties or tracts was again assigned to l the CRAC2 population grid by ratioing by area. Extrapolation to the year 2000 was done by using regional growth rates from the Census Department's Bureau of Economic Affairs, for the USA, and

! similar regional growth rates for Canada.

Rev. 12, 04/83 7.1-28

DRAFT LGS EROL 7.1.4.1.2.4 Evacuation Modeling and Other Protective Measures l The site-specific offsite emergency response plans are not complete at this time. Certain features of these plans, however, are considered to be sufficiently defined so as to be used in this analysis (e.g., 360-degree evacuation of the EPZ). These features were combined with a generic evacuation model, which was developed at Sandia Laboratories, on the basis of U.S. avacuation experience. It is described in the PRA Procedures Guide. This evacuation model is used with three alternative evacuation scenarios; 1, 3- or 5-hour delay times with relative probabilities of 30, 40 and 30 percent, and a subsequent evacuation speed of 10 mph (4.5 m/sec). This is considered to be a "best estimate" model.

The source terms considered in Tables 7.1-22 and 7.1-23 include some with contributions from earthquakes. For evacuation for these sequences, the model was modified to incorporate a 3-hour delay for th? whole population and an effective evacuation speed of 0.5 m/sec.

The "best estimate" model also includes an estimate of the response of people beyond the EP2 in the range 10 to 25 miles.

They are assumed to continue their normal activities for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the passage of the cloud, at which time they are rapidly relocated. In the event of an earthquake, this period is assumed to be 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Equivalent reductions in predicted dose'could be achieved by other countermeasures such as assuming that people shelter in their basements or large buildings for a day or two before relocating; that is, significant reductions in predicted dose could be achieved by a choice of simple countermeasures.

The outer limit of 25 miles is chosen because, in general, calculations with CRAC2 show that, even with conservative fission product source terms, life-threatening acute doses are rarely predicted beyond this distance, even in the most adverse of weather conditions.

7.1.4.1.2.5 Economic Costs l The necessary input to the calculation of economic costs in CRAC2 includes several unit costs such as the cost of evacuating or relocating a person and the. cost of decontaminating an acre of farm land or developed land. These costs are given in Reference 7.1-20 and have been updated to 1980 to allow for inflation. In addition, land use statistics, farm land values, farm product values, dairy production, and growing season information are required by CRAC2. These statistics are provided on a county-7.1-29 Rev. 12, 04/83

DRAFT LGS EROL wide basis within 50 miles and on a state-wide basis for larger distances. The various economic inputs are tabulated in Reference 7.1-22.

7-1.4.1.3 Uncertainty l Reference 7.1-23 lists 51 modeling assumptions or parameter variations to which the complementary cumulative distribution functions (CCDFs) may be sensitive. HoweVer, an uncertainty analysis taking account of all 51 parameters would be rohibitively time consuming. Instead, four major sources of p'ncertainty u were chose; (a) the frequencies'of the source terms given in Table 7.1-23; (b) the magnitude and associated characteristigs of the source terms; (c) the evacuation and sheltering modeling; and (d) the modeling of. health effects.

Consideration of this' limited set of uncertainties is sufficient to establish plausible bounds on the CCDFs; that is, more detailed uncertainty analysis would not be expected to produce results that are likely to lie outside the bounds established by the more limited uncertainty analysis. Justification for this view is given in Reference 7.1-22.

7.1.4.1.3.1 Uncertainty in Frequencies l Probability distributions on the frequencies of the source terms contributing to the various results were constructed. For accident sequences originating from internal and seismic j initiating. events, distributions were obtained by propagating l uncertainties on input parameters to the fault tree and event tree analyses through the algebraic expressions for accident l class frequencies in terms of those parameters, using Monte Carlo i methods. The distributions on the input parameters were assigned in a manner that follows currently accepted practice as described, for example, in Reference 7.1-23. For initiating events originating from fires in the plant, the probability l distribution on accident class frequency was constructed on the basis of a sensitivi.ty analysis of the more.important assumptions and parameters. They are discussed in detail and documented in Reference 7.1-22.

i 7.1.4.1.3.2 Uncertainty in Source Terms l One of the greatest sources of uncertainty in the CCDFs is the magnitude of the source terms. Sensitivity studies have been Rev. 12, 04/83 7.1-30 l

DRAFT LGS EROL carried out to determine the effect of a range of source term magnitudes and times of release for: (a) VR and VRH20; (b) C47, C4r' and C4r" (both seismic and internal); (c) OPREL (latent effects only); and (d) RB. These source terms were chosen because, on the basis of runs of CRAC2 carried out with the source terms and point estimate frequencies given in Table 7.1-23, it was established that they represent the major contributors to public risk. Details of these sensitivity studies and their effect on the CCDFs are provided in Reference 7.1-22.

