NOC-AE-000529, Forwards Response to NRC 990414 RAI Re Util Proposed Amend on Radiological Aspects of Operation at Reduced FW Temp & of Operation with Replacement Sgs.Ltr Contains No New Commitments

Forwards Response to NRC 990414 RAI Re Util Proposed Amend on Radiological Aspects of Operation at Reduced FW Temp & of Operation with Replacement Sgs.Ltr Contains No New Commitments

ML20206P237
Person / Time
Site:	South Texas
Issue date:	05/14/1999
From:	Leazar D HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NOC-AE-000529, NOC-AE-529, TAC-MA3820, TAC-MA3821, NUDOCS 9905180195
Download: ML20206P237 (17)

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May 14,1999 NOC-AE- 000529 File No.: G20.02.01 G21.02.01 l 10CFR50.90 l

U. S. Nuclear Regulatory Commission Attention: DocumentControlDesk Washington,DC 20555 South Texas Project i Units 1 and 2 Docket Nos. STN 50-498 and STN 50-499 Response to Request for AdditionalInformation -

Proposed Amendments on Radiological Aspects of Operation at Reduced Feedwater Temperature and of Operation with Replacement Steam Generators, South Texas Proiect. Units 1 and 2 (STP) (TAC Nos. MA3820 AND MA3821)

References:

1) Letter from T.H. Cloninger to U.S. Nuclear Regulatory Commission dated September 30,1998, (NOC-AE-0140)

2) Letter from U. S. Nuclear Regulatory Commission to W. T. Cottle, " REQUEST FOR ADDITIONAL INFORMATION - PROPOSED AMENDMENTS ON RADIOLOGICAL ASPECTS OF OPERATION AT REDUCED FEEDWATER TEMPERATURE AND OF OPERATION WITH REPLACEMENT STEAM GENERATORS, SOUTH TEXAS PROJECT, UNITS 1 AND 2 (STP) (TAC NOS. MA3820 AND MA3821)," dated April 14,1999 Attached is South Texas Project Nuclear Operating Company's response to the referenced U. S.

Nuclear Regulatory Commission Request for Additional Information. Each request is repeated in its original form with the answer immediately following.

There are no commitments contained in this response that have not been specifically made otherwise in separate correspondence. If there are questions regarding this subject, please contact Mr. D. E.

Gore at (512) 972-8909, or me at (512) 972-7795.

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• ibd D. A. Leazar (

i Director 3 p Nuclear Fuel & Analysis C' DEG/MTVN C:WiARKSTUBSGRP.JMes\Submittals\RAls\TSC . 220\TSC_220 RAI. NoC-AE-000529. doc STI: 30870065 9905180195 990514 PDR ADOCK 05000498 P PDR w -

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NOC-AE- 000529 File No.: G20.02.01 G21.02.01 Page 2 l

cc:

Ellis W. Merschoff Jon C. Wood Regional Administrator, Region IV Matthews & Branscomb l U. S. Nuclear Regulatory Comnussion One Alamo Center  ;

611 Ryan Plaza Drive, Suite 400 106 S. St. Mary's Street, Suite 700 l Arlington, TX 76011-8064 San Antonio,TX 78205-3692 Thomas W. Alexion Institute of Nuclear Power ,

Project Manager, Mail Code 13H3 Operations - Records Center l U. S. Nuclear Regulatory Commission 700 Galleria Parkway ,

Washington, DC 20555-0001 Atlanta, GA 30339-5957 '

Cornelius F. O'Keefe Richard A. Ratliff i Sr. Resident Inspector Bureau of Radiation Control c/o U. S. Nuclear Regulatory Commission Texas Department of Health P. O. Box 910 1100 West 49th Street Bay City, TX 77404-0910 Austin, TX 78756-3189 J. R. Newman, Esquire D. G. Tees /R. L. Balcom Morgan,12wis & Bockius Houston Lighting & Power Co.

1800 M. Street, N.W. P. O. Box 1700 Washington, DC 20036-5869 Houston,TX 77251 M. T. Hardt/W. C. Gunst Central Power and Light Company City Public Service ATI'N: G. E. Vaughn/C. A. Johnson P. O. Box 1771 P. O. Box 289, Mail Code: N5012 San Antonio,TX 78296 Wadsworth,TX 77483 A. Ramirez/C. M. Canady U. S. Nuclear Regulatory Commission City of Austin Attention: Document Control Desk Electric Utility Department Washington, D.C. 20555-0001 721 Barton Springs Road

. Austin, TX 78704

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NOC-AE- 000529 Page 1 of 15 l Response to Request for AdditionalInformation l Radiological Asoects of Ooeration at Reduced Feedwater Temocrature and of Operation with Replacement Steam Generators l South Texas Proiect. Units 1 and 2 l

1) Why do the consequences of a locked rotor and rod ejection accident result in an increase only in the whole body and skin doses and not the thyroid?

