ML17229A298

Final Analysis of Radiological Consequences of Main Steam Line Break Outside Containment for St Lucie Unit 1 NPP Using NUREG-0800 Std Review Plan 15.1.5 App A.

ML17229A298
Person / Time
Site:	Saint Lucie
Issue date:	03/06/1997
From:	Hannaman G, Karimi R, Otis M SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:
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ML17229A297	List: ML17229A296 ML17229A298
References
RTR-NUREG-0800, RTR-NUREG-800 SAIC-97-1008, NUDOCS 9704100061
Download: ML17229A298 (51)

Text

'I SAIC-97/1008 r

ANALYSIS OF THE RADIOLOGICALCONSEQUENCES OF A MAINSTEAM LINE BREAK OUTSIDE CONTAINMENTFOR THE ST. LUCIE UIAT 1 NUCLEAR POWER PLANT USING NUREG'-0800 STANDARD REVIEW PLAN 15.1.5 APPENDIX A Prepared for:

APTE<CH Engineering Services, Inc.

Pittsburgh, Pennsylvania Prepared by:

Science Applications International Corporation Germantown, Maryland and Reston, Virginia Roy Karimi, Sc.D, Steven . MirSk, P.E., and Joseph D. Price, PhD f 4% ~ h '4 i P~

Reviewed by:

Mark Otis Approved by:

G.W. Hannaman March 14, 1997 FINAL 9704100061 970404 PDR ADQCK 05000335 P PDR

r TABLE OF CONTENTS

1.0 INTRODUCTION

~ ..

2.0 INPUT DATA ..

3.0 ASSUMPTIONS 4.0 CALCULATIONS .6 4.1 'ite Boundary and Low Population Zone Dose Calculations .6 4.1.1 Pre-existing Transient with an I-131 PCS Equilibrium Concentration of 60 pCi/g .. ~ ~ ~ ~ ~ ~ ~ 7 4.1.2 Concurrent Iodine Spike with the Outside Containment Main Steam Line Break Accident . .8 4.1.3 Evaluation of Effects of Fuel Failure after the MSLB Accident . ~ ~ ~ ~ ~ ~ ~ 9 4.2 Control Room Dose Calculations 10 4.2.1 Methods of Control Room Analysis .. 10 4.2.2 Control Room Source Term Characterization . 13 4.2.3 Results of the Control Room Dose Analysis .. ~ ~ ~ ~ o 13 5.0

SUMMARY

AND CONCLUSIONS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 21

6.0 REFERENCES

23 Appendix A Memorandum NF-97-065 from J.N. Kabadi (FPkL) to G.L. Boyers(FPAL),

"Initial Steam Generator Mass for St. Lucie Unit 1 MSLB Dose Calculations", February 18, 1997 . ~ A-1 Appendix B Florida Power and Light Letter ENG-SPSL-97-0068, to Steven Mirsky (SAIC) from Carl Bible (FP8rL), "St. Lucie Unit 1 Transmittal of Review &

Verification of Values 8t: Input Parameters - File: Engineering Evaluation JPN-PSL-SESS-96-076 Rev. 0", dated February 26, 1997 .......... ~.... B-1 ~

Appendix C Facsimile from J. Kabadi and Chris Buehrig of Florida Power and Light to Steve Mirsky at SAIC, "ANF-88-113(P), St. Lucie Unit 1 Assessment of Radiological and Rod Bow Effects for Increased Burnup", July 1988, Advanced Nuclear Fuels Corp., March 11, 1997 C-1

LIST OF FIGURE Figure 1 Control Room Habitability System Model

LIST OF TABLES Table 1 List and Values of Input Parameters Table 2 Control Room Analysis Iodine Species Release Quantities Table 3 Control Room Analysis Noble Gas Release Quantities Table 4 Control Room Analysis for 1 gpm Steam Generator Tube Leak Rate......

Table 5 Control Room Dose Analysis Results ..

1.0 INTRODUCTION

SAIC has been contracted by APTECH Engineering Services, Inc. (APTECH) to perform a licensing analysis of the radiological consequences of an unisolable postulated main steam line break (MSLB) outside containment accident at the St. Lucie 1 Nuclear Power Plant. This analysis was performed in accordance with NUREG-0800, the U.S. Nuclear Regulatory Commission (USNRC) Standard Review Plan (SRP) Section 15.1.5 Appendix A, "Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR", Revision 2, July, 1981."'he purpose of this analysis is to calculate whole body and thyroid doses to the site boundary (or EAB), low population zone, and occupants in the control room resulting from radionuclide releases during the postulated MSLB accident outside containment. In addition, skin doses to control room occupants were also calculated. These doses are calculated using suitable conservative licensing assumptions as delineated in the SRP and the St. Lucie Final Safety Analysis Report (FSAR) and presented in terms of the St. Lucie 1 steam generator tube primary-to-secondary side leak rate.

2.0 INPUT DATA The input data for this analysis was developed by reviewing and evaluating information available in the St. Lucie Unit 1 Final Safety Analysis Report (FSAR)" and its Technical Specifications,"

along with the FSAR for St. Lucie Unit 2."'n addition, The NRC Standard Review Plan (NUREG-0800) 15.1.5 Appendix A"'as used for the determination of iodine spiking effects.

The FP&L engineering staff provided some data directly and confirmed the appropriateness and applicability of all input parameter values used in this analysis."" All input data and source for the data is given in Table 1.

Table 1 List and Values of Input Parameters Parameters Numerical Values Reference atmospheric dispersion factor to site boundary 8.55E-5 (0-2 hour) sec./m'.97E-6 atmospheric dispersion factor to low population zone sec./m'.47E-4 breathing rate m~/sec.

steam generator hot full power secondary side 127,602 pounds water mass Primary coolant system (PCS) water volume 10,400 ft 1,3 primary coolant system water mass 2.13E+8 grams calculated time to MSIV closure after MSLB 70 sec. 1,2 I-131 Thyroid dose factor 1.08E+6 rem/Ci time to SIAS after the MSLB 66.1 sec.

time to shutdown cooling condition after MSLB 12,240 sec.

Limit on SG secondary coolant I-131 0.1E-6 Ci/g concentration Iodine partition factor in SG and main steam line 1.0 assumed Noble gas partition factor in the SG and main 1.0 assumed steam line failed fuel followirigan MSLB

'ercent 2.0 .

%( Table 1 List and Values of Input Parameters (Continued)

Parameters Numerical Values Reference CVCS PCS let down flow rate 40 gpm CVCS I-131 decontamination factor 1,000 pre-existing PCS iodine concentration 60.0 NCi/g 3,4 maximum time period for a 60 pCi/g iodine 106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br /> 3 concentration control room volume 62,700 ft'50 normal control HVAC outside air intake flow cfm control room HVAC isolation damper closure 35 seconds time unfiltered air leakage into the control room 100 cfm control room recirculation flow rate 2,000 cfm atmospheric dispersion factor to control room 4.86E-4 sec./m'9 Control Room HEPA filter efficiency per cent P

Control room occupancy factor for 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 06 ~

96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 0.4 control room HVAC charcoal filter iodine removal efficiency Elemental iodine 95 per cent Organic iodine 95 per cent Particulate iodine 95 per cent Xe-133 PCS concentration 100/8 pCi/g I-131 PCS concentration technical specification 1 p Ci/g primary to secondary leak rate 1 gpm

3.0 ASSUMPTIONS The following conservative assumptions were made in performing this analysis in accordance with the licensing requirements set forth in NUREG-0800 SRP 15.1.5 Appendix A.