7.1.4.1.3.3 Uncertainty in Evacuation and Sheltering l The CCDF for early fatalities is particularly sensitive to the choice of evacuation delay time (Ref. 7.1-23). Se'nsitivity studies were carried out in which they delay time was varied from 1 to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The evacuation velocity was varied from 2.5 to 10 mph. For seismically initiated sequences, it was assumed for the sensitivity study that evacuation assumptions would be unaffected.

The 10 to 25 mile sheltering assumptions were changed to simulate sheltering in basements for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , followed by rapid relocation. In addition, the outer 25 mile radius was changed to 50 miles.

The effect that these variations have on CCDFs is described in Reference 7.1-22.

7.1.4.1.3.4 Uncertainty in Health Effects Modeling l For early fatalitiefs, Reference 7.1-20 provides dose-response relationships for minimal, supportive, and heroic medical treatment. In the sensitivity analysis, each of these was chosen in turn. The standard dose-response relationship used for latent cancers in CRAC2, the central estimate, was varied to allow the simple linear dose-response relationship. The effect that these variations have on the CCDFs is described in Reference 7.1-22.

7.1.4.2 Analysis l The first step in the analysis was to use the point estimate source terms and point estimate frequencies in Tables 7.1-22 and 7.1-23, respectively, in CRAC2 and to produce a single CCDF for 7.1-31 Rev. 12, 04/83

LGS EROL DRAFT each health or economic effect. This single CCDF is called

" point estimate" because it is obtained using single or point estimates of each of the important input parameters. For each health or economic effect, the significant contributors to risk, determined by comparing the size of each contributor to the area under the point estimate CCDFs, were (a) VR and VRH20; (b) RB; (c) C47, C4r' and C4r"; and (d) OPREL (latent effects only). .

In the second step, an uncertainty analysis of the frequency of each source term was carried out as described in Section 7.1.4.1.3.1.

The third step was to establish a range of conditional CCDFs for each source term and each of the health or economic effects that are being considered. Upper and lower estimates on this range were taken as upper and lower percentiles on a lognormal distribution. The upper percentiles were chosen as the 95th or .

99th, depending on how likely the estimates are expected to be, and the lower estimate was chosen to be the 5th percentile. This is sufficent to fix the two independent parameters in the lognormal distribution.

The fourth step was to use this lognormal distribution in combination with the uncertainty distribution on frequencies to given an overall uncertainty distribution on the CCDFs. The uncertainty distributions are presented in Reference 7.1-22.

The final step was to extract from the uncertainty distribution the medians that are presented in Section 7.1.4.3. ;

7.1.4.3 Results l The results of the analysis are given in Figures 7.1-2 to 7.1-7 and in Table 7.1-26. These results give the total contribution from all source terms for seismic, internal, and fire initiators.

The CCDFs for individual source terms, as well as upper and lower estimates and point estimates, are given in Reference 7.1-22.

All of the results presented here are median CCDFs.

7.1.4.3.1 CCDFs l Figure 7.1-2 contains the median CCDF for the number of people receiving a bane marrow dose in excess of 200 rems from early exposure. (Early exposure is confined to that portion of the Rev. 12, 04/83 7.1-32

DRAFT LGS EROL radiation dose that is accumulated within 7 days, due to inhalation of radioactive materials, cloudshine and groundshine.)

This level of dose roughly corresponds to a need for hospital treatment.

Figure 7.1-3 shows the median CCDF for the total population exposure in person-rems for the population out to 500 miles (that is, the probability per reactor year that the total population exposure will equal or exceed the values given). The figure also gives a similar CCDF for the population within 50 miles.

