ML17229A370

Final Analysis of Radiological Consequences of Main Steam Line Break Outside Containment for St Lucie Unit 1 Nuclear Power Plant Using NUREG-0800 Std Review Plan Section 15.1.5, App a, Rev 1

ML17229A370
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Site:	Saint Lucie
Issue date:	04/29/1997
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
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ML17229A369	List: ML17229A368 ML17229A370
References
RTR-NUREG-0800, RTR-NUREG-800 L-97-141, SAIC-97-1008, SAIC-97-1008-R01, SAIC-97-1008-R1, NUDOCS 9706090313
Download: ML17229A370 (55)

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St. Lucie Unit 1 Docket No. 50-335 L-97-141 Attachment 2 SAIC-97/1008 REVISION 1 ANALYSIS OF THK RADIOLOGICALCONSEQUENCES OF A MAINSTEAM I,INKBREAK OUTSIDE CONTAINMKNTFOR THK ST. LUCIKUNIT 1 NUCLEAR POWER PLANT USING NURKG-0800 STANDARD REVIEW PLAN SECTION 15.1.5, APPENDIX A Prepared for:

APTECH Engineering Services, Inc.

Pittsburgh, Pennsylvania Prepared by:

Science Applications International Corporation Germantown, Maryland and Reston, Virginia April 29, 1997 REVISION 1 FINAL

'77060903i3 970603 PDR ADOCK 05000335

( P PDR

TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 INPUT DATA 3.0 ASSUMPTIONS 4 4.0 CALCULATIONS .6 4.1 Site 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 .

4.1.2 Concurrent Iodine Spike with the Outside Containment Main Steam Line Break Accident ...9 4.1.3 Evaluation of Effects of Fuel Failure after the MSLB Accident ~ ~ ~ ~ ~ ~ 10 4.2 Control Room Dose Calculations ~ 12 4.2.1 Methods of Control Room Analysis . ...... 12 4.2.2 Control Room Source Term Characterization . 15 4.2.3 Results of the Control Room Dose Analysis .......... .... 15 5.0

SUMMARY

AND CONCLUSIONS 23

6.0 REFERENCES

..25 Appendix A 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 A-1 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 . 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 Appendix D Florida Power and Light Letter ENG-SPSL-97-0190, to Steven Mirsky (SAIC) from Carl Bible (FP&L), "St. Lucie Unit 1 Transmittal of Review Comments &

Input Parameters for Revision of Main Steam Line Break Outside Containment Radiological Consequences Analysis - SAIC-97/1008 - File: Engineering Evaluation JPN-PSL-SESS-96-076", dated April 24, 1997............ .. D-1 ~

LIST OF FIGURE Figure 1 Control Room Habitability System Model (t4' LIST OF TABLES Table 1 List and Values of Input Parameters Table 2 Control Room Analysis Iodine Species Release Quantities 17 Table 3 Control Room Analysis Noble Gas Release Quantities 18 Table 4 Control Room Analysis for 1 gpm Steam Generator Tube Leak Rate......... 19 Table 5 Control Room Dose Analysis Results ..................... ~ .20

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 an individual at 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 leak rate. Revision 0 of this report presented results for a case in which pre- and post-accident primary-to-secondary leak rates were assumed to be equal.

The analysis described in this revised report (Revision 1) considers cases in which the pre-accident leak rate is constant at a value of 1 gallon per minute (gpm) and the post-accident leak rate is varied over a range of potential values. Variation of this leak rate parameter identifies a value of post-accident primary-to-secondary leak rate at which predicted doses are equal to acceptance criteria.

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.<'> In addition, The NRC Standard Review Plan (NUREG-0800) Section 15.1.5, Appendix A<'> was 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 'eference 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'reathing rate 3.47E-4 m'/sec.

steam generator hot full power secondary side 127,602 pounds water mass Primary coolant system (PCS) water volume 10,400 1,3 ft'.13E+8 primary coolant system water mass 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 0.1 pCi/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 m

percent failed fuel following an MSLB 1.61 1,18 (4

Table 1 List and Values of Input Parameters (Continued)

Parameters Numerical Values Reference

'0 CVCS PCS let down flow rate gpm CVCS 1-131 decontamination factor 1,000 pre-existing PCS iodine concentration 60 pCi/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 /> concentration control room volume 62,700 ft3 normal control HVAC outside air intake flow 750 cfm control room HVAC isolation damper closure 35 seconds time unfiltered air leakage into the control room cfm control room recirculation flow rate 2,000 cfm 0

atmospheric dispersion factor to control room 4.86E-4 sec./m'9 Control Room HEPA filter efficiency per cent 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 /> 0.6 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 rate'00 per Elemental iodine Organic iodine 95 95 per cent cent Particulate iodine 99 per cent Xe-133 PCS concentration 100/E pCi/g I-131 PCS concentration technical specification 1p Ci/g pre-accident primary to secondary leak 1 gpm 3, 18 a Units and decimal places are carried to match the original references.

The primary to secondary leak rate could change after an accident. The effect of this change is analyzed in this report, see Section 4.1.

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 Section 15.1.5, Appendix A.

Three cases'"'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)).

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.'"

For the pre-existing Iodine spike PCS case (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 will accumulate in the SG secondary coolant water volume until the MSLB occurs. The iodine inventory also includes the initial concentration of 0.1 pCi/g of SG secondary water as specified in the FSAR."'uring this time period, the primary-to-secondary side leak rate is constant at 1 gpm.""

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 operating condition (about 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the accident), iodine and noble gases continue to be released directly to the environment from the unisolable SG with no iodine or noble gas removal inside the SG or main steam line piping. The post accident primary-to-secondary leak rate is constant (although different than the pre-accident primary-to-secondary leak rate of 1 gpm) throughout the accident case even though the PCS pressure is being reduced from 2250 psia to 275 psia which would cause a lower primary-to-secondary leak rate.

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

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 case 1(b) source term assumption 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 case source term assumption 1(c) above was obtained from an ANF report provided by FPAL<"> coupled with the assumption that 1.61%< "> 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 3.4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time period.

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

14. The doses at the site boundary location are calculated for a 0-2 hour exposure to the atmospheric releases. The LPZ doses are calculated for the full duration of the release, which is about 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The control room doses are calculated for a 30-day duration.