Three scenarios"'re evaluated for I-131(DEC) in the primary coolant system (PCS): (a) pre-existing equilibrium concentration of 60 pCi/g,"'b) MSLB accident induced a release rate spike of 500 times the release rate that corresponds with the technical specification limit of 1 pCi/g, and (c) the PCS concentration associated with the maximum MSLB FSAR calculated fuel failure (fuel failure is assumed for all fuel that is calculated to experience departure from nucleate boiling (DNB)).

2. Xe-133 PCS concentration is 100/E bar where E bar is the average Beta and Gamma energy of all noble gas radioisotopes present in the PCS.'"

3. For the pre-existing Iodine spike PCS scenario (assumption 1(a)), the maximum technical specification allowable time for this concentration to exist in the PCS prior to the plant being put into hot shutdown is 106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br />."'uring these 106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br />, all iodine introduced into the steam generator secondary coolant by the steam generator (SG) tube leakage in the SG secondary coolant water volume until the MSLB occurs. The iodinewill'ccumulate inventory also includes the initial concentration of 0.1 pCi/g of SG secondary water as .'.

specified in the FSAR."

At the time of the MSLB, all Iodine present in the entire SG water inventory willbe released to the atmosphere with no iodine removal within the SG internals or main steam line (i.e. iodine partition factor = 1.0).

5. The entire inventory of iodine present in both SGs is completely released directly to the atmosphere from the initiation of the MSLB to the time of MSIV closure.

After MSIV closure and up to'the time of cooldown to shutdown cooling system operation, iodine and noble gases continue to be released directly to the environment from the unisolable SG at the same rate as the tube leak rate with no iodine or noble gas removal inside the SG or main steam line piping. The tube leak rate is the same constant value throughout the accident scenario even though the PCS pressure is being reduced from 2250 psia to 275 psia which would cause a lower tube leak rate.

7. The primary to secondary leak through SG tubes occurs in the unisolable SG.

8. The Xe-133 (DEQ) noble gas release directly to the environment is at the same rate as the SG tube leak rate with no removal within the SG or main steam lines.

9. Xe-133 (DEQ) noble gas does not accumulate in the SG secondary system prior to the MSLB, but is continuously released by the condenser air ejectors because of its chemical form.

10. The Iodine PCS concentration for the scenario source term assumption 1(b) above does not credit any iodine removal by the CVCS system or radioactive decay after the iodine release rate is increased by a factor of 500.

11. The Iodine and noble gas PCS concentration for the scenario source term assumption 1(c) above was obtained from an AbP report provided by FPEcL"" coupled with the assumption that 2% of the core gap inventory was released to the PCS.

12. Although the LPZ atmospheric dispersion factor in the FSAR is for the 2 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, this same value is also applied to the 0 to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time period.

13. The control room unfiltered inleakage throughout the MSLB Accident is 100 CFM in accordance with FSAR"'ontrol room dose calculations.

4.0 CALCULATIONS ..

Dose impacts for the MSLB outside containment accident were evaluated for three cases characterized by: 1) a pre-existing iodine spike, 2) a concurrent iodine spike, and 3) accident initiated 2% fuel failure with no iodine spike. Hypothetical receptors at the exclusion area boundary, low population zone, and control room were considered.

4.1 Site Boundary and Low Population Zone Dose Calculations'he following equations were used to perform the thyroid and the external whole body dose calculations associated with the releases of I-131 and noble gases (in terms Xe-133 dose equivalent concentration) after a postulated outside containment main steam line break accident.

a. Thyroid dose from I-131 D,= [A,,] x [X/g] x [DCF...] x [BR]

b. External whole body dose from P particles and y rays"'~8

[0 25 Ef + 0 23 Q] [Ai33] x (X/g]

Where:

Thyroid dose from I-131 inhalation, (rem).

Ai3i Dose equivalent activity of released I-131, (Ci).

x/Q 0-2 hour dispersion coefficient for site boundary, or 0-8 hour dispersion coefficient for low population zone, (sec/m').

DCF f3) I-131 Thyroid dose conversion factor, (rem/Ci).

BR Breathing rate, (m'/sec).

Dw.s. Whole body dose (rem) from immersion in a semi-infinite cloud of Xe-133.

Average energy release by y decay (MEV/disintegration).

Average energy release by P decay (MEV/disintegration).

133 Dose equivalent noble gas (Xe-133) activity release (Ci).

The relevant data for each of the above parameters are given in Table 1. It was assumed that the total primary to secondary leak rate through steam generators is 1 gpm, (2702.8 g/min). [The PCS water specific density of 0.724 is based on 2250 Psia and 575'F.]

4.1.1 Pre-existing TransieJit with an I-131 PCS Equilibrium Concentration of 60 pCi/g In this scenario, it was assumed that I-131 concentration in the PCS has reached an equilibrium value of 60 pCi/g well before the MSLB accident. Based on the St. Lucie 1 technical specification limiting condition for operation,"'his condition could exist for about 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> before the plant is brought down to hot shut down condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The noble gas concentration is at the technical specification"'imit of 100/E pCi/g. Using the assumptions and the data listed in Section 3.0, and Table 1, the following doses where calculated.

a. Site boundary (exclusion area boundary)

Thyroid Dose I-131 activity release during the time period after MSLB to shutdown cooling condition:

Pre-break PCS to SCS I-131 activity transfer 2702.8 x 60x 10~x 106 x 60 = 1032.39 Ci initial activity of I-131 in the SCS, 2 x 127,602 x 453.6 x 0.1 x 10~ = 11.58 Ci w sum of I-131 released prior to MSIV closure 1032.39+ 11.58+ 2702.8 x 70/60 x 60 x 10~ = 1044.16 Iodine release after MSIV closure 2702.8 x 60 x 10~ x 202.83 = 32.89 Ci over remaining minutes (12,240/ 60-70/60).

D~ = [1044.16+ 32.90] x [8.55 x 10 ] x [1.08 x 10 ] x [3.47 x10 ]

= 34.51 rem

2. External whole body dose Noble gases are directly released to atmosphere after the break for 204 minutes.

~ 2702.8 x 100/8 x 10~ x 204 = 55.14/8 Ci The value of 8 for noble gases in the PCS was calculated to 0.23 MEV based on

. the concentration of the gases given in Table 12.1-3 and.the average energy per disintegration given in ICRP publication No. 38. The calculated beta and gamma average energy per disintegration were 0.15, and 0.08 MEV, respectively. For this average energy, the total noble gas activity released would be 239.7 Ci.

E D~ a = [0.25 x 0.08+ 0.23 x 0.15] x [8.55 x 10'] x [239.7 ]

= 1.2 x 10'em

Low population zone An individual in this low population zone is subjected to a different diffusion coefficient than that of the site boundary. All other parameters are similar to those calculated above .

1. Thyroid Dose D~ = [1044.16+ 32.90] x [7.97 x 10+] x [1.08 x 10 ] x [3.47 x10 ]

= 3.22 rem

2. External whole body dose D~ a = [0.25 x 0.08 + 0.23 x 0.15] x [7.97 x 10 ] x [239.7 ]

= 1.1 x 10 rem 4.1.2 Concurrent Iodine Spike with the Outside Containment Main Steam Line Break Accident In this scenario the reactor trip resulting from primary system depressurization associated with the MSLB creates an iodine spike in the primary system. SRP Section 15.1.5 assumes that the iodine release from the fuel rods to the PCS would increase to a value 500 times greater than the.

release rate corresponding to the iodine concentration at the equilibrium value stated in the plant'.

technical specifications. The equilibrium iodine (I-131) concentration prior to the spike is assumed to be the technical specification"'imit of 1 pCi/g. Using the data provided in Table 1, the iodine production rate producing this equilibrium concentration is calculated based on the following equations:

R =A,x A A =C,xM XP+ XR Ap = 0.693/T,~

~R = F x (1.0-1/DF) /V where:

R Iodine production rate, (Ci/sec)

I-131 decay and removal rate, (per sec).