Figure 7.1-4 shows the median CCDF for acute fatalities, representing radiation injuries that would produce fatalities within about one year after exposure.

Figure 7.1-5 gives the median CCDFs for latent cancer fatalities.

CCDFs for the total population and the population within 80 km (50 miles) are shown separately, and the latent cancers have been subdivided into that attributable to exposures of the thyroid and all other organs.

Figure 7.1-6 shows the CCDF for ex-plant costs in 1980 dollars.

In general, these costs are dominated by decontamination of urban or agriculatural land. Additional economic costs include decontamination of the facility itself and the cost of replacement power. These impacts are discussed in Section 7.1.4.3.2. section on risk considerations below.

7.1.4.3.2 Risk Considerations l The foregoing discussions have dealt with both the frequency (or likelihood of occurrence) of accidents and their impacts (or consequences). Because the ranges of both factors are broad, it is also useful to combine them to obtain average measures of environmental risk. Such averages can be particularly useful as an aid to the comparison of radiological risks associated with accidental releases, or those arising from other accidents.

A common way in which this combination of factors is used to estimate risk is to multiply the frequencies by the consequences.

The resultant risk is then expressed as the number of consequence expected per unit time. Table 7.1-26 shows average values of risk associated with population dose, acute fatalities, latent fatalities, and costs for protective actions and decontamination.

These average values are obtained by summing the frequency -

7.1-33 Rev. 12, 04/83

DRAFT LGS EROL multiplied by the consequences over the entire range of the median CCDFs. They are equal to the areas under the corresponding CCDFs. Because the probabilities are on a per-reactor-year basis, the averages shown are also on a per-reactor-year basis.

The acute fatality risk of 4.1x10-5 deaths per reactor year at the median level may be put into perspective by noting that 60 fatalities from motor vheicle accidents, 24 from falls, 8 from burns, and 3 from firearms are likely to occur each year within 10 miles of the plant. These figures are based on U.S. averages.

The indivi M $ risk of acute fatality as a function of distance is displayeT on Figure 7.1-7. The risk to the average individual living within one mile of the site boundary is 2.2x10-' per reactor year. This risk is small. For comparison, the following risks of fatality per year to an individual living in the United States may be noted; 2.2x10-* per year from automobile accidents and 1.2x10-5 per year from firearms.

The average population exposure is 70 person-rem per reactor year. This value may be compared with the annual average population exposures from routine operation given in Tables 5.2-15 and 5.2-17.

The average number of latent cancer fatalities (summing those due to thyroid dose and those in all other organs) within the population to 500 miles is 0.013 per reactor year. The equivalent average latent cancer fatalities for the population within 50 miles is 0.008 per reactor year. These figures may be put in perspective by noting that, in the population of 8,100,000 that is predicted to live within 50 miles of the Limerick reactor in the year 2000, there will be about 20,000 cancer fatalities per year from all causes. This figure was obtained by multiplying the figure for the population within 50 miles by 2.5x10-3, which, according to the Statistical Abstract of the United States, is the chance per year that an individual will die of cancer.

The ex-plant economic risk, in 1980 dollars, associated with the Limerick Generating Station is predicated to be $6,000 per reactor year at the median level. This figure is small compared with the estimated property damage caused by other accidents within 50 miles of the Limerick site (e.g., of the order of $10 million per year for automobile accidents. This figure is based on U.S. average statistics).

Rev. 12, 04/83 7.1-34 L . _

DRAFT LGS EROL There are other economic impacts and risks that are not included in the calculations discussed above. These costs would be for decontamination and repair or replacement of the facility, and for replacement power. Experience with such costs is currently being accumulated as a result of the Three Mile Island accident.

It is already clear that such costs can equal or exceed the original capital cost. The cost for decontamination and restoration is in the region of $2 billion. Replacement power costs for two units at the Limerick site are estimated at $580 million per year. If it is assumed that both units on the site are out of operation for 8 years, the total cost of the accident would be $6.64 billion. The accident sequences considered in this report and shown in Table 7.1-22 would all lead to core melt and would in turn lead to costs of the size described above. The predicted median frequency of core melt is 3.0x10-5 per year so that the economic risk due to the accident sequences considered in this report is predicted to be $200,000 per year. This estimate is in 1980 dollars.