Locked Rotor Accident The bounding dose analysis for the Model E steam generators (at 420 F feedwater temperature) and the A94 steam generators (at 390 F feedwater temperature) for the locked rotor accident shows that the  ;

thyroid doses are about one-third lower than the current UFSAR analysis, but the whole body and skin doses for LPZ are higher than the current analysis (see Table 4.7-2, Reference 1). Both analyses have been reviewed and contributing factors for the differences are described below:

a) The bounding dose analysis for the Model E steam generators (at 420 F feedwater temperature) and the A94 steam generators (at 390 F feedwater temperature) analysis uses a lower steam release value but a higher steam generator mass (see Table 4.7-1). Use of these values will result in an offsite dose that is about 1/3 lower than the original analysis of the Model E generators (at 440 F feedwater temperature).

b) The analysis for the Model E generators, at 440 F feedwater temperature, properly models the release of nuclides with the steam release from the steam generators and accordingly reduces the activities in the steam generator (a " depleting" model). The bounding dose analysis for the Model E steam generators (at 420 F feedwater temperature) and the A94 steam generators (at 390 F feedwater temperature) analysis, however, conservatively assumes that the steam release from the steam generators does not reduce the activities in the steam generator (a "non-depleting" model).

This assumption results in doses that are higher by about 1/3 at the EAB (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) and about double at the LPZ (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).

c) The analysis for the Model E generators (at 440 F feedwater temperature) was performed in two parts: (1) calculation of the dose from the release ofiodine isotopes alone; and, (2) doses from noble gases. The bounding analysis for the Model E steam generators (at 420 F feedwater temperature) and the A94 steam generators (at 390 F feedwater temperature) determined the contributions from allisotopes at one time. Noble gases dominate the contribution to skin and whole body doses.

At the EAB, the one third decrease in steam release rate is offset by the "non-depleting" versus

" depleting" modeling change. Hence, the skin and whole body doses at EAB are expected to be about the same for both analyses.

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NOC-AE- 000529 Page 2 of 15 However, for the LPZ dose analysis, the "non-depleting" modeling effect is dominant since the time frame is expanded from 0-2 hours to 0-8 hours. Although the flow is reduced by 1/3, the "non-depleting" modelincreases the dose by a factor of 2. Therefore, the doses at the LPZ from the bounding analysis for the Model E steam generators (at 420 F feedwater temperature) and the A94 steam generators (at 390 F feedwater temperature) are expected to be about 1/3 higher (i.e.,

(1-1/3)

• 2 = 4/3, or an increase of 1/3) than the analysis for the Model E generators (at 440 F feedwater temperature). The skin and whole body doses on Table 4.7-2 shows this expec:ed trend.

d) An iodine partition factor of 0.01 was assumed in both analyses for iodine released from the steam generator inventory. The analysis for the Model E generators (at 440 F feedwater temperature) reduced the steam release rate to 1% of the total steam release rate to simulate this effect. However, since the steam release rate is reduced by a factor of 100, there is no significant " depletion" of activity in the steam generator due to the release of steam. This modeling methodology is similar to the "no depletion" assumption used in the bounding analysis for the Model E steam generators (at 420 F feedwater temperature) and the A94 steam generators (at 390 F feedwater temperature).

The bounding analysis for the Model E steam generators (at 420 F feedwater temperature) and the A94 steam generators (at 390 F feedwater temperature) uses a 99% efficient iodine filter on the steam release to model the partition factor. As a result, the differences in steam release masses and steam generator mass ((a) above) determines the differences in thyroid doses. Table 4.7-2 shows that thyroid doses from the bounding analysis for the Model E steam generators (at 420 F feedwater temperature) and the A94 steam generators (at 390 F feedwater temperature) are about 1/3 lower the Model E generators (at 440 F feedwater temperature), consistent with the expected trend.

Control Rod Eiection Accident The control rod ejection accident is modeled with two release paths: (1) all activity is released into the containment building and then to the environment; and, (2) all activity released into the reactor coolant system, to the secondary side via a 1 gpm primary-to-secondary leak, and then to the environment. The dose from each pathway is then added to give a total dose. This methodology includes uncertainties associated with the actual distribution of the released radioisotopes retained in the reactor coolant system or released into the containment building.