A equilibrium activity of the I-131 in the PCS, (Ci)

Tl/2 8.02 days, Half life of I-131, (6.93 x 10', sec)

F letdown flow rate, (40 gpm, or 0.667 system coolant volume, (10, 400 ft', or 7.78 x gals/sec)'rimary V 104 gallons)

DF I-131 decontamination factor in the CVCS, (1000)

M Primary system coolant mass (2.13 x 10~ grams)

Xp 1.0 x 10~ (per sec)

XR 8.56 x 10~ (per sec)

C, 1.0x 10 (Ci/g)

The equilibrium I-131 production rate prior to the accident is 2.04 x 10'i/sec.

Using the SRP"'ssumption, the spike would produce an I-131 production rate of 0.102 Ci/sec, (500 x 0.00204). To calculate the time dependent I-131 concentration in the PCS after the accident, the following conservative assumptions were made:

ao The released iodine instantaneously mix with the PCS water,

b. No credit will be taken for the iodine dilution due to ECCS injection.

C. No credit willbe taken for iodine decay and removal.

Losses in PCS mass during the accident is negligible.

Based on the above assumptions, the iodine concentration is:

C(t) =C,+Kt Where:

K = [0.102 /M] x 10' 4.79 x 10~ pCi/sec t time after the accidents, (sec)

At the end of accident, 204 minutes after the break, the iodine concentration would be 6.86 pCi/g. Iodine release using this concentration would result in a much smaller (a factor of 285 smaller) consequences that those analyzed earlier in Section 4.1.1. Therefore the calculated doses, (thyroid and external whole body) at the site boundary and the LPZ would be much smaller than those specified in the SRP, Section 15.1.5.

4.1.3 Evaluation of Effects of Fuel Failure after the NISLB Accident In this scenario, it was conservatively assumed that the transient following the steam line break would result in 2% failed fuel,"" even though the engineering analysis performed in support of the St. Lucie 1 FSAR,"'ndicated that no fuel failure is expected in an outside containment MSLB accident. The inside containment MSLB accident analysis resulted in a 1.6% fuel failure.

Using the ANF calculation"" of core and fuel gap nuclide inventories, and assuming that 2 percent of the fuel would fail following an MSLB accident, an I-131 and Xe-133 dose equivalent concentration (DEC) in the primary coolant system was calculated. For this calculation, it was assumed that the noble gas and iodine nuclides in the gap of the failed fuel would instantaneously release from the fuel and mix with the primary coolant system. Based on this assumption, the PCS I-131 (DEC) is calculated to be 1.089 x 102 Ci/g. Assuming that this concentration would remain constant throughout the accident duration, a total of 602 Ci of I-131 (dose equivalent quantities) would be released to the environment. The thyroid dose to an individual at the site boundary and low population zone would be:

a. Site boundary (exclusjon area boundary) Dose D~ = [602] x [8.55 x 10'] x [1.08 x 10 ] x [3.47 x 10 ]

~ ~

= 19.3 rem

b. Low population zone Dose D~ = [602] x [7.97 x 10~] x [1.08 x 10~] x [3.47 x 10~]

= 1.8 rem.

Using the same assumptions as indicated above, an Xe-133 (DEC) value of 1.83 x 10'i/g was calculated. Assuming a constant concentration over the 204 minutes duration of the accident, an equivalent of 1009.6 Ci of Xe-133 would be released through the affected steam generator.

a. 'ite boundary (exclusion area boundary)

External whole body dose D~ a = [0.25 x 0.046+ 0.23 x 0.135] x [8.55 x 10'] x [1009.6]

= 3.67 x 10'em

b. Low population zone External whole body dose D~a = [0.25 x 0.046+ 0.23 x 0.135] x [7.97 x 10 ] x [1009.6 ]

= 3.42 x 10~ rem 4.2 Control Room Dose Calculations Radionuclides released from the broken steam line may be transported through the air to the auxiliary building where they may enter the control room ventilation system. Analysis of this potential scenario involves specification of the steam generator release rate, degree of atmospheric dispersion, operational conditions for the control room ventilation system, and estimation of radionuclide concentrations and doses in the control room. This section presents a discussion of methods of analysis, specification of source term, and estimation of control room doses for the hypothetical MSLB event.

4.2.1 Methods of Control Room Analysis Section 15.1.5 of the SRP, Regulatory Guide 1.4, and the TMI Action Plan document"" discuss elements of analyses appropriate for the estimation of control room doses under accident conditions. This guidance has been incorporated into versions of the control room habitability evaluation computer code"'i which was used in this analysis. The central feature of the code is solution of lumped parameter,.transient radionuclide activity balances formulated around the control room. The time variation of atmospheric dispersion conditions, radionuclide ingress rates, and control room ventilation system function are represented through time periods defined by the analyst. Model parameters are constant during a time interval but are varied from one time interval to the next. The code considers filtered and unfiltered inflows and removal of contaminants in control room ventilation system recirculation filter trains. In any time interval, the concentration of a radionuclide in the control room represented in Figure 1 is calculated as:

C = C, EXP[-a(t-t,)]+([R;J(a V)] { 1-EXP[-a(t-ti,)] } )

where:

R Aj + LpFpCpV Aj~ C)F) + C~F~ + Ci(1 Ei)F3 + C4(1 E4)(1 E~)F4 F, = Fraction of parent decaying to daughter (F, + E5F5+ L~V)/V C CR concentration at the beginning of a time interval CP CR concentration of the nuclide's parent time at the end of a time interval tb time at the beginning of a time interval C, concentration at unfiltered inleakage point number 1 F, unfiltered air inleakage rate at point 1 concentration at unfiltered inleakage point number 2 F~ unfiltered air inleakage rate at point 2 Ci concentration at filtered inflow point 3 Eq filtration efficiency at filtered inflow point 3 F~ air inflow rate at filtered inflow point C4 at filtered inflow point 4 (CR recirculation train) 3'oncentration E4 'iltration efficiency at filtered inflow point 4 F4 air inflow rate at filtered inflow point 4 E5 CR recirculation loop filter removal efficiency F~ CR recirculation/filtration loop flow rate F, CR effluent flow rate L, radionuclide decay constant L radionuclide parent decay constant and control room volume Control room doses for a time interval are calculated as the time integral of the above concentration multiplied by the appropriate dose factor, breathing rate, and occupancy factor.

Scenario conditions specified in Table 1 are such that control room ventilation system behavior does not change over the time for the postulated accident. As described in Section 4.1, the radionuclide source time does change over the time frame of the release event.

0 Figure 1 Control Room Habitability System Model Outside Alr Intake 2 (F4)

Outside Alr Intako 1 (Fs) Filter (Ni)

Filter (N

Recirculation Loop (Fo)

Control Room Unllltered lnleakage 1 (Fi)

Volume Unfiltered Inleakage 2 (v) (Fs)

Filter

(%

Flow Out (FQ 4.2.2

~ ~ Control Room Source Term Characterization The estimation of radionuclide source terms for the pre-existing iodine spike and 2% failed fuel cases is described in Section 4.1. For the purposes of control room dose analysis, the pre-existing iodine spike case iodine source term expressed as DEC I-131 is converted into nuclide specific release quantities using the definition of dose equivalent. For iodines, the dose equivalency is expressed as:

A~. DCF,. = Z AJ DCFJ where:

Ad dose equivalent activity of nuclide i DCF; thyroid dose conversion factor for nuclide I AJ activity of nuclide j DCF) thyroid dose conversion factor for nuclide j and the summation is taken over all'nuclides. Using the iodine species relative distribution specified in FSAR Table 12.1-3, the above relation may be solved for individual species activities using the dose equivalent activity estimate derived in Section 4.1. The results of this calculation are summarized in Table 2 for iodine species. The source term for the first 70 seconds is due primarily to the pre-existing iodine inventory in the steam generator while the release for the remaining time interval is due to continuing leakage from the PCS. A similar relation may be proposed for the noble gas species with the substitution of average energy for dose conversion factor in the dose equivalency relation. As stated in Section 1.4, the total activity of noble gas species released during 204 minutes for the pre-existing iodine spike case is 239.7 Ci. Using the relative concentrations of noble gas species reported in FSAR Table 12.1-3, the individual noble gas releases presented in Table 3 were calculated for the pre-existing iodine spike case. For the 2% failed fuel case, the distributions of iodine and noble gas nuclides are known directly as the distributions present in the gap inventory and conversion using the DEC concept is not required. Release estimates for the 2% failed fuel case for iodine and noble gas nuclides are presented in Tables 2 and 3, respectively.