7.1.4.4 Conclusions l The previous sections consider the potential environmental impacts of severe accidents at the Limerick facility. These have covered a broad spectrum of hypothetical accidental releases and a range of possible health and economic impacts. The comparisons in the section on risk considerations show that the public risk associated with these impacts is small.

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7.1-35 Rev. 12, 04/83 l

DRAFT LGS EROL i

7.

1.5 REFERENCES

7.1-1 NRC Regulatory Guide, 4.2, Rev 2, Preparation of Environmental Reports for Nuclear Power Stations, Nuclear Regulatory Commission, Office of Standards Development, Washington, D.C. (July 1976).

7.1-2 NRC Regulatory Guide 1.3, Assumptions Used for Evaluatino the Potential Radiological Consequences of a Loss of Coolant Accident for Boi Lina Water Reactors 7 NRC Washington, D.C. (June 27, 1974J,.

7.1-3 NRC Regulatory Guide 1.25, Assumptions Used for ,

Evaluatino the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handlino and Storace Facility for Boilina an3 Pressurized Water Reactors, NRC, Washington, D.C. (March 23, 1972).

7.1-4 Report of ICRP Committee II, " Permissible Dose for Internal Radiation (1959)," Health Physics, 3:30 (1969) pp 146-153.

I l' 7.1-5 Slade, D.H., " Meteorology and Atomic Energy," AEC Report Number TID-24190, AEC, Washington D.C.-(January 1969).

7.1-6 Meek, J.E., and Rider, B.F., " Summary of Fission Product Yields for U-235, Pu-239, and Pu-241, at Thermal, Fission Spectrum and 14 Mev Neutron Energies," Report l Number APEC-3398 (March 1, 1964).

l 7.1-7 DiNunno, J.J., et al, " Calculation of Distance Factors i

for Power and Test Reactor Sites," AEC Report Number TID-14844, AEC, Washington, D.C. (March 1, 1962).

7.1-8 Blomeke, J.O., and Todd, M.F., " Uranium-235 Fission-Product Production as a Function of Thermal Neutron Flux, Irradiation Time, and Decay Time," AEC Report Number ORNL-2127, AEC, Washington, D.C. (August 19, 1957).

7.1-9 Lederer, C.M., et al, Table gf Isotopes, 6th edition (1968).

I Rev. 12, 04/83 7.1-36

LGS EROL DRAFT 7.1-10 ICRP Publication 2, Report of Committee II, Permissible Dose for Internal Radiation (1959).

7.1-11 Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boilina Water Reactors, NUREG-0016, U.S. Nuclear Regulatory Commission, Washington D.C. (April 1976).

/

145 7.1-12 NRC Regulatory Guide 1 NRK, Atmospheric Dispersion Models for Potential Accident Consecuence Assessments at Nuclear Power Plants (Draft), NRC, Washington, D.C.

(September 23, 1977).

7.1-13 NRC Regulatory Guide 1.111, Methods for Estimatina Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Licht-Water-Cooled Reactors, NRC, Washington, D.C. (1977).

7.1-14 NRC Regulatory Guide 1.23, Onsite Meteorolooical Procrams, NRC, Washington, D.C. (1972).

7.1-15 Pasquill, F., "The Estimation of the Dispersion of Windborne Material," Meteoroloov Macazine, Vol 90, (1961) pp 33-49.

7.1-16 Gifford, F.A., "Uses of Routine Meteorological Observations for Estimating Atmospheric Dispersion,"

Nuclear Safety, Vol 2, (1961) pp 47-51.

7.1-17 Eimutis, E.C., and Konicek, M.G., " Derivations of Continuous Functions for the Lateral and Vertical Atmospheric Dispersion Coefficients," Atmos Environ, Vol 6, (1972) pp 859-863.

7.1-18 Busse, A.D., and Zimmerman, J.R., " Users Guide to the Climatological Dispersion Model," EPA Report Rr-73-024, EPA, Washington, D.C. (1973).