Table 4.9-2 of the submittal (Reference 1) shows that the majority of the thyroid dose is from the containment leakage path. During the preparation of this submittal the change in thyroid doses between the analyses was noted. At that time it was determined that the current calculation (as reflected in the UFSAR) contained a math error in the determination of the amount ofI-131 in the fuel pin gap. However, during the preparation of this response (and the response to Question 8, below) it was noted that the additior,alI-131 was due to the use of the use of 0.12 as the gap fraction  ;

for I-131. Since NUREG/CR- 5009 (Reference 2) on high burnup fuel used the factor of 0.12 for  !

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i NOC-AE- 000529 Page 3 of 15 only the fuel handling accident, the revised analysis reduced the gap fraction of I-131 to the original 0.1. This results in a correspondingly lower thyroid dose.

Note that the whole body and beta-skin doses from the containment leakage remain the same between the current analysis and the new, bounding analysis. This is to be expected since the proposed changes (lower feedwater temperature and replacement steam generators) will have negligible effect on parameters describing this release pathway.

The increase in the dose from the secondary side is the result of both a decrease in the time needed to release the contents of a steam generator (300 seconds to 191 seconds) and the longer time needed to equalize pressure between the primary and secondary sides (1250 seconds to 4500 seconds).

The current analysis assumes that the minimum time needed to release the initial steam generator mass of 72,300 lbs is 300 seconds, resulting in an average flow rate of 14,460 lbs/ min. However, the time needed to equalize the pressure between the primary and secondary sides is 1250 seconds.

Therefore, the primary-to-secondary leakage of 1 gpm will continue out to 1250 seconds. The analysis conservatively assumes that the release from the steam generators will continue out to 1250 seconds at a rate of 14,460 lbs/ min.

During the development of the bounding analysis, the source document for the steam releases in the original analysis was questioned. Due to the low radiological consequences of this accident, a conservative set of values for steam releases was independently generated. Instead of determining how much steam would be released, it was assumed that the entire mass of the steam generators would be released. It was determined that the minimum time needed to release the initial steam generator mass of 659,000 lbs (total mass for four steam generators) is 191 seconds, resulting in an average total flow rate of 207,000 lbs/ min. {

The same methodology that was used in the current analysis to determine the rate of steam release is used in the bounding analysis. The bounding analysis assumes that the minimum time needed to release the initial steam generator mass of 659,000 lbs (total mass for four steam generators) is 191 seconds, resulting in an average total flow rate of 207,000 lbs/ min. However, the time needed to equalize the pressure between the primary and secondary sides is increased to 4500 seconds.

l Therefore, the primary-to-secondary leakage of I gpm will continue out to 4500 seconds. The analysis conservatively assumes that the release from the steam generators will continue out to 4500 seconds at a rate of 207,000 lbs/ min, for a total release of 1.55 x 10'lbs of steam. Note that this is

! overly conservative since the steaming rate would decrease significantly over time. A total release of 1.55 x 10' lbs of steam is unrealistically high.

Note that the release from the secondary side has a small effect on the thyroid doses. The major impact is in the whole body and skin doses. The large steam release assumed in the bounding analysis accounts for the increase in the whole body and skin doses.

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2) Explain what the "MSIV [ main steam isolation valve] above seat drains" leak rate is? (sic) ,

i During plant start-up, the above main steam line isolation valves (MSIV) seat drain line valves are l opened for removal of accumulated condensate to protect the turbine from water induction damage and to prevent water hammer in the steam lines. During normal operations, manual valves isolate the above MSIV seat drain lines. Due to the use of restricting orifices (3/8"), flow from the lines is limited and no 1 operator action is required to close the above MSIV seat drain line isolation valves. This results in an additional, minor, steam release path that is considered in the radiological analyses. See UFSAR sections 15.1.5.4,15.1.5.2, and15.6.3.3. I

3) Why wasnt an analysis performed for the pre-existing spike case for the voltage-based analysis?

Based on previous calculations for the voltage-based repair, the preexisting spike cases are not limiting.

The Main Steam Line Break dose analyses for UFSAR Chapter 15.1.5 (see Table 4.4-2 of Reference 1) shows that the doses for the preexisting spike case are significantly lower than the accident-initiated iodine spike case. Additionally, dose limits for the pre-existing iodine spike case are a factor of 10 higher than the accident-initiated iodine spike case. The preexisting spike case was far less limiting and, therefore, was not performed for the voltage based analyses.