4.2.3 Results of the Control Room Dose Analysis Control room doses were estimated for the pre-existing iodine spike and 2% failed fuel cases using the models and source terms described in Section 4.2.1 and 4.2,2 and the control room ventilation system parameters presented in Table 1. These conditions are those defined in other analyses discussed in the St. Lucie Unit 1 FSAR. For these conditions, thyroid, skin, and whole body doses over the entire time frame of the event are summarized in Table 4. More detailed output of the computer code, including ventilation system conditions and doses at significant time steps are presented in Table 5 for the pre-existing iodine spike case. The results indicate 0

~

~

that, for the pre-existing iodige spike case, the major portion of the dose occurs after release of the initial steam generator inventory and during a time period in which the portion of this initial pulse of activity which entered the control room is removed by the recirculation filters. The portion of the activity which is due to continued leakage from the PCS subsequent to release of the initial steam generator inventory is a small contributor to the total dose. Unfiltered inleakage accounts for approximately 80% of the dose. The calculated control room dose for the bounding MSLB source term is 9.9 rem thyroid, 1.3E-4 rem whole body, and 9.1E-3 rem skin for a 1 gpm SG tube leak rate. Comparison of the doses for the pre-existing iodine spike and the 2% failed fuel cases indicates that thyroid dose is dominated by the iodines, but that both iodines and noble gases contribute to skin and whole body doses. The noble gas contribution to skin and whole body dose for the 2% failed fuel case is approximately 3 times the contribution of noble gases to the some tissues for the pre-existing iodine spike case. All of these calculated doses are less than their regulatory limits."'"'

~ I I~

0 ~

Table 2 Control Room Analysis Iodine Species Release Quantities Releases (Ci)

Pre-Existing Iodine Spike 2% Failed Fuel Nuclide Time Period (s) 0 to 70 0 to 12,240 0 to 12,240 I-131 836.4 862.7 528.0 I-132 215.0 221.8 111.0 I-133 1,192.7 1,230.3 402.0 I-134 132.5 136.7 103.0 I-135 568.4 586.3 215.0 Table 3 Control Room Analysis Noble Gas Release Quantities Releases (Ci)

Pre-Existing Iodine Spike 2% Failed Fuel Nuclide Time Period (s) 0 to 70 0 to 12,240 0 to 12,240 Kr-85m 0.011 1.92 8.80 Kr-85 0.007 1.20 8.70 Kr-87 0.005 0.96 8.39 Kr-88 0.018 3.12 19.70 Xe-131m 0.011 1.92 2.40 Xe-133 1.265 221.24 379.0 Xe-135 0.052 9.11 6.26 Xe-138 0.003 0.50 14.20 Table 4 Control Room Analysis for 1 gpm Steam Generator Tube Leak Rate Dose (rem)

Tissue Pre-Existing Iodine Spike 2% Failed Fuel Thyroid 9.9 5.5 Skin 9.1E-3 9.2E-3 Whole Body 1.3E-4 1.1E-4

Table 5 Control Room Dose Analysis Results Pre-Existing Iodine Spike Case TIME STEP START: 0.000000E+00 hr TME STEP END: 2.000000E-02 hr Buildin cross sectional area m"2: 0.000000E+00 Buildin hei ht m: 0.000000E+00 Releasehei ht m: 0.000000E+00 Effluent vertical velocit m/s: 0.000000E+00 Effluent flow rate m"3/s: 0.000000E+00 Horizontal distance to rece tor m: 0.000000E+00 Air intake hei ht m: 0.000000E+00 Winds eed m/s: 0.000000E+00 Vertical dis ersion class:

Horizontal dis ersion class:

X/ s/m"3: 4.860000E-04 Flow in from unfiltered source 1 m"3/s): 0.000000E+00 Flow in from unfiltered source 2 m"3/s): 0.000000E+00 Filtered intake flow source 1 m"3/s: 4.700000E-02 Filter efficienc 41:

elemental fraction or anic fraction 0000 articulate fraction 0000 Recirculation flow rate m"3/s: 9.440000E-01 Recirculation filter efficienc:

elemental fraction .9500 or anic fraction 9500 articulate fraction 9900 Filtered intake flow 2 feeds recirc: 2.120000E-01 Intake 2 filter efficienc:

elemental fraction or anic fraction articulate fraction Bottled air flow rate m"3/s: 0.000000 E+00 Control room volume m"3: 1775.200000 CUMULATIVEDOSE END WH BODY SKIN THYROID LUNG BONE LIVER TIME REM REM REM REM REM REM HOURS 2.000E-02 2.535E-06 1.217E-04 2.221E-01 1 ~ 137E-08 7,989E-04 1.262E-03 Table 5 Control Room Dose Analysis Results Pre-Existing Iodine Spike Case (Continued)

TIME STEP START: 2.000000E-02 hr TIME STEP END: 3.400000E+00 hr Buildin cross sectional area m"2: 0.000000E+00 Buildin hei ht m: 0.000000E+00 Releasehei ht m: 0.000000E+00 Effluent vertical velocit m/s: 0.000000E+00 Effluent flow rate m"3/s: 0.000000E+00 Horizontal distance to rece tor m: 0.000000E+00 Airintakehei ht m: 0.000000E+00 Winds eed m/s: 0.000000E+00 Vertical dis ersion class:

Horizontal dis ersion class:

X/ s/m"3: 4.860000E-04 Flow in from unfiltered source 1 m"3/s: 0.000000E+00 Flow in from unfiltered source 2 m"3/s: 0.000000E+00 Filtered intake flow source 1 m"3/s: 4.700000E-02 Filter efficienc 01:

elemental fraction or anic fraction articulate fraction Recirculation flow rate m"3/s: 9A40000E-01 Recirculation filter efficienc:

elemental fraction .9500 or anic fraction .9500 articulate fraction ~ 9900 Filtered intake flow 2 feeds recirc: 2.120000E-01 Intake 2 filter efficienc:

elemental fraction or anic fraction articulate fraction .0000 Bottled air flow rate m"3/s: 0.000000E+00 Control room volume m"3 1775.200000 CUMULATIVEDOSE END WH BODY SKIN THYROID LUNG BONE LIVER TIME REM REM REM REM REM REM HOURS 3.400E+00 1.273E-04 8.280E-03 9.830E+00 1.297E-04 3.525E-02 5.559E-02 Table 5 Control Room Dose Analysis Results Pre-Existing Iodine Spike Case (Continued)

TIME STEP START: 696.000000E+00 hr TIME STEP END: 720.000000E+00 hr Buildin cross sectional area m"2: 0.000000E+00 Buildin hei ht m: 0.000000E+00 Releasehei ht m: 0.000000E+00 Effluent vertical velocit m/s: 0.000000E+00 Effluent flow rate m"3/s: 0.000000E+00 Horizontal distance to rece tor m: 0.000000E+00 Airintakehei ht m: 0.000000E+00 Winds eed m/s: 0.000000E+00 Vertical dis ersion class:

Horizontal dis ersion class:

X/ s/m"3: 6.360000E-05 Flow in from unfiltered source 1 m"3/s: 0.000000E+00 Flow in from unfiltered source 2 m"3/s: 0.000000E+00 Filtered intake flow source 1 m"3/s: 4.700000 E-02 Filter efficienc ¹1:

elemental fraction or anic fraction .0000 articulate fraction .0000 Recirculation flow rate m"3/s 9 440000E-01 Recirc filter efficienc:

elemental fraction .9500 or anic fraction 9500 articulate fraction 9900 Filtered intake flow 2 feeds recirc: 2.120000E-01 Intake 2 filter efficienc:

elemental fraction or anic fraction articulate fraction Bottled air flow rate m"3/s: 0.000000E+00 Control room volume m"3: 1775.200000 CUMULATIVEDOSE END WH BODY SKIN 'HYROID LUNG BONE LIVER TIME REM REM REM REM REM REM HOURS 7.200E+02 1.337E-04 9.140E-03 .872E+00 1.708E-04 3.540E-02 5.584E-02 5.0

SUMMARY

AND CONCLUSIONS An analysis of the radiological consequences of a postulated MSLB outside containment accident at the St. Lucie Unit 1 nuclear power plant was performed in accordance with SRP 15.1.15.

Appendix A (4). This analysis was performed in the following three steps.

First, using the conservative assumptions in the SRP, three scenarios were used to calculate the radioisotope source term released to the environment from this postulated accident. These scenarios are: (1) pre-existing iodine spike, (2) accident induced iodine spike, and (3) accident induced fuel failure. The bounding iodine source term was calculated with the pre-existing iodine spike scenario, while the bounding noble gas source term was calculated with the accident induced fuel failure scenario.

The second step involved the calculation of the site boundary (exclusion area boundary) and low population zone thyroid and whole body doses. These doses were calculated using the bounding source terms from the first step and appropriate conservative parameters for atmospheric dispersion, breathing rate and dose conversion factor.

The third step calculated the thyroid, whole body, and skin doses inside the control room from the bounding source terms which were determined in the first step. The control room doses were calculated using conservative assumptions regarding the control room HVAC system operation the performance of radionuclide absorbing filters in the HVAC system, and conservative

'nd atmospheric dispersion parameters.

The bounding results of this analysis are tabulated below and show the margin to the regulatory limits set forth in 10 CFR 100"" and 10 CFR 50 GDC 19."" All doses are normalized to 1 gpm steam generator tube leak rate, but they can be linearly extrapolated to high leak rates.

MSLB Outside Containment Doses (rem) (for 1 gpm SG Tube Leak Rate)

Thyroid Whole Body Skin Maximum Limit Maximum Limit Maximum Limit Calculated Calculated Calculated EAB (Site Boundary) 34.51 300 0.00367 25 N/A N/A 3.22 300 0.000342 25 N/A N/A Control Room 9.9 30 0.00013 5 0.0092 30 A second comparison of calculated doses to the regulatory limits presented in SRP 15.1.5, Appendix A, Section II Acceptance Criteria is delineated below.

MSLB Source Term Scenario Calculated Dose SRP Criteria Pre-Accident Iodine Spike 34.51 Rem Thyroid 300 Rem 0.0012 Rem Whole Body 25 Rem Accident Initiated Iodine . 0.12 Rem Thyroid 30 Rem Spike 0.0012 Rem Whole Body 2.5 Rem As discussed in Section 4, the accident initiated iodine spike results in an iodine source term which is a factor of 285 lower than the pre-accident iodine scenario.

A licensing analysis of the radiological consequen'ces of a main steam line break outside containment (MSLBOC) at the St. Lucie Unit 1 nuclear power plant has been performed. All input data and assumptions are based on appropriate conservative and consistent information from USNRC regulations (i.e. standard review plan and code of federal regulations) and St.

Lucie 1 plant sources (i.e. FSAR and technical specifications). The assumptions and data have been confirmed"" by Florida Power and Light Company, the licensee which is operating St.

Lucie 1. For all postulated radiological source term scenarios for the MSLBOC accident, the calculated EAB(or site boundary), LPZ, and control room thyroid and whole body doses as well as control room skin doses are considerably less than the relevant regulatory limit.

6.0 REFERENCES

1. St. Lucie Unit 1 Updated Final Safety Analysis Report (UFSAR), Amendment 15.

2. St. Lucie Unit 2 Updated Final Safety Analysis Report (UFSAR).

3. St. Lucie Unit 1 Technical Specifications.

4. NUREG-0800, USNRC Standard Review Plan 15.1.5 Appendix A, "Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR", Revision 2, July, 1981.

5. USNRC Regulatory Guide (R.G.) 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors", Revision 2, June, 1974.

6. USNRC R.G. 1.52, "Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light Water Cooled Nuclear power Plants", Revision 2, March, 1978.

7. ICRP 30, "Limits for Intake by Workers", International Commission on Radiological Protection.

8. ICRP 38, "Radionuclide Transformations - Energy and Intensity of Emissions",

International Commission on Radiological Protection, Pergamon Press.

9., Memorandum NF-97-065 from J.N. Kabadi (FP&L) to'G.L.,Boyers(FP&L), "Initial Steam Generator Mass for St. Lucie Unit 1 MSLB Dose Calculations", February 18, 1997.

10. ABB-CE Calculation A-SL2-FE-0072, Rev. 00 (page 53 of 187).

11. Florida Power and Light Letter ENG-SPSL-97-0068, to Steven Mirsky (SAIC) from Carl Bible (FP&L), "St. Lucie Unit 1 Transmittal of Review & Verification of Values & Input Parameters - File: Engineering Evaluation JPN-PSL-SESS-96-076 Rev. 0", dated February 26, 1997.

12. 10 CFR Part 50, Appendix A, GDC 19, "Control Room".

13. 10 CFR,Part 100.11, "Determining of Exclusion Area, Low Population Zone and Population Center Distance.

~ ~

0

14. NUREG/CR-5659, Centrol Room Habitability System Review Models, H.Gilpin, Science Applications International Corporation, December 1990.

15. NUREG-0737, "Clarification of TMI Action Plan Requirements," Item III.D.3.4, "Control Room Habitability," November 1980.

16. NUREG-0800, USNRC Standard Review Plan 6.4, Control Room Habitability System, July, 1981.

17. Facsimile from J. Kabadi and Chris Buehrig of Florida Power and Light to Steve Mirsky at SAIC, "ANF-88-113(P), St. Lucie Unit 1 Assessment of Radiological and Rod Bow Effects for Increased Burnup", July 1988, Advanced Nuclear Fuels Corp., March 11, 1997.

Appendix A Memorandum NF-97-065 from J.N. Kabadi (FPRL) to G.L. Boyers(FP &L), "Initial Steam Generator Mass for St. Lucie Unit 1 MSLB Dose Calculations", February 18, 1997

{REFERENCE 9]

e

~e Ento'Of fice CorresyozuKence NF-97-065 To: G. L. Boyers Date: February 18, 1997 e

From: J. N. Kabadl Department: ENG/NF/JB

Subject:

Initial Steam Generator Mass for St Lucle Unit 1 MSLB Dose Calculations

Reference:

JPN Calculation PSL-1FJF-S5-155, Revision 1 This memo provides the initial mass in the St. Lucie Unit 1 steam generators. The full power (HFP) values provided are verified to be those from the reference calculation which documents the St. Lucie Unit 1 Cycle 14 Groundrules.