1 7.1-37 Rev. 12, 04/83

DRAFT LGS EROL 7.1-19 Markee, E.G., and Levine, J.R., "Probabilistic Evaluations of Atmospheric Diffusion Conditions for Nuclear Facility Design and Siting," Proceedinas of the American Meteorolocical Society Conference on Probability and Statistics in Atmospheric Sciences, Las Vegas, Nevada (1977) pp 146-150.

7.1-20 U.S. Nuclear Regulatory Commission, Reactor Safety Study, An Assessment of Accident Risk in U.S. Commercial Nuclear Power Plants (WASH-1400, NUREG-75/014), 1975.

7.1-21 R.E. Hall, et.al., A Risk Assessment of a Pressurized Water Reactor for Class 3-8 Accidents, Brookhaven National Laboratory, 1979.

7.1-22 NUS Corporation, Limerick Generatino Station Severe Accident Risk Assessment, NUS-4161, 1983. -

7.1-23 U.S. Nuclear Regulatory Commission, PRA Procedures Guide; A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, NUREG/CR-2300, 1983.

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e

}

SOURCE TERM CHARAt CRAC 2 Input T (2) T (3) T (*) h(5) Oca) X_E r a w GROUP (br) (hr) (hr) (m) (cal /sec)

OXRE 4.0 0.5 3.0 27 8. 4 (6) < 7 ) 1.t OPREL 7.0 2.0 6.0 27 8. 4 (6) 1.1 C4y 1.5 2.0 1.0 27 7.0(4) 1.(

C4y' 1. 5 2.0 1.0 27 7.0(4) 1.<

C4y" 1. 5 2.0 1.0 10 7.0(4) 1.t C123y" 7.0 2.0 6.0 10 7.0(4) 1.4 LEAK 1 7. 0 ?.0 6.0 27 7.0(4) 0.'

LZAK 2 7.0 2.0 6.0 27 7.0(4) O.

RBce) 1.5 3.0 1.5 10 8. 4 (6) 1.4 VRC') 0.25 3.5 0.25 10 1. 4 (4) 1.1 VRH20( 10 ) 0.34 0.65 0.34 10 2 (6) 1.(

C1) The final CCDFs given in Figures 7.1-2 through on the source term characteristics.

(2) T = time of release r

(3) T = duration or release d

(*) T = warning time w

C5) h = height of release ca) Q = rate of release of energy (7) 8. 4 (6) = 8.4 x 106 ce) Reactor building failure (9) Vessel rupture without water in vessel (1o) vessel rupture with water in vessel v

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LGS EROL ABLE 7.1-22 ,

AERISTICS - POINT ESTIMATEC 1)

RADIONUCLIDE RELEASE FRACTIONS QI Il Cs le jr Ru La 3 (-4) 0.20 0. O.50 0.007 0.40 1. 0 (-5) 3 (-4) 0.11 0.09 0.016 0.01 3 (-3) 3 (-4) 3 (-4) 0.261 0.202 0.434 0.029 0.095 5. 2 (-3) 3 (-4) 0.07 0.09 0.20 0.016 0.008 5. 0 (-3) 3 (-4) 0.73 0.70 0.55 0.09 0.12 7. 0 (-3) 3 (-4) 0.13 0.17 0.50 0.02 0.08 6. 2 (-3) 3 3 (-4) 1. 9 (-2) 9. 8 (-3) 4. 6 (-2) 1. 6 (-3) 3. 2 (-3) 5. 8 (-4) 3 3 (-4) 2. 7 (-3) 9.8(-5) 4. 6 (-4) 1. 6 (-5) 3. 2 (-5) 5. 8 (-6) 3 (-4) 0.05 0.09 0.09 4. 0 (-3) 0.02 5.0(-3) 3 (-4) 0.1 0.33 0.33 0.15 0.04 0.02 3 (-4) 0.5 0.73 0.75 0.35 0.07 0.05 1.2-6 are medians and are obtained from an uncertainty analysis l

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FREQUENCIES OF TJ CRAC 2 POINT ESTIMATE (YR-1)

INPUT INTERNAL jEIg[qC FIRJ GROUP OXRE 4. 4 (-8) 1. 3 (- 8) 6. 9 (-!