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 1.61% 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 The 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 Da = [Ai3i] x [</Q] x [DCFoi-3 x [BR]

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

[0.25 Z( + 0.23 Q] [A,33] x (g/Q]

Where:

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

Ai3 Dose equivalent activity of released I-131, (Ci).

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

'-8 DCFi3i-e I-131 Thyroid dose conversion factor, (rem/Ci).

BR Breathing rate, (m'/sec).

Dw.8. Whole body dose (rem) from immersion in a semi-infinite cloud of Xe-133.

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

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

Aip3 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, (or 2702.8 g/min based on PCS specific density of 0.714), prior to the MSLB accident. [The PCS water specific density of 0.714 is based on the PCS pressure and temperature of 2250 Psia and 575'F, respectively.]

4.1.1 Pre-existing Transient with an I-131 PCS Equilibrium Concentration of 60 pCi/g In this case, 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,<'> this 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<'> limit of 100/E pCi/g. Using the assumptions and the data listed in Section 3.0, and Table 1, the following doses were calculated.

a0 Site boundary (exclusion area boundary)

The following calculations estimate the thyroid and the whole body doses to an individual located at the site boundary and exposed to iodine releases after an MSLB for the period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (or 7,200 seconds), and to an individual at the low population zone area who is exposed to the releases for a period of 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (or 12,240 seconds). The calculations indicated that the limiting primary-to-secondary leak rate after the MSLB accident is 430 gpm. This leakage would not result in a consequences that exceed the SRP criterion of 300 rem thyroid dose.

1. 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 60 x 10~ x 106 x 60 = 1032.39 Ci, [or 858.12 Ci, ifiodine radioactive decay half-life of 8.04 days were to be considered] 2 Initial activity of I-131 in the SCS, 2 x 127,602 x 453.6 x 0.1 x 10~ = 11.58 Ci sum of I-131 in the SCS at the time of the break 1032.39+ 11.58 = 1044.0 Ci, [or 869.70 Ci, if iodine decay is considered; decay of the initial activity of I-131 in the SCS was not considered.)

Iodine release to the SCS after the break (70 seconds when the MSIV closes, and 7,130 seconds after the MSIV closure), ifthere is no change in the primary to the secondary leak rate, 2702.8 x 70/60 x 60 x 10~+ 2702.8 x 60 x 10's x 118.83= 19.5 Ci The thyroid dose based on a continuous 1 gpm leak rate with no credit from iodine decay is:

Pre-break incremental change of iodine activity in the SCS over a small dt is equal to the iodine decay plus the primary-to-secondary I-131 activity transfer, i.e., dA A 2 dt+ C< Q dt, where C, is pre-break PCS I-131 concentration, g primary-to-secondary leak rate, A, iodine decay rate, and A the iodine activity. The equation for the integrated accumulations of I-131 over 106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br /> is:

A= (C~ g/XJx(1.0- e'~" J.

D,= [1044.0+ 19.5] x [8.55 x 10'] x [1.08 x 10'] x [3.47 x 10"]

= [1044.0+ 19.5] x [3.20 x 10']

= 34.1 rem The contribution to the thyroid dose from the initial iodine (0 to 70 seconds) release is 33.45 rem, and that from the releases afterward (70 to 7,130 seconds) is 0.62 rem. The SRP criterion on the thyroid dose is 300 rem. Therefore, in order to not exceed this acceptance criterion, the primary-to-secondary leak rate should be less then 430 gpm, ([300-33.45]/0.62).

~ If the iodine decay in the SCS were to be considered, then the thyroid dose would be:

D~ = [869.9+ 19.5] x [3.20 x 10~] = 28.5 rem In this case, the primary-to-secondary leak rate should be less than 438 gpm in order not to exceed the SRP criterion of 300 rem thyroid.

2. External whole body dose Noble gases are directly released to atmosphere after the break for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (7,200 seconds). Assuming that the primary-to-secondary leak rate is 1 gpm, the noble gas curies release would be:

2702.8 x 100/E x 10+ x 7200/60 = 32 44/ E Ci The value of E for noble gases in the PCS was calculated to 0.23 MeV based on the concentration of the gases given in FSAR 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 141 Ci.

D~ a = [0.25 x 0.08+ 0.23 x 0.15] x [8.55 x 10'] x [141]

=7.0x10 rem The SRP criterion on whole body dose is 25 rem. A very large (greater than 10,000 gpm) primary-to-secondary leak rate is needed to exceed this acceptance criterion.

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

1. Thyroid Dose

~ Iodine release to the SCS after the break (70 seconds when the MSIV closes, and 12,170 seconds after the MSIV closure), if there is no change in the primary-to-secondary leak rate, 2702.8 x 70/60 x 60 x 10 + 2702.8 x 60 x 10 x 12170/60= 33.1 Ci D,= [1044.0+ 33.1] x [7.97 x 10 ] x [1.08 x 10 ] x [3.47 x 10 ]

= [1044.0+ 33.1] x [ 2.99 x 10']

= 3.22 rem Consideration of iodine decay before the accident would result in an even smaller dose than calculated here.'or low population zone, the limiting primary-to-secondary leak rate is 3000 gpm.

2, External whole body dose The total noble gas activity release over 204 minutes (12,240 seconds) with 1 gpm post accident leak rate would be 239.7 Ci. The external whole body dose is:

D~ a = [0.25 x 0.08+ 0.23 x 0.15] x [7.97 x 10 ] x [239.7 ]

=1.1x10 rem The SRP criterion on whole body dose is 25 rem. A very large (greater than 20,000 gpm) primary-to-secondary leak rate is needed to exceed this acceptance criterion.

4.1.2 Concurrent Iodine Spike with the Outside Containment Main Steam Line Break Accident In this case 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'" limit 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 where:

A =C,xM XP+ AR Ap = 0.693 /T,

~R = F x (1.0-1/DF) /V 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 8.04 days, Half life of I-131, (6.93 x 10', sec)

F letdown flow rate, (40 gpm, or 0.667 gals/sec)

V Primary system coolant volume, (10, 400 ft', or 7.78 x 10'allons)

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

M Primary system coolant mass (2.13 x 10'rams) p 1.0 x 10~ (per sec)

~R 8.56 x 10~ (per sec)

C, 1.0 x 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 1.02 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:

a. 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 will be taken for iodine decay and removal.

d. Losses in PCS mass during the accident is negligible.