Total mass per steam generator at HFP = 137,970 Ibm Liquid mass per steam generator at HFP ~ 127,602 Ibm Prepared By: Verified By:

~~+

Mis4~ 8/~r Distribution:

'C. J. Buehrig K, R. Craig C. G, O'Farrlll C. Viliyrd

[~ '

~

Appendix B Florida Power and Light Letter ENG-SPSL-97-0068, to Steven Mirsky (SAIC) from Carl ..

Bible (FP&L), "St. Lucie Unit 1 Transmittal of Review & Verification of Values & Input--

Parameters - File: Engineering Evaluation JPN-PSL-SESS-96-076 Rev. 0", dated February 26, 1997

[REFERENCE 11]

ENG-SPSL-97-0068 FPL FEB 26 1997 Scientific Applications International Corporation 20201 Century Boulevard Germantown, Maryland 20874 Attention: Mr. Steven M. Mirsky Manager, Nuclear Facilities Safety St. Lucle Unit 1 Transmittal of Review & Verification of Values & input Parameters File:

Reference:

1 ~ NRC Letter dated January 23, 1997 from L.A. Wiens to T.F. Plunkett, "St.

Lucie Unit 1 Steam Generator Run Time Analysis".

2. FPL Purchase Order 00019096, Blanket Release 002 to APTECH Engineering Services.

3. SAIC letter dated February 24, 1997, Steven M. Mirsky to Chris Buehrig.

Gentlemen:

This letter formally transmits to you our review and verification of values and input parameters identified by SAIC in Reference 3. These values and input parameters are to be used by SAIC to recalculate the dose assessment for MSLB outside containment in accordance with the guidance in SRP 15.1.5. This work was requested by NRC Staff in Reference 1 and authorized by FPL in Reference 2.

If you have any further questions or need additional information contact Chris Buehrig at (561) 467-7507 or Gary Boyers at (561) 694-4909.

Sincerely, Carl R. Bible Engineering Manager CRB/GLB A at'B Quuia&q;

Attachment:

SAIC Letter, S.M. Mirsky to Chris K.R. Craig Buehrig, dated 2/24/97 (4 Pages)

C. Buehrig G.L. Boyers J.. Begley (APTECH)

W. Hannaman (SAIC) an FPL Group company

EN G-SPSL-97-0068 C.R. Bible to SAICPage 2 of 5 ITEM PARAMETER SOURCE REVIEWERS COMMENTS Site Boundary X/Q FSAR Sec 2.3.4.2 Values Correct, References Correct Change Reference to: FSAR Sec.

8.55 E-5 sec/m'AIC sec/m'ow Population Zone X/Q FSAR Tbl 15.4.1-4 2.3.4.3 for both items 7.97 E-6 Breathing Rates Reference Correct Oto8hr: Changes: 8 to 24 hr, 3.47 E-4 m'/sec 1.75 E-4 m'/sec, 24 to 720 hr, 2.32 Reg Guide 1.4 EA m'/sec 8to24hr: Note: Accident does not use 2.32 EA m'/sec values past 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 24 to 720 hr:

1.75 E-4 m'/sec I-131 Thyroid Dose Factor Value and reference is correct.

1.08 E+6 rem/Ci Inhaled ICRP 30 Change: ADD:

'Inhaled'rimary to Secondary Leak FSAR Section Technical Spec... Change Reference to Just:

1 gpm Technical Specification 3.4.6.2.c Max. time of Opm at or Technical Specifications Value Correct above 1 uCI/gr if I- 3.4.8 Reference Correct 131(DEC): 106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br />... Change: "Hot Shutdown to Hot Standby" (Ref. TS T1.2, pl-9)

I-131 (DEC) conc. in Technical Specifications Value Correct

...(PCS) 3.4.8 & Reference Correct 1 uCVgm... fig. 3 4-1 Change Unit of measure to uCVgm 60 uCVgm. . ~ & add the "(DEC)"

I-131 (DEC)conc in FSAR tbl 15.2.11-5 secondary: Value Correct 0.1 uCVgr Reference Correct Change Unit of measure to uCI/gm 8 add the "(DEC)"

PCS Gross Activity: Technical Specification Value Correct 100/ Ebar 3.4.8 Reference Correct Q1 z.4 Reviewed By: Date:~Z. Verified By:

'2 zC 97

~ l

~~

ENG-SPSL-97-0068 C.R. Bible to SAICPage 3 of 5 ITEM PARAMETER SAIC SOURCE REVIEWERS COMMENTS 14 Noble Gas Release ICRP 38 & Value Correct FSAR tbl 11.1-1 References correct CHANGE DISPLAY OF VALUE &

SOURCE (example for Value) (example for Source) 239.7 Xe-131 (DEQ) Ci ICRP 38, 'Radio....'or the energies and yields.

No accumulation of noble gas in the SG prior to break. St. Lucie FSAR Table 11.1-1 for the noble gas concentrations Noble gases...(existing words)...55.14/E Ci.

Average E for noble gases is 0.23 (Ebar p of 0.23+ Ebar T of 0.084).

The Ci released would be 55.14/0.23 = 239.7 Ci (delete: 'This Ci will ....(DEQ) release) 18 CVCS PCS radwaste FSAR Sec. 9.3.4 Values Correct removal and Table 11.2-4 References Correct Letdown flow = 40 gpm ADD(between DF and =) "for--

CVCS DF for Iodlnes "

iodines = 1000 Control Room HVAC damper close times: Values Correct a.SIAS 66.1 sec St Lucie 2 FASR References Correct Table 15.1.5.1-1 b.35 sec to close FSAR 9.4 20 Control Room HVAC design FSAR Values Correct features References Correct a.M/U 750 cfm Pg 12.2-9, Sec 9.4.1.3 b.recirc 2000 cfm Pg 12.2-8

c. X/Qs Pg 15.4.1-17 d.Char Filt Eff Sec 15.4.1 e.Occup Factors Sec 15.4.1Sec 9.4 f.Post-Acc inflow Sec 9.4 Sec 6.4.1.3.1 21 Control Room Volume FSAR Pg 12.2-8 Value Correct Reference Correct 22 Control Room FSAR Sec. 15.4.1 Value Correct Unfiltered Inieaka e Reference Correct 23 Iodine Chemical FSAR Sec 15.4.1 Value Correct Com osition Reference Correct 24 Control Room Reg Guide 1.52 Value Correct HEPA Eff. Reference Correct I

~91 Reviewed By: Date: Verified By:

ENG-SPSL-97-0068 C.R. Bible to SAICPage 4 of 5 ITEM PARAMETER SAIC SOURCE REVIEWERS COMMENTS 10 Activity Release for (100% of initial iodine in the Conservative Assumption Steam Line Break secondary side) + (iodine transferred Outside Containment from the primary system after iodine spike equilibrium in the primary system) will provide conservative results.

11a MSIV Closure Signal Time The value of 63.9 seconds is based St. Lucie Unit 2 FSAR Table 15.1.5.1-on the St. Lucia Unit 2 FSAR Table 1 15.1.5.1-1 for steam line break outside containment. MSIV closure affects the unaffected steam generator releases. In the dose calculations where all the secondary mass will be assumed to be released from the steam generators, this timing will not affect the results. This value of 63.9 seconds is, therefore,,

acceptable for this purpose.

11b MSIV Closure Delay Time 6.9 seconds St. Lucie, Unit 1 FSAR Table 15.4.6-.2 P.SL-1FJF-95-155, Rev. 1 12 Cooldown Duration St. Lucia Unit 2 analysis showed that St. Lucie Unit 2 FSAR Table After Break (Shutdown shutdown cooling is initiated at 15..1.5.1-1 Cooling Initiated) 12,240 seconds after the break. It has been stated in the ABB-CE ABB-CE Gale A-SL2-FE-0072, Rev.

referenced calculation that the 00 (page 53 of 187) cooldown rate is not used in the dose calculation as it is assumed that all the SG activity is released to the atmosphere. Under similar conditions the time of 12,240 seconds for initiating shutdown cooling is acceptable for St. Lucle Unit 1.