OPREL 7. 0 (-6) 2. 0 (-6) 1.1 (-!

c4y 6. 4 (-8) 6.3 (-8) O c4y' 5. 6 (-8) 5. 6 (-8) O c4y" 6. 4 (-9) 6. 3 (-9) O c123y" 3. 6 (-7) 1. 0 (-7) 5. 8 (-1 LEAK 1 1.1 (-6) 3. 3 (-7) 1. 8 (-f LEAK 2 6.1 (-6) 1. 7 (-6) 9.9(-f RB 0 1. 2 (-6) 0 VR 1. 4 (-8) 3. 7 (-7) 0 VRH2O 1. 4 (-8) 4.1 (-8) 0 I

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@S EROL h j(

iLE 7.1-23 5 3BLE 7.1-22 SOURCE TERMS MEDIAN (YR-1) j I!TTERNAL SEISMIC FIRE

!) 3. 3 (-8) 7. 5 (-10 ) 2. 6 (- 8)

>) 5.3 (-6) 1. 2 (-7) 4. 2 (-6)

6. 4 (-8) 2. 0 (-9) 0

5. 6 (-8) 9. 0 (-10) 0

6. 2 (-9) 1. 0 (-10) 0

) 2. 8 (-7) 6. 3 (-9) 2. 2 (-7 )

) 8. 8 (-7) 2.0(-8) 6. 8 (-7)

) 4. 6 (-6) 1.1(-7) 3. 7 (-6) 0 7. 6 (-9) 0

5. 0 (-9) < 1 (-10) 0

5. 0 (-9) < 1 (- 10 ) 0 i

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Rev. 12, 04/83

LGS EROL TABLE 7.1-24 (Page 1 of 2)

ACTIVITY IN THE LIMERICK REACTOR CORE AT 3293 MWt Radioactive inventory Half-life Group /radionuclide (million of Curies) (days)

NOBLE GASES Krypton-85 0.57 3,950 Krypton-85m 28 0.183 Krypton-87 55 0.0528 Krypton-88 77 0.117 Xenon-133 184 5.28 Xenon-135 34 0.384 ,

IODINES

Iodine-131 83 8.05 Iodine-132 128 0.0958 Iodine-133 183 0.875 Iodine-134 202 0.0366 Iodine-135 172 0.280 ALKALI METALS Rubidium-86 0.061 18.7 Cesium-134 5.7 750 Cesium-136 1.9 13.0 Cesium-137 5.6 11,000 TELLURIUM-ANTIMONY Tellurium-127 5.8 0.391 Tellurium-127m 0.79 109 Tellurium-129 21.8 0.048 Tellurium-129m 5.8 34.0 Tellurium-131m 11.4 1.25 Tellurium-132 122 3.25 Antimony-127 6.0 3.88 Antimony-129 23.2 0.179 AKALINE EARTHS Strontium-89 102 52.1 Strontium-90 4.8 10,300 Strontium-91 130 0.403 Barium-140 163 12.8 Rev. 12, 04/83

i LGS EROL TABLE 7.1-24 (Cont'd) (Page 2 of 2) f Radioactive inventory Half-life Group /radionuclide (million of Curies) (days)

COBALT AND NOBLE METALS Cobalt-58 0.0 71.0 Cobalt-60 0.0 1,920 Molybdenum-99 166 2.80 Technitium-99m 143 0.25 Ruthenium-103 114 39.5 Ruthenium-105 67 0.185 Ruthenium-106 42 366 Rhodium-105 60 1.5 RARE EARTHS, REFRACTORY OXIDES AND TRANSURANICS Yttrium-90 504 2.67 Yttrium-91 127 59.0 Zirconium-95 152 65.2 Zirconium-97 156 0.71 Niobium-95 145 35.0 Lanthanum-140 166 1.67 Cerium-141 151 32.3 Cerium-143 148 1.38 Cerium-144 90 284 Praseodymium-143 147 13.7 Neodymium-147 61 11.1 i Neptunium-239 1,670 2.35 Plutonium-238 0.036 32,500 Plutonium-239 0.02 8.9x10*

Plutonium-240 0.024 2.5x10*

Plutonium-241 5.5 5.350 Americium-241 0.0034 1.6x105 Curium-242 1.1 163 Curium-244 0.013 6,630 Note: The above grouping of radionuclides corresponds to that in the Reactor Safety Study i