Based on the above assumptions, the iodine concentration is:

C(t) =C,+Et Where:

K = [1.02/M] x 10 =4.79 x 10'Ci/sec/g t time after the accidents, (sec)

At 120 minutes (7,200 seconds) after the break, the iodine concentration would be 35.5 pCi/g.

Iodine release using this concentration would result in a smaller (a factor of 92 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. The SRP criterion limits the thyroid dose to an individual located at the site boundary to 10% of the 10 CFR Part 100, or 30 rem. The final PCS iodine concentration is 35.5 pCi/g, and, ifthis value were to be conservatively used with a 1 gpm primary-to-secondary leak rate after the accident, the dose to an individual would be 0.37 rem.

To remain below the SRP thyroid dose criterion of 30 rem, the maximum primary-to-secondary leak rate should be about 81 gpm or less. However, over the 120 minutes (7,200 seconds) of release, the average concentration of iodine in the PCS would be 18.3 NCi/g leaUing to a release of 5.9 Ci of 1-131 (DEC). Use of this averaged concentration value would result in a limiting leak rate of 157 gpm.

4.1.3 Evaluation of Effects of Fuel Failure after the MSLB Accident In this case, it was conservatively assumed that the transient following the steam line break would result in 1.61% failed fuel, '" even though the engineering analysis performed in support of the St. Lucie 1 FSAR,'" indicated that no fuel failure is expected in an outside containment MSLB accident. The inside containment MSLB accident analysis resulted in a 1.61% fuel failure. Using the ANF calculation"" of core and fuel gap nuclide inventories, and assuming that 1.61 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 8.77 x 10" Ci/g. Assuming that this concentration would remain constant throughout the accident duration, and the primary-to-secondary leak rate would remain at 1 gpm, a total of 284 Ci of I-131 (dose equivalent quantities) would be released to the environment over 120 minutes (7,200 seconds), and a total of 484 Ci over 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (12, 240 seconds). The thyroid dose to an individual located at the site boundary and low population zone would be:

a. Site boundary (exclusion area boundary) Dose D~ = [284] x [3.2 x 10~]

= 9.1 rem The SRP limiting dose criterion for this accident is 300 rem thyroid dose. Therefore, the primary-to-secondary leak rate should 33 gpm or less in order not to exceed the SRP criterion.

b. Low population zone Dose D~ = [484] x [2.99 x 10']

= 1.5 rem For the low population zone, the limiting leak rate is 200 gpm.

Using the same assumptions as indicated above, an Xe-133 (DEC) value of 1.47 x 10'i/g was calculated. Assuming a constant concentration over the 120, and 204 minutes (7,200, and 12,240 seconds) duration of the accident, and a 1 gpm primary-to-secondary leakage after the accident, an equivalent of 477, and 811 Ci of Xe-133 would be released through the affected steam generator, respectively. ~

a. Site boundary (exclusion area boundary)

The site boundary external whole body dose is calculated using a 2-hour release, or 477 Ci of Xe-133 (DEC) release.

D~a = [0.25 x 0.046+ 0.23 x 0.135] x [8.55 x 10'] x [477]

= 1.73 x 10'em The SRP dose criterion for this accident is 25 rem. A very large leak rate (14, 500 gpm) is needed to exceed this dose limit.

b. Low population zone The external whole body dose is calculated using the releases for 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or 811 Ci of Xe-133 (DEC):

D~a = [0.25 x 0.046+ 0.23 x 0.135] x [7.97 x 10 ] x [811 ]

= 2.75 x 10" rem For the low population zone, an even larger leak rate is needed to exceed the SRP dose limit.

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 case 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. As described earlier, the pre-accident steam generator primary-to-secondary leak rate is constant at 1 gpm, but the post-accident primary-to-secondary leak rate is varied to determine the dose limiting leak rate. 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<"> 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,'XP[-a(t-t)]+([Rj(aV)] ( 1-EXP[-a(t-t)] ))

where:

R; A;+ LP'pCpV A.in C,F, + C2F2+ C3(1-E3)F3+ C4(1-E4)(1-E,)F4 FP Fraction of parent decaying to daughter a (Fo+ EsFs+ LdV)/V C CR concentration at the beginning of a time interval CP CR concentration of the nuclide's parent t 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 C, concentration at filtered inflowpoint 3 filtration efficiency at filtered inflowpoint 3 F, air inflow rate at filtered inflow point 3 C4 concentration at filtered inflow point 4 (CR recirculation train)

E4 filtration efficiency at filtered inflowpoint 4 F4 air inflow rate at filtered inflow point 4 E, CR recirculation loop filter removal efficiency F, CR recirculation/filtration loop flow rate F0 CR effluent flow rate Ld radionuclide decay constant LP radionuclide parent decay constant and V control room volume Control room doses for a time interval in each case are calculated as the time integral of the above concentration multiplied by the appropriate dose factor, breathing rate, and occupancy factor. Case 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 term does change over the time frame of the release event.

Figure 1 Control Room Habitability System Model Outside Air Intake 2 (Fi)

Outside Air Intake 1 (F0 Filter (N,)

Filter (NQ Recirculation Loop (FG Control Room Unfiltered Inleakago 1 (Fi)

Volume Unfiltered Inloakago 2

(~) (FD Filter (N~)

Flow Out (FQ 4.2.2 Control Room Source Term Characterization The estimation of radionuclide source terms for the pre-existing and concurrent iodine spike cases and the 1.61% failed fuel case is described in Section 4.1. For the purposes of control room dose analysis, the pre-existing and concurrent iodine spike case iodine source terms expressed as DEC I-131 are converted into nuclide specific release quantities using the definition of dose equivalent. For iodines, the dose equivalency is expressed as:

deci i j J where:

Adec.i dose equivalent activity of nuclide I DCF; thyroid dose conversion factor for nuclide I A.l activity of nuclide J DCF.J 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 of the pre-existing iodine spike case 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. In all cases, for the 70 to 12,240 second period, iodine species release is proportional to the post-accident primary-to-secondary leak rate. 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 iodine spike cases is 239.7 Ci at a post-accident primary-to-secondary leak rate of 1 gpm. Noble gas release rates increase in proportion to the post-accident primary-to-secondary leak rate. 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 1.61% 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. Iodine and noble gas nuclide release estimates for the 1.61% failed fuel case at 1 gpm post-accident primary-to-secondary leak rate 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 and concurrent iodine spike and the 1.61% 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 were calculated for a range of values of post-accident primary-to-secondary leak rate. Dose impacts for a 1 gpm post-accident primary-to-secondary leak rate and for a post-accident primary-to-secondary leak rate at which the control room dose criterion would be reached are presented in Table 4 for the three release cases. In each case, the dose criterion for the thyroid (30 rem) was limiting. The lowest allowable post-accident primary-to-secondary leak rate of 6.8 gpm was estimated for the 1.61%

failed fuel case. This occurs because the PCS radionuclide inventory is much greater for the failed fuel case than for the iodine spike cases. For example, PCS initial inventories for the pre-existing iodine spike and 1.61% failed fuel cases are 12,780 and 186,600 iodine-131 dose equivalent curies, respectively. More detailed computer output for the 1.61% failed fuel case is presented in Table 5. Approximately 90% of the dose for this case occurs in the first 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the accident. Control room unfiltered air inleakage accounts for approximately 80% of the dose.

As indicated in Table 4, limiting primary-to-secondary leak rates for the iodine spike cases were much higher than that estimated for the failed fuel case. The analysis indicates that thyroid dose is dominated by the iodine nuclides but that both iodine and noble gas nuclides contribute to skin and whole body doses.

Table 2 Control Room Analysis Iodine Species Release Quantities for 1 gpm Post-Accident Steam Generator Tube Leak Rates Releases (Ci)

Pre-Existing Iodine Spike Concurrent 1.61%

Iodine Spike Failed Fuel Nuclide Time Period (s) 0 to70 0 to 12,240 0 to 12,240 0 to 12,240 I-131 836.4 862.7 13.4 425.0 I-132 215.0 221.8 3.4 89.0 I-133 1,192.7 1,230.3 19.1 323.0 I-134 132.5 136.7 2.1 83.0 I-135 568.4 586.3 9.1 173.0

Table 3 Control Room Analysis Noble Gas Release Quantities Releases (Ci)

Pre-Existing and Concurrent iodine Spike 1.61% Failed Fuel Nuclide Time Period (s)

Oto70 0 to 12,240 0 to 12,240 Kr-85m 0.011 1.92 7.1 Kr-85 0.007 1.20 7.0 Kr-87 0.005 0.96 6.8 Kr-88 0.018 3.12 15.8 Xe-131m 0.011 1.92 1.9 Xe-133 1.265 221.24 305.0 Xe-135 0.052 9.11 18.2 Xe-138 0.003 0.50 11.5 Table 4 Control Room Analysis for 1 gpm and Limiting Post-Accident Steam Generator Tube Leak Rates Dose (rem)

Tissue Pre-Existing Concurrent 1.61%

Iodine Spike Iodine Spike Failed Fuel 1 gpm 67 gpm 1 gpm 196 gpm 1 gpm 6.8 gpm Thyroid 9.9 29.5 1.5E-1 30.0 44 30.0 Skin 9.1E-3 2.0E-1 2.9E-3 5.7E-1 7.4E-3 5.0E-2 Whole Body 1.3E-4 1.5E-3 2.0E-S 4.0E-3 8.5E-5 5.8E-4 Table 5 Control Room Dose Analysis Results for the 1.61'ailed Fuel 6.8 gpm Post-Accident Primary-to-secondary Leak Rate TIME STEP START: 0.000000E+00 hr TIME STEP END: 2.000000E-02 hr Building cross sectional area (m"2): 0.000000E+00 Building height (m): 0.000000E+00 Release height (m): 0.000000E+00 Effluent vertical velocity (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 height (m): 0.000000E+00 Winds eed (m/s): 0.000000E+00 Vertical dis ersion class:

Horizontal dis ersion class:

X/Q (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 efficiency 01: 'ilter

'lemental fraction .0000 or anic fraction .0000 articulate fraction .0000 Recirculation fiow rate (m"3/s): 9.440000E-01 Recirculation filter efficiency:

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

elemental fraction .0000 or anic fraction 0000 articulate fraction .0000 Bottled air flow rate (m"3/s): 0.000000E+00 Control room volume (mA3): 1775.200000 CUMULATIVE DOSE END TIME WH BODY REM SKIN THYROID LUNG BONE LIVER HOURS REM REM REM REM REM 2.000E-02 5.818E-08 4.331E-06 4.094E-03 1.262E-06 1.614E-05 2.391E-05 3

This flowrate and the filter efficiency reflect the unfiltered inleakage as modeled in code.

Table 5 Control Room Dose Analysis Results for the 1.61% Failed Fuel 6.8 gpm Post-Accident Primary-to-secondary Leak Rate (Continued)

TIME STEP START: 2.000000E-02 hr TME STEP END: 3.400000E+00 hr Building cross sectional area (m 2): 0.000000E+00 Buildino height (m): 0.000000E+00 Release height (m): 0.000000E+00 Effluent vertical velocity (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 height (m): 0.000000E+00 Winds eed (m/s): 0.000000E+00 Vertical dispersion class:

Horizontal dis ersion class:

X/Q (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): 3 4.700000E-02 Filter efficiency 01:

'lemental fraction .0000 organic fraction 0000 articulate fraction .0000 Recirculation fiow rate (m"3/s): 9.440000E-01 Recirculation filter efficiency:

elemental fraction .9500 organic fraction 9500 articulate fraction .9900 Filtered intake flow 2 (feeds recirc): 2.120000E-01 Intake 2 filter efficiency:

elemental fraction .0000 or anic fraction 0000 articulate fraction .0000 Bottled air flow rate (m"3/s): 0.000000E+00 Control room volume (m*3): 1775.200000 CUMULATIVE DOSE END TIME WH BODY REM SKIN THYROID LUNG BONE LIVER HOURS REM REM REM REM REM 3.400E+00 4.510E-04 3.744E-02 2.627E+01 18.880E-02 1.777E-01 2.294E-01 Table 5 Control Room Dose Analysis Results for the 1.61% Failed Fuel 6.8 gpm Post-Accident Primary-to-secondary Leak Rate (Continued)

TME STEP START: 696.000000E+00 hr TIME STEP END: 720.000000E+00 hr Building cross sectional area (m 2): 0.000000E+00 Building height (m): 0.000000E+00 Release height (m): 0.000000E+00 Effluent vertical velocity (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 height (m): 0.000000E+00 Winds eed (m/s): 0.000000E+00 Vertical dis ersion class:

Horizontal dis ersion class:

X/Q (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) 4.700000E-02 Filter efficiency 01 elemental fraction .0000 organic fraction 0000 aruculate fraction .0000 Recirculation flow rate (m"3/s): 9.440000E-01 Recirc filter efficiency:

elemental fraction .9500 organic fraction 9500 articulate fraction .9900 Filtered intake flow 2 (feeds recirc): 2.120000E-01 Intake 2 filter efficiency:

elemental fraction .0000 organic fraction .0000 articulate fraction ~ 0000 Bottled air flow rate (m~3/s): 0.000000E+00 Control room volume (m~3): 1775.200000 CUMULATIVE DOSE END TIME WH BODY REM SKIN THYROID LUNG BONE LIVER HOURS REM REM REM REM REM 7.200E+02 5.790E-04 5.021 E-02 3.001E+01 6.415 E-01 4.313E-01 4.813E-01 5.0 SUKIRIARYAi ID 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 Section 15.1.5, Appendix A (4). This analysis was performed in the following three steps.

First, using the conservative assumptions in the SRP, three cases were analyzed to calculate the radioisotope source term released to the environment from this postulated accident. These cases are: (1) pre-existing iodine spike, (2) accident induced iodine spike, and (3) accident induced 1.61% fuel failure.

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 fr'om 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 conser vative assumptions regarding the control room HVAC system operation and.the performance of radionuclide absorbing filters in the HVAC system, and conservative atmospheric dispersion parameters.

The bounding results of this analysis are tabulated below and show the maximum allowable post-accident primary-to-secondary leak rate to reach the regulatory limits set forth in 10 CFR 100""

and 10 CFR 50 GDC 19"" as indicated in SRP Section 15.1.5, Appendix A. Since the thyroid dose is more limiting than the whole body dose, the following table only presents the limiting results.

Regulatory Thyroid Dose Limiting Post-Accident Primary-to-secondary Leak Rate Dose Location Regulatory Concurrent 1.61% Failed Thyroid Dose Iodine Spike Iodine Spike Fuel Case 300're-Existing Limit (rem) Case (gpm) Case (gpm) (gpm)

Control Room 30 67 196 6.8 EAB (Site Boundary) 30 or 300'0 430 157 33 LPZ or 3,000 988 200 For the concurrent iodine spike case, the regulatory dose limit is 30 rem; for the other cases the regulatory dose limit is 300 rem.

These results show that the limiting post-accident primary-to-secondary leak rate is 6.8 gpm which is for the 1.61% failed fuel source as applied to the control room thyroid regulatory dose limit.

A licensing analysis of the radiological consequences 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.

6.0 REFERENCES

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.

NUREG-0800, USNRC Standard Review Plan Section 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.

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.

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

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 &

Parameters - File: Engineering Evaluation JPN-PSL-SESS-96-076 Rev. 0", dated 'nput 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.

14. NUREG/CR-5659, Control 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.

18. Florida Power and Light Letter ENG-SPSL-97-0190, to Steven Mirsky (SAIC) from Carl Bible (FP&L), "St. Lucie Unit 1 Transmittal of Review Comments & Input Parameters for Revision of Main Steam Line Break Outside Containment Radiological Consequences Analysis - SAIC-97/1008 - File: Engineering Evaluation JPN-PSL-SESS-96-076", dated April 24, 1997.

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

[REFERENCE 9]

Cg Xnter-Of fice Cora esyondence NF-97-065 To: G. L. Boyers Date: February 18, 1997 r

From: J. N. Kabadi Department; ENG/NF/JB

Subject:

Initial Steam Generator Mass for St Lucie Unit 1 MSLB Dose Calculations Reference; JPN Calculation PSL-1FJF-96-156, 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 Yerifled Sy:

Prepared By;

+~+

Mi~il~~

Distribution:

'C. J. Buehrig K. R. Craig C. G. O'Farrlll C. Vlliard

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 8c Verification of Values 4 Input Parameters - File: Engineering Evaluation JPN-PSL4ESS-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 Yatues & 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.

lf 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 A4B CRB/GLB I~a, K.R. Craig

Attachment:

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

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

W. Hannaman (SAIC) an fpi. Group company

ENG.SPSL-97-pp68 C.R. Bible to SAICPage 2 of 5 ITEM PARAMETER SOURCE REVIEWERS COMMENTS Site Boundagy X/Q FSAR Sec 2.3.4.2 Values Correct, References Correct Change Reference to: FSAR Sec.

sec/m'AIC 8.55 E-5 sec/m~

Low 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 24hr, 3.47 E-4 m'/sec 1.75 E-4 m'/sec, 24 to 720 hr, 2.32 Reg Guide 1.4 E-4 m'/sec 8 to 24hr: Note: Accident does not use 2.32 E-4 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 EA m'/sec 1-131 Thyroid Dose Factor Value and reference is correct.

1.08 E+6 rem/Ci lnha/ed ICRP 30 Change: ADD: "Inhaled" Primary 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 5 Reference Correct 1 uCi/gm... fig. 3A-1 Change Unit of measure to uCUgm 60 uCi/gm... 5 add the "(DEC)"

1-131 (DEC)conc in FSAR tbl 15.2.11-5 secondary: Value Correct 0.1 uCi/gr Reference Correct Change Unit of measure to uCVgm 5 add the "(DEC)"

9 PCS Gross Activity: Technical Specification Value Correct 100/ Ebar 3.4.8 Reference Correct Reviewed By: Verified By:

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) Cl ICRP 38, "Radio...." for the energies and yields.

No accumulation ot 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 II of 0.23 + Ebar Y 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 lodlnes "

!odines = 1000 19 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 teatures 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 Untiltered Inleaka 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. Reterence Correct Reviewed By: ~4 ~W Date:~>5' Veritied By: C Z.-2t "

e

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. Lucie 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. Lucle Unit 1 FSAR Table 15.4.6-2 PSL-1FJF-95-155, Rev. 1 12 Cooldown Duration St. Lucie Unit 2 analysis showed that St. Lucia 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 ls 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. Lucie Unit 1.