13 Steam Generator Hot 127,602 Ibm PSL-1FJF-95-155, Rev. 1 Full Power Secondary Side Water Inventory 15 Iodine and Noble Iodine: Use FSAR Table values + 5% St. Lucie Unit 1 FSAR,Table 12.1-3 for 1% Failed Gas'eleases Fuel Noble gas: Use FSAR Table values Reviewed By: Oate:~2M Verified By: 4',91 3~y k'a,b~cfi

ENG-SPSL-97-0068 C.R. Bible to SAICPage 5 of 5 ITEM PARAMETER SAIC SOURCE REVIEWERS COMMENTS t16 Fractlonzf,GoreFuel ~ '.-"I 2'0 St LucieUnit.1,'FSAR,Sectlorj15 4,'6 3'he.S+t.,L'ucfe'Unit,1.'Inside, the g7F48-'113{P),;July 1988 containment steam',line break anaIysis has;;1'.61,k'failed.fuel,.forcthe Failure'CS moist,case~it'lias been.,stated,ln ANF<8;1,13(P)7page',13,",that;no fuel, failure is',expected, for outsIde Ke'containment steam lije',break j Also per; FSAR.'Section;,15;-'4;6,

)nside,the containment steam, line

'aiiuie than. outsIde the:contaInment steam liiie.break e'vent;: A"value'o f 2%,fuel failure. Is th'erefoie conservative',for use In thedose calculations',fo', outside, the containment steam linebaal 17 Liquid Volume 10,400 ft St. Lucie Unit 1 FSAR Table 5.1-1 This value which appears in the St. Lucle Unit 1 Technical FSAR Table 5.1-1 is also consistent Specification 5.4.2 with the Tech Spec value of -11,100 ft'otal RCS volume, which includes the pressurizer gas volume of 700 ft.

Shaded';item'sirepresentchained;v'glues.

Reviewed By Date:

2. aG ~ ~

Verified By: C-

FEB 24'97 15:34 No.035 P.02 ID:

Rh"~cl~~e y +o F46-.$ / S/ 7'7 - Oc bg Science Apptleetlone Internatlonel Corporation An Employeo Owned Company February24, l997 Chris Buehrig Florida Power and I.,ight Company 650l South Ocean Diivc Jensen Beach, Florida 34957

Dear Mr. Buehrig:

In accordance with our milestone schedule for the MSLB outside containment SRP analysis task for St. Lucie I, I am providing the following list of input data values and associated d t a e assumed for tlus analysis. To meet our schedule, please review these parameters and their assumed values and confirm in writing that they arc appropriate for this licensing consct~ative analysis no later than February 26, 1997.

List and.values of input parameters Parameters Vn)ues Source 0-2 hour Z/Q (site 8.55 x 10" St. Lucie I FSAR boundary) sec/in'.97 Section 2.3.4.2 0-8 hours gQ (low x St. Lucic l FSAR population xone) 10'ec/m'.47 Section 15.4, '1'able 15.4.1-4 Breathing rate 5t 10'"/sec for 0 t8 hrs.; 2.32E-4 for 8-24 USNRC Reg Guide 1.4 hrs., and 1.75EA for 24 to 720 hrs. (c.2.c)

)-131 Thyroid dose 1.08 x 10'ein/Ci ICRP publication 30.

conversion factor Primary to secondary I gpm (or 2,702.8 gr/min (primaty system condition; St. Lueic l FSAR leak rate and HFP specific density of 0.724 based on 2? SO psia and 575 Section Tcclinicel primary coolant F)] Specilication Leak density Limit (Scc. 3.4.6.2)

Maximum time of 106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br />, l 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of operation above l pCi/gr, St. Lucia 1 Tech Spec.

(o operation at abnve 1 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />'to bc in Hot,Shut down.) Andrew, 3.4.&

pCi/gr of I-131 (DEC) l-)31 concentration s 1 pCi/gr Tech Spec. Limit, with a maximum of St. Lucie l Tech. Spec in Primary Caolant 60 pCi/g in mode 1 with a ~ 80% power level Section 3.4.8 and System (PCS) Figurc 3.4-1 20Z05 Century 8oulevant. Gennantown, Afatyland 208T4 ~ (305) 3534550 Oow sllc sllC cÃcec 05anc Abuyuenyue, Ao cab'& sprinpl, AI~fa& chunA l4nlsvst, Lao vertu. Lot Ahba. l ce Iktoui oak 0+t. cHeae. san A~ soatua Tactor

FFB 24'97 15:34 Ho.055 P.05 ID:

P+ool ~e~ P fo g~@ -sPs L-97-Dab/

Parameters Values Source J-131 concentration O. I pCi/gr St. Lucic I FSAR table in Secondary C:oolant I S.2,J I-S System (SCc))

PCS s peciTic activity ~ 100/Ii (E is thc sum of'thc avcragc p and y cocrlrics St. I,ucic I Tech. Spec.

per disintegration ) MEV] for isotopes other than 3.4.lt iodide's.)

Steam linc break main assumplinn is that 100% of lodii)c in thc SCS outside contaiomcnt: (initial amount plus that transferred aAcr iodine spike equilibrium in thc PCS) would bc released to thc aunnsphcl'c.

l. MSIV closure time a) MSIS 63.9 seconds aAcr tho brcak, with a I.OOP St. Lucic 2 FSAR a)) d tl oil lite aller turbine trip Table IS.J.S. I-I b) MS)V closure 6 seconds afb:r MSIS St. Lucie I FSAR Table I5.4.6-2 L. Cooldown 12,240 seconds, (using SL Lucic 2 time duration tn St. I,ucic 2 I'SAI<

duration after brcak shutdown cooling, froio condition a above.) Ta'l)lc 15.J.5.1-1 Steam generator hot 127,602 Ib Kabadi to Boycrs full power secondary memo dated 2-18-1997 side water inventory (FP&I.)

Noble gee releere oo accumulation of noble gases. in thc S(i prior to Avcragc P. values werc brcak. Noble gases arc directly rclcascd to calculalcd based oi) atmosphcrc aJler lhc brcak for 204 m)ou'les. data provided in

-0 2702.8 )o IOO/E x 10>> x 204 55. J4/I) Ci tion ICRI'ublic')

avcragc E for noble gases pcr PCS concentration 38,"Radionuclidc Table 11.1-1 is 0.23 (0. I 5)Q+ 0.084 ) I:.'Y])

ol'SAR Traitsformatioos For this avcragc cocrgy, the Ci rclcascd would bc Iiocfgy aod Jotcllsity 239.7 (5$ .14/0.23). 'I Itis Ci will bc usual to reprcscn[ lanissions", St. Lucic I, thc Xc- 131 (DLQ) release. FSAR 'I'able I I. I -I lodh)e and Noble gas '1-131 (DHC) 7.0 pCi/gr. (Thyroid equivalency) St. Lucio I FSAR release fur 1 "/>> failed Xo-133 (DEC) 353.6 I)CJ/gr. 12, 1-3, and '1'able fuel VSNRC R.(i. I.4 Fraction of core fuel '.61 "/>> SL I.ucic I 1'SAR failure for the MSf 8 Tablo 15.4.6-4