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~.

k LGS 1 TABLE 1 PERMANENT RESIDENT POPULA1 S rrt or Dire smiles) N NNE NE EAE F ESL 51 55l 0-0.5 0 0 0 0 0 0 0 0 0.5-1.0 61 141 27 32 23 73 0 18 1.0-1.5 308 110 40 38 60 70 222 - 204

1. 5- 2 .0 432 153 55 53 84 97 311 286 2.0-2.5 243 151 128 159 183 192 675 949

2. 5- 3.0 297 184 157 194 223 234 826 1.160 3.0-3.5 316 214 191 218 281 172 2.622 2.669 1.5-4.0 365 246 220 252 325 199 3.025 3,079 4.0-4.5 472 92 187 109 227- 198 745 1.126 4.5-5.0 527 102 210 121 253 221 833 1.258 5-6 1.306 585 5%9 345 2.248 2.692 598 5.913 6-7 1.544 691 660 407 2.657 3.182 707 6.989 7-8.5 2.761 1.236 1.181 729 4.751 5.691 1.264 12.499 8.5-10 3.295 1.476 1.410 870 5.671 6.792 1.508 14.918 i 10-12.5 1,280 4.739 6.146 9,828 12.472 31.605 21.922 7,194 l 12.5-15 1.565 5.792 7.512 12.012 15.243 38.629 26.794 8.792 l 15-17.5 1.850 6.845 8.877 14,197 18.014' 45.652 31,666 10.391 l

. 17.%-20 2.134 7.897 10,243 16.381 20.786 52.675 36.537 11,990 l 20-25 20.829 97.040 10.711 27.827 63.046 336.450 563.411. 121.367 25-30 25.457 118.604 13.091 34,010 77.056 411.217 688.613 148,337 l 30-35 21.716 85.094 14,733 11,780 122.464 324.681 336.351 16,314 l 35-40 2).057 90.186 16.999 13,592 141.305 374.632 388.097 18.823 l 40-45 11.888 17.743 24.911 18.800 225.218 49.936 67.649 13,997 4%-50 13.286 19.811 27.841 21.011 251.715 55.811 75.607 15.643 50-55 6.886 32.970 30.854 53.592 187,792 49.511 161,447 30,528 55-60 17.057 16.913 64.100 105.293 174,828 59,913 102.131 39.055 60-65 10.623 17.742 66.292 171.163 162.803 72.760 47.391 34,900 65-70 35.151 15.206 62.170 2723048 160,844 78.672 55,195 32.108 70 85 155.810 36.828 296.821 2.001.226 351.491 177.523 128,551 69.479 85-100 113.867 53.596 456.449 6.070.018 0 0 0 27,737 100-150 271.093 258.729 1.244.443 5.114.585 0 0 0 8.231 150-200 4R2.H02 568.895 1.151.835 1.802.514 0 0 0 0 200-350 1.6S0.580 1,194.147 3.569.922 4.813.485 0 0 0 0 350-500 819.581 5,136.991 Y49. 375 0 0 0 0 0 1

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.1-25

' ION FOR THE LIMERICK SITE i

ctinn 5 55w Sw W5W k WN= hm NNh 0 0 0 0 0 0 0 0 5 0 88 67 65 10 24 11 259 305 123 136 33 45 199 317 362 427 173 190 46 63 278 444 175 207 117 339 715 2.083 1,517 565 213 254 142 414 874 2.5a6 1,878 690 50 232 208 293 740 7,310 4,029 572 57 268 239 339 853 8.435 4,648 661 253 168 200 713 1,197 2,232 960 434 283 187 223 796 1,337 2.494 1,073 486