13 Steam Generator Hot 127,602 Ibm PSL-1FJF-95-155, Rev. 1 Full Power Secondary Side Water Inventory Iodine and Noble Gas Iodine: Use FSAR Table values+ 5% St. Lucie Unit 1 FSAR Table 12.1-3 Releases for 1% Failed Fuel Noble gas: Use FSAR Table values Reviewed By: /7 Date:~i~ Verified By: 2 2I,)1 3~y g~ bn.d,

ENG-SPSL-97-0068 C.R. Bible to SAICPage 5 of 5 ITEM PARAMETER SAIC SOURCE REVIEWERS COMMENTS 18 Fraction, of, Core Fuel;,:

-"'ailure 2.04lo St'ucie Unit 1 FSAR Section 15.4.6 The St. Lucie Unit.1 inside the ANF48-.113(P) July,1988 containment steam line break analysis has 1.61% failed fuel for the worst case. It has been stated Iri ANF-88-113(P), page 13, that.no fuel failure is expected for'outside the containment steam line break" Also per FSAR Section 15 the',containment'steam line 4;6,'nside break Is more. severe in'erms of fuel failure'han outside,tlie'.containment steam line break event .. A'value.of 2% fuel failure is, therefore conservative. fo' u'se tn:the'dose calculations for'o'utslde.'the coiitainment steam line b'reap 17 RCS Uquid Volume 10,400 ft St. Lucie Unit 1 FSAR Table 5.1-1 This value which appears in the St. Lucie 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~.

'" "ev4t'~i.."'.pppwpAs,~~~wR~i..

444~% ilf~~~l.~q"5~sr~

~

anadeci::iiems','repre'senffchanaed vaiues.

Reviewed By Date:

2. 2g ~ ~ Verified By: C ~

.Z zf,'f t

24'97 15:S4 No,035 P.02 Science Appllcetlene International Corporation An Emptayeo Owned Company February 24, ) 997 Chris Buehrig Florida Power and l.,ight Conipany 6501 South Ocean Diivc Jensen Beach, Florida 34957

Dear Mr. Buehrig:

ln accordance with our milestone schedule for the MSLB outside containment SIP analysis task or St. Lucie 1, am providing the faHowing list nf input data values and associated data sources 1

that we have assumed for tltis analysis. To meet our schedule, please review these parameters and their assumed values and confirm in writing that they arc appropriate ior this licensing conservative analysis no later than February 26, 1997.

'List and values of input parameters Parameters Vnfues 0-2 hour X/Q (site houndary) 10'ec/m'ource 3.55 x 10" scc/m'.97 St Lucie l FSAR Section 2.3.4,2 0-8 hours QQ (low x St. Lucic FSAR 1

population xone) ~ Scctioll 15.4, 1 able 15.4.1-4 Breathing rate 3.47 x 10'm"/sec for 0 tg hrs.; 2.32'or R-24 USNRC Rag Guide 1.4 hrs., and 1.75'or 24 to 720 hrs. (c.2,c) l-131 Thyroid dose LOS x 10'em/Ci ICRP Publication 30.

conversion factor Primary to secondary 1 gpm for 2,702.8 gr/min (primary systeni condition; St. Lucic 1 FSAR leak rate and HFP specific density of 0.724 based on 2250 psia and 575 Section Tcchnical primary coolant F)] Specification I.cak 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 1 pCi/8r, St Lucic 1 Teel) Spec operation at above 1 aiid 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.j 3.4.8 pCi/gr of I-131 (DEC) 1-131 concentration s 1 pCi/gr Tech Spec. Limit. with a maximum of St. Lucie I '1'ech. Spec in Primary Coolant 60 pCi/y'n mode 1 with a ~ 80% power level Section 3.4.8, and System (PCS) Figurc 3.4-1 2020t Centuty Boulevent, Gennantewn, Matyland 20874 o (30'53@f50 Othe Sile Cvf~'Abuqvenwt, cobrgoospnngy, conn. Alo Church. IMl~o too vtgar, LooAAe. tooAnpew, NctoaL oak$ 4o. &Oooo. Son Ooyo Sootllr. factm

FFB 24'97 15:34 No.035 P.03 g g, -S P> L

- 97-Cc'4 X EW Paratneters Values Source 1-131 concentration St. Lucic I FSAR table 8 in Secondary (.'oolant I S.2,11-5 System (SCS) c{ PCS specific activity ~

.100/E (E is thc sum of the average P and y cnei'gics R. I.ucic Tech. Spec.

1 per disintegration )MEV] for isotopes other than 3.4.8 lodidc's.)

Steam linc brcak lnalil assitlnptiol) ik tllat 00"la of iodine iii thc SCS 1

t> outside eontainmcnt: (initiiilamount.pink tliat transferred aAcr iodine spike equilibrium in thc I'CS) tvould bc re)cased to thc atmos phcrc.

1. MSlV closure time a) MS1S 63.9 seconds aQcr tlic brcak, tvith a I.OOP SL Lucic 2 FSAk and tinlinp after turbine trip Table I S,1,5.1-1 b) MSIV closure G seconds at4',r MSlS St. Lucie I CESAR Table 15.4.G-2

2. Cooldown 12,240 seconds, (using SL Lucic 2 tim<<duration to St. I.ucic 2 VSAR

~ g duration after brcak shutdown cooling, from condition a above.) '1'able 15,1.5,1-1 Steam generator hot 127,602 lb Kabadi tn Boycrs full power secondary memo dated 2-18-1997 side water inventory (FP&I.)

Nable gasrrdcssc no accumulation of noble gases in thc S(j prior to Average E values werc brcak. Noble gases arc directly rcfcascd to calculated based on atmosphere aAer thc brcak for 204 minutes. (lata provided in