~ ~ ~

)0,400 cubic feel St. l.ucic 1'SAR 1'CS Iitluid water 1 volume 'I'able ~ l-l

FEB 24'97 15:35 No.035 P.04 ID:

g+wsh~e~f 7 O

~s ~

~

5@5 L- 9 7- Gog~g foal Paranietcrs Values Source CYCS PCS Ictdowri IIow rate.= 40 Iym St..Lucie 1 I'SAR

]8 radwaste CVCS DI:= Iono Sectiori!).3.4, and re ill o vill Table 11,2-4 Control room JIVAC a) SIAS on low pressurize prcssure at 66. I scc. St. Lucic 2 I'SAR I damper closul c time 'I'able l5.1.&. 1-1 b) danipcr closure time aAcr signal is 35 scc.(i.c. St. Lucio I FSAR controf 1'ooiil daniper closes nt 101.1 scc. Aller Section 9.4 MSLB)

Control Room IIVAC a) normal uiitrltcrcd makeup flow rate is 750 cfin St. Lucic I FSAR Pat,c dcslgn features 12.2-9 b) rccirculatio>> IIow rate is 2000 cfin St. Lucic ) FSAR I'agc, 12.2-8 c) 0 to 8 hr. Atninsphcric dispcrsinn factor 4.116E-4 St. Lucio l FSAR page sc<<jcrrbic mctcr, 8-24 hr value&.17F 4, 24-96 hr 15.4.1-12 value 1.68&,96-720 hr value 6.361'.-5 d) charcoal filter iodine rcinoval cGicicricy r) $ % for St. Lucio I FSAk elcmc>>tal and organic a>>d 99% for particulate Scctioii 15.4.1 c) occupa>>cy factor '.0 for 0-24 lirs, 0.6 for 24-96 St. Lucic I I'SAR hrs., 0.4 fur 96 to 720 hrs. Section I S.4. l f) controlled post-accident filtcrcd iiiflow450 <<fin St, Lucic I FSAR Section 9.4 Control Room 62.700 cubic feet St. I.ucic I FSAR 1'age z Volume )22-8 Control Room l QO cfin St. Lucie 1 FSAR 2 4- lJnfiltercd lnlcakage Section IS 4 I Iodllle Chenucat 91% clcincntal, 4% organic,-5% particulate St.Lucio I FSAR Section I SA. I Comp orition Control Room HEPA 99% t)SNRC R.G. 1.52 filter efficiency

ID: FEB 24'97 15:55 No.035 P.05 e

(

II ~gncA'~e,~+ ~~

gD4- SPSL - ~7->>48

'l'he above list constitutes the second n>ilestone of this task. On February 27, l 997, l will bc verbally communicating our third milestone which will be the preliminary results of our analysis.

Sincerely, Steven lVJ. Mirsky, P,L.

Manager, Nuclear 1'acilitics Safety cc: Jim Beglcy, APTECH Gary Doyers (FP&L)

Bill Hannanian (SAl(:)

Appendix C Facsimile from J. Kabadi and Chris Buehrig of Florida Power and Light to Steve Mirsky at SAIC, "ANF-88-113(P), St. Lucie Unit 1 Assessment of Radiological and Rod Bow Effects for Increased Burnup", July 1988, Advanced Nuclear Fuels Corp., March 11, 1997

[REFERENCE 17]

r 5 ~ fi  %

a t < hatt ~

gee - -. s ~ 'te "a l q )- g <<g ~'g a a a ~ 4t tti S I H I I, )

M%I

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( I

~ ~

CtM H I ~ It 4 ~I P o 0 l-

,g 4

~

k r / ,

~ +J

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4 e w gOgv e oo e a

TABLE 3.3 CORE AND GAP FISSION PROQUCT ACTIVITIES CcQ)

Oesign Base 14x14 Fuel Assembly ANF 14x14 Fuel Asselbly Design EOL Averaged Core Exposure of 25 GWd/HTU EOL Average Core Exposure of 40 GMd/HTU Fraction of Fraction of Curies in Activity Curies Curfes in Activity Curies Embalm Mm ~p ~re ~a 1-129 2.32E>00 .166 3.84E-D1 3.68E+00 .217 7.99E-01 1-131 7.88Ew07 .104 8.20E106 7'.89E+07 .)29 1.02E+07 I-132 1.12E%08 .014 1.57E+06 1.12BOS .019 2. 14E+06 i-133 1.48EeOS .04 5.92E+06 1.46E+OS .053 7. 76E+06 I-134 1.69B08 .009 1.52E+06 1.66E+08 .012 1.99E+06 1-135 1.3IE408 .024 3.16E+06 1.30E+08 . 032 4.15E+06 Cs-134 2.75M)7 .155 4.26E+06 Cs-134N 6.61E+06 .012 7,93E~04 Cs-135 2.88EIOI .155 4.46E+00 Cs-136 6.43E>06 .098 6.31K+05 Cs-137 1.08E+07 .155 1.67E+06 Cs-138 1.38E+OS .005 6.91E+05 Cs-139 1.36E+08 .003 4.07E<05 Cs-140 1.24E+OS .0009 1.12E+05 Cs-141 8.77B97 .0006 5.26E+04 Cs-142 7.03E+07 .0002 1.41E+04 Cs-143 3.61E+07 .0002 7.21E+03 Te-123M 7.61E+01 .336 2.56E+Ok Te-125M 3 07E+05

~

'.3 9.20E+04 g Te-127 6.63E+06 .071 4.71E+05 Te-127N

~

l

~ g

}.42E+06 ..333 4.74E~05 ~~

Q~Q

TABLE 3.3 CORE AND GAP FISSION PRODUCT ACTIVITIES (CONT,) g~'gi 'm va4. >

'vi.'NF Design Base 14x14 Fuel Assembly 14x14 Fuel Assembly Design EOL Averaged Core Exposure of 25 GWd/NU EOL Average Core Exposure of ig Nd/HTU CR g Fraction of Fraction of Cuties in Activity Curies Curies in Activity Curies Mme ~CX ~m Te-129 3.0BEe07 .027 8.33Ee05 Te-129H 5.26Ee06 .271 1.43Ee06 Te-131 6.92Ee07 .017 1.18E+D6 Te-131K 1.18E+07 .112 1.32Ee06 Te-132 1.08Ee08 .154 1.67EeO/

Te-133 4.12E+07 .012 5.00E+OS Te-133H 1.14E+DB .024 2.74E+06 Te-134 1,47E+08 .022 3.24Ee06 Te-135 1.27Ee08 :002 2.55EeOS I C

Kr-BS 7.33E+05 .102 7.42Ee04 1.12E+06 .149 le68EeOS Kr-85N 1.79E+D1 .00& 1,43Ee05 1.70Ee07 .01 1,70EeOS Kr-87 3.46E+07 . .004 1.38Ee05 3.24E+07 .005 1.62BOS Kr-88 5.04Ee07 .006 3.02E+05 4.75Ee07 .008 3.80E<05 Kr-89 6.30E+07 .0008 '5.04E+04 5.91E+07 .001 5.91E+04 Xe-131M 6.38EeOS .055 3.51E+04 6.43Ee05 .072 4.63E+04 Xe-133 1.48E+08 .038 5.63E+06 1.46E+OB .05 7.32E+06 Xe-133N 3.58E+D6 ..025 8.95E+04 3.56Ee06 .034 1.21K+05 Xe-135 3.13E+07 .011 3.44E+05 3.12E+07 .014 4.36E+05 Xe-135H 3.97Ee07 .002 7.94E+04 3.94E+07 .002 7.87Ee04 1.43E+05 O I Xe-137 1.44E>08 .0009 1.30E+05 1.43E+08 .001 gss Xe-138 1.38f+DB .002 2.76E+05 1.37E+08 .002 2.75E+05 ggs cep Q

g Not examined in St. Lucie Unit 1 FSAR.

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