- 944 472 745 261 60 1,724 164 1.032 1.115 558 880 309 70 2.038 193 1.219 1,995 998 1,574 552 126 3,645 346 2.181 2,381 1,191 1,879 659 150 4.150 413 2,603 17,907 8,376 1,211 2.068 737 24.907 1.578 1,986 21,887 10.237 1.481 2.528 901 30.442 1,928 2,428 25.866 12.098 1,750 2,987 1,065 35.976 2,279 2,869 29,846 13,960 2,019 3.447 1,229 41,511 2,629 3,310 17,609 17,078 23,839 10,670 8,012 34,626 8,212 7,096 21.523 20.873 29,137 13,041 9,793 42,320 10.037 8,67) 202,552 24.450 6.281 34,785 23.142 9.433 6.615 2.663 233,714 28.212 7,247 40,136 26,703 10.884 7.632 3.072 11,712 32.128 9,777 71,801 -37,361 12,542 24.250 13,994 13,106 35,907 10.927 80.248 41,756 14.017 27,101 15.640 53,899 8,247 27,223 34,766 32.841 22.307 20,197 22.146 68.312 9.516 42,384 36.859 44,757 31.575 23.085 45,025' l 53,574 9,185 50,422 43.002 49,577 45.250 19,869 47,911 l18.010 5,570 40,968 58.190 46.595 42,210 21,365 39,144 l57,227 22.168 730.546 206,022 116,878 74.437 72.049 141,117 ,

l 61.571 43.637 942,506 101,937 110.012 41.010 59.92R 20,755 !

'209.523 362.871 2,739,529 1,062.112 218.115 295.032 140,775 182,565 52.287 166.772 329.908 287.951 520.117 164.714 142,422 324,640- 1

%2,89 ) 3.071.062 1.R79,393 1.030,760 4.504.704 5.42%.119 6,677,69) 2,105.064 1 31.9%9 2.036,392 4,558,30) 2,849,465 6.044.539 9,035,547 504.911 30s.549 l i

4 i 1 I l

l i

Rev. 12, 04/33 l

- i 4

LGS EROL TABLE 7.1-26 AVERAGE VALUES OF ENVIRONMENTAL RISKS DUE TO ACCIDENTS PER REACTOR-YEAR Environmental Risk Average /RY (Median)

Population exposure Person-rems within 50 miles 40 Total person-rems 70 Acute fatalities 4.1 x 10-s Latent cancer fatalities

All organs excluding thyroid 0.012 Thyroid only 0.001 Cost of protective actions $6,000 and decontamination l

Rev. 12, 04/83

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LGS Egot OUESTION E450.1 In accordance with NRC's Interim Policy (45FR40101) revise Section 7.111 to include a probabilistic evaluation of impacts of accidents including those formerly called Class 9 accidents.

RESPONSE

The requested evaluation is provided in Section 7.1.4. l E450.1-1 Rev. 12, 04/83

LGS EROL OUESTION E450.2 Please provide your assessment of accidents formerly clagsified as classes 3-8.

RESPONSE

The assessment of accidents formerly classified as classes 3-8 is provided in Section 7.1.

I l

E450.2-1 Rev. 12, 04/83

LGS EROL QUESTION E450.3 Figures 7.1-1 and 7.1-2 (the two CCDFs) have been superseded by subsequent PRA revisions; and therefore are no longer valid.

Please provide updated information.

RESPONSE

on 7.1.4. 6 i

E450.3-1 Rev. 12, 04/83 l

t . . - -

LGS EROL OUESTION E450.4 Please provide information on the following specific items you consider appropriate to your PRA which is now recognized as part of the ER-OL and bases therefore;

a. population distribution for the plant mid-life years;

b. site specific off-site emergency response parameters such as delay time before evacuation, evacuation speed,

evacuation distance etc.;

c. site-specific land-use and economic data;

d. assumption of the availability of supportive medical treatment to highly exposed individuals to reduce early fatality;

e. other categories of consequences and risk such as:

1. delayed cancer fatality within 50-mile

11. person rems )

) within the 50-mile 111. thyroid effects )

) and the entire regions iv. genetic defects )

v. offsite and onsite property damage vi. risks to individuals as functions of distance from the reactor, or individual risks isopleths;

f. liquid pathway considerations; and

! g. comparison of risks from accidents with those from plant j operation.

RESPONSE

l The requested information is provided in Section 7.1.4, with the exception of Jtem f which is provided in the response to Question E240.21.

l I

E450.4-1 Rev. 12, 04/83 i

ML20072M690

Text

Subject:

Reference:

Subject:

Dear Mr. Schwencer:

1.5 REFERENCES

RESPONSE

RESPONSE

RESPONSE

RESPONSE

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