-~ 2702.8 x 100/E x 10 ' 204 55,14/I: Ci ICR1'ublication average E for noble gases pcr PCS concentration of 38,"Radionuclidc FSAR Table 11.1-1 is 0.23 (0.15!E ] + 0.084 ) I>]) Traitsforinations For this average cncrgy, the Ci released would bc 1;ncrgy and lnteitkity 239.7 (5 S. I 4/0.23). 'I'his Ci will bc used to reprckcnt 1'.missions", St. Lucio I thc Xc-131 (DLQ) release. FSAR 'l'able I I. 1- I iodine and Noble gas 1-131 (DEC) 7.0 ItCi/gr. (Thyroid equivalency) St. Lucic 1 FSAR rcleasc fur 1% failed Xc-I33 (DE(:) 3S3.G ItCi/gr. 'l'able 12 1-'t and fuel USNRC R.6'. 1.4 Fraction of core fuel 1.61 % SL I.ucic. FEAR 1

failure for the MSLB Table 15.4.6-4

~

P(.'S liquid water 10,400 cubic feet SL I.ucic 1 FSAR volume ~

Table 5.1-1

FEB 2d '97 15: 35 Na . OSS P . Od 1D:

gg>QI~IYlg.i p 7 Q l~6-585 L - 9 7-c,~g.g E~i Paranteters Values Source CVCS PCS letdown Qow rate = 40 Iym St. Liicic I FSAR i8 radwnste CVCS DI'-1000 Sectinii 9.3.4, and rein oval Table 11.2-4 Con(rot room 11VAC a) SIAS on low prcssuriicr pr<<ssurc at 66.1 sc<<. Nt. Lui.ic 2 I:SAR damper closure time '1'nbtc 15.1.5.1-1 b) damper closure tiinc aAcr signaI is 35 scc.(i.c. St. Lucic I I:SAR control rooin dauiper <<loses n( 101.1 scc. Aller Section 9.4 MSLB)

Control I(oom ) IVAC n) normnl untiltcrcd ninkcup flow rate is 750 cfm St. l,ucic 1 FSAR Page dcsigii features 122-9 b) rccirculntioii flow rate is 2000 clm St. Lucic I FSAR Page 12.2-$

c) 0 to & hr. A tniosphcric dispersion factor 4. &6E-4 SL Lucio l I'SAR Page sccjcubic inctcr, &-24 hr value&.17';4, 24-~Xi hr 15.4.1-17 value 1.688-4,96-720 hr value ~ 6.36r:-5 d} cllnfconl filler iodine rcnioval coicicncy 95% for S(. Lucio 1 FSAR e)cmental and organic nnd 99% for particulate Section 15.4.1 c) occupancy factor 1.0 for 0-24 1trs, 0.6 for 24-96 St. Lucic I I'SAR hrs., 0.4 for 96 tn 720 hrs. Scc(ion 15.4,1 I) controlled post-accident filtcrcd inflow 450 <<fin St, Lucio 1 FSAR Section 9.4 Colltl'ol Roolii 62.700 cubic fcct SL l.ucic 1 FSAR Page Volume 12.2-&

Con(rol Room I QD cfin St. Lucio 1 FSAR 2 Z- Unfiltered Inlenknge Sec(inn 15.4.1 Iodine Chemical 91% elemental, 4% organic. 5% particulate S(.Lucio FSAR 1

Composition Section 15.4.1 Control Room HEPA 99% tlSNRC R.G. l.52 filter efficiency

ID: FEB 24'97 15:35 No.035 p.05 Pg~cA ~g.iP g~ p). 5pgt - F7 c'Pc Q

'I'he above list constitutes the second milestone of this task. On February 27, 1997, I will bc verbally communicating our third milestone wltich will be the preliminary results of our analysis.

Sincerely, Steven M. Mirsky, P,J..

Manager, Nuclear I'acilitics Safety cc: . Jim Beglcy, Al"I'ECH Gary Doyers (FPE'.L}

Bill Hannarnan (SAIC)

C t

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 Vnit 1 Assessment of Radiological and Rod Bow Effects for Increased Burnup", July 1988, Advanced Nuclear Fuels Corp., March 11, 1997

[REFERENCE 17]

Fll CLEAP, FUEL JB SHOUT OHi@5tAL ~8K SK QiSGViQHH . YK +

IF g~ p~~ABQNG- HH PBX<P BY NYCT BQSNERS GAY "UHQGTB"KNAVE ~

smHsRN.

SRIQ TQ'EEWQNK NQSNSh

0 0

TABLE 3.3 CORE AND GAP FISSION PRODUCT ACTIVITIES ~) j.4e craig Design Base 14xi4 Fuel Assembly ANF 14x}4 Fuel Asseibly Design EOL Averaged Core Exposure of 25 GMd/HTU EOL Average Core Exposure of 40 GMd/HTU Fraction of Fraction of Curies in Activity Curies Curies in Activity Curies Mhp Mm. ~im I-}29 2.32E+00 .166 3.84E-01 3.68ft00 .217 7.99E-01 I-131 7.BSE+07 .104 8,20E<06 . 7.89E<07 .129 1.02E+07 I-132 1.12E+08 .014 1.57E>06 1.12Ew08 .019 2.14E+06 I-133 1.48E+08 ;04 5.92Et06 }.46ftOS .053 7.76E+06 I-134 1.69E108 .009 1.52E<06 1.66E+08 .012 1.99E<06 I-135 '.31E+08

.024 3.16E+06 1.30E<08 .032 4. 15E+06 Cs-134 2.75E>07 4. 26E+06 .

Cs-134N &.6}E+06 .012 7.93Ew04 2.88E~OI .155 4.46E+00 Cs-135 Cs-136 Cs-137 Cs-138 6.43E<06 1.08E+D7 1.38EIOS

'155 .098

.155

.005 6.3}E+05

1. 67E+06 6.91E+05 Cs-139 1.36E+08 .OD3 4.07E~05 Cs-140 1.24E+08 .0009 1.12Et05 Cs-141 8.77E+07 .0006 5.26E+04 Cs-142 7.03E<07 .0002 1.41E104 3.6}E+07 .0002 7'. 21E+03 Cs-143 Te-123M 7.61E+0} .336 2.56fwOI Te-125M 3.07E405 .3 9.20B.04 ~

Te-}27 6.63E+06 .071 4.1IE ~05 Te-127H ).42E+06 .333 4.74E~05 ~~

O~9

~

ss ' 'I ~

TABLE 3.3 CORE AND GAP FISSION PRODUCT ACTIVITIES (CONT.) LL~ 'gi Design Base 14x14 Fuel Assembly ANF 14xl4 Fuel Assembly Design EOL Averaged Core Exposure of 25 GMd/HTU EOL Average Core Exposure of IO Ggd/NTU n

Fraction of Fraction of Curies in Activity Curies Curies in Activity furies UdKQ ~ha Te-129 3.08007 .027 8.33905 Te-l29H 5.26E~06 .271 1.43Ew06 Te-131 6.92Et07 .017 1.18E+06 sg Te-13)H 1.18E+07 .112 1.32E106 I LI Te-132 1.08E>08 .154 1.67Ee07 Lsl Te-133 4.17007 .012 5.00E+05 fsl Te-133H 1.14E+DB .024 2.74E<06 s5)

Te-134 1.47E+08 .022 3.24E+06 Te-135 1.27E408 .002 2.55E+05 L (sl iO Ig Kr-85 7.33E>05 .102 7.47B04 1.12E+06 .149 1.68f>05 Vl Kr-85H 1.79E007 .008 1.43&05 1.70E+07 .01 1.7DEw05 Kr-87 3.46E+D7 ,004 1.38Ew05 3.24E+07 .005 1.62Ew05 Kr-88 5.04Ee07 :006 3.02E405 4.75E+07 .008 3.80B05 Kr-89 6.30F+07 .0008 5.04&04 5.91E+D7 .001 5.91fw04 Xe-13N 6.38E>05 .055 3.5IE<04 6.43f+05 .072 4.63E+04 Xe-133 1.48E<08 .038 . 5.63E>06 1.46E+08 .05 7.32E+06 Xe-133H 3.5BE<06 .025 8.95E+04 3.56E+06 .034 1.21E+05 Xe-135 3.13E>07 e011 3.44E+05 3. l2E+07 .014 4.36E~05 ll I

Xe-135H 3.97007 .002 7.94Ew04 3.94E+07 .002 7.87E404 OCI Co Xe-%37 1.44E<08 '0009 1.30E+05 1.43E+OB .OOI 1.43E105 O t PJ Xe-138 1.38E+08 .002 2.76E~05 1.37E+08 .002 2.75E+05 ill Gas

~ Hot examined in St. Lucie Unit 1 FSAR.

I I:l 0

Appendix D Florida Power and Light Letter ENG-SPSL-97-0190, to Steven Mirsky (SAIC) from Carl Bible (FP&L), "St. Lucie Unit 1 Transmittal of Review Comments A Input Parameters for Revision of Main Steam Line Break Outside Containment Radiological Consequences Analysis - SAIC-97/1008 - File: Engineering Evaluation JPN-PSL4ESS-96-076", dated April 24, 1997

[REFERENCE 18]

ENG-SPSL"97-0190 APR 2 4 $ 997 Scientific Applicatrons International Corporation 20201 Centu~ Boulevard German town, Maryland 20874 Attention: Mr. Steven M. Mirsky Manager, Nuclear Pacx 1ities Saf ety St, Zucie Quit 1 Trans'.ttal of Review Coameuts fe Zuput Parameters for Revision of Main Steam Line Break Outside Contaimoeat Radiaioqf.cel Consequences Analgesia - SAIC-97/1008

Reference:

1. SAIC Report SAIC-97/1008. AnalysiS of the Raliologic'al Consequence's of a Main Steam Line Break Outside Containment for the St. Lucio Unit 1 Nuclear Power Plant (Tsing NURZG-4800 Standard Review Plan 15.1; 5 AppencUpc.A.

sAIC Fax, steve Mfrsky to Rick Noble O'PL) Bated AP'ril 24, 1997 PPL PurChaae Order 00019096, Blanket ReleaSe 002 tO APTECH Engineering Services Gentlemen:

This letter formally transmits to you our review comments of SAIC report SAIC-97/1008. Zn addition, provided is the input parameter for percent failed fuel. These comments and input value are to ba used by SAIC to'revise the dose assessment for MSLB outside containment in accordance with the guidance in SRP 15.1.5. This work is authorised by PPL in Reference 3.

If you have any further guestions or need additional information contact Chris Buehrig at (561) 467-7507 or Rick Noble at (561) 467-7022.

SrncQrelye Carl R. Bible Engineering Manager c &

CRB/C

Attachment:

Review Comments (2 pages)

K.R. Craig M. Hannaman (SAIC)

Begley (APTECH) on fPL Group coupon'f

'PC7 I~ f I ENG" SPST 0190 Page 2 of 3 Reviepr of SAXC=97/3.008, March 14, 1997, The primary to secondary leak rate prior to the HSLB accident should be a constant 1.0 gpm whi.ch is. consistent with the Technical Specification maximum, This initial leak is not affected by the MSLB accident. Therefore, the initial accumulated steam generator iodine inventories available for release will not vary with the" accident initiated leak rate.

2. The post-accident primary to secondary leak that is assumed, to be initiated by the MSLB should be varied from a minimum of 1.0 gpm up to a maximum value that result in .radiological doses that would not exceed the most limi.ting of either 10CFR100 (as stated in the SRP 15.1.5 acceptance criteria) or GDG criteria.

3. The site boundary dose should be calculated for a 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration rather than the duration oK the accident, 4, A mathematical error was introduced on page 9 of the 'report.

The 1-131 production rate should be 1.02 Ci/sec rather than 0.102 Ci/sec. This leads to an iodine concentration of approximately 59.6 uCi/g at 204 minutes rather than 6.86 uCi/g as stated in the report. The conclusion that the concurrent iodine spike and MSLB is bounded by the pre-existing transient should be revisited. Zn addition, Standard Review Plan 15.1.5 states that the acceptance criteria for this case (e.g., MSLB with an assumed accident initiated iodine spike) is that the calculated doses should not exceed a small fraction (3.0%) of 10CFR100 or 2.5 rem and 30 rem for whole-body and thyroid doses respectively. This should be reflected in the report conclusions for completeness.

A typographical error exists on gage 9, The X-133:

concentration shoed be 1.089 x 10 Ci/g rather than the stated 1.089 x 10 Ci/g. This error did not affect the iodine released.

ENQ-SPSL-97-0190 Page 3 of The control room model used reflects the St. Lucie 1 control room, It is suggested that the unfiltered inleakage of 100 c fm (4 7E-02 /s) be moved from a filtered intake f low" (with filter

~

efficiencies equal to zero), Table 5, to an flow". This is not critical, since "unfiltered intake not change the results, but since the model is going it it'hould to he re-run anyway, would make the report clearer,

7. The summary and conclusions should be revised to incorporate the previous comments and provide the maximum post-accident primary to secondary leak that can be tolerated without exceeded the radiological doses allowed by the regulations.

8. The percent failed fuel value to use for the recalculation should be 1.61%.

go~

Reviewed By: Verified By:

Date: Date:

l' V