During the NRC review of the latest voltage-based repair submittal (References 3,4, and 7), the voltage-based repair analysis performed for this submittal was used to answer NRC staff questions (References 5 and 6). This question was raised on this analysis by the NRC staff. The staff agreed on the results, but requested that in subsequent revisions to the analyses the preexisting spike also be analyzed. STP agreed to that request.

1

4) Why was the duration of the accident for the revised main steam line break (MSLB) analysis 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and that for the voltage-based repair criteria 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />?

It is assumed that 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after an MSLB accident occurs, the residual heat removal system ('RHR) starts and the primary- to-secondary leakage ends. The steam release through the MSIV Above Seat

- Drain orifices (see Question #2, above) continues for a total of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The MSLB dose analysis, however, simplifies the modeling of the accident by assuming that the primary-to-secondary leakage -

continues for the same 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> duration. After 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, all steam releases end. This extra conservatism does not result in a significant dose increase and the results remain a small fraction of the regulatory limits.

However, because of the desire to reduce unnecessary conservatism in the voltage-based repair criteria analysis, the timing of the steam releases is more exactly modeled. The primary- to-secondary leakage ends at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with the initiation of RHR and the steam release through the MSIV Above Seat Drain orifice-s continues for a total of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (see Table 4.5-1 of Reference 1).

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5) For the steam generator tube rupture (SGTR) accident, was an analysis performed of the impact of the reduced feedwater temperature on steam generator overfill, flashing fraction, and break flow?

The RSG overfill analysis included the impact of the reduced feedwater temperature. The reduced feedwater temperature is a (small) benefit for the offsite dose analysis and therefore was not modeled in the analysis.

The scope of this submittalis the radiologicalimpact of operation of the Model E steam generators at reduced feedwater temperatures and the use of the replacement A94 steam generators at feedwater temperatures as low as 390 F. Other impacts of these changes are being handled via 50.59 evaluations and submittals, as appropriate (see Reference 8).

6) For the voltage-based MSLB analysis, what is the accident-induced primary-to-secondary leakage value?

The analysis varied the primary to secondary leak rate in the faulted steam generator to yield a dose equivalent to about 90% of the applicable dose limit for the scenario (pre-existing iodine spike o, accident induced iodine spike), type of dose (thyroid, whole body, skin), and receptor location (EAB, LPZ, or control room). The limiting leak rate (15.4 gpm) was then reported as the maximum accident-induced primary to secondary leakage that could be tolerated. This maximum leakage value is then used as a success criterion when determining the accident-induced primary to secondary leakage value for the voltage-based repair data analysis.

7) Have the control room X/Q values in Table 5.2-1 included the occupancy factors?

The atmospheric dispersion factors in the tables in Section 5 (specifically Table 5.2-1 of Reference 1) do not contain the occupancy factors. The computer code used applies the occupancy factors separately.

8) Does this plant have extended burnup fuel?

Yes, STP has licensed extended burnup fuel (UFSAR Sections 11.1.6,12.2.3, and 15.A.4). The nominal cycle length is assumed to be 20,000 MWD /MTU, and the core average burnup is assumed to be 40,000 MWD /MTU. Reference 9 is the licensing submittal and Reference 10 is the NRC Safety Evaluation Report (SER).

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9) Explain why it takes only 862 seconds to cool the plant down via the secondary side in the event of an SGTR7 (sic) l This is the code-calculated time. This is the time required to cool the reactor coolant system (by l

dumping steam from the three intact steam generators) to the temperature corresponding to saturation conditions at the ruptured steam generator pressure at the start of the cooldown (plus subcooling margin & uncertainties) in accordance with the emergency operating procedures. Additional time is required, and accounted for in the calculations, to cool to the point at which the residual heat removal system may be used to further cool the plant and end the steam releases from the secondary side. This is not done until after break flow termination.

10)What is the basis for halting the iodine spike at 11676 Ci for the SGTR accident-initiated spike case?

This is the I-131 activity corresponding to the pre-accident iodine spike. The Standard Review Plan (SRP) specifies that the accident-initiated iodine spike should be modeled as being 500 times the equilibrium appearance rate. However, the SRP gives no guidance to the duration of the iodine spike.

The typical approach followed in modeling the accident-initiated iodine spike duration when performing accident dose analysis is to terminate the iodine spike when the primary coolant activity concentration is raised to the Tech Spec upper limit identified for transients (60 p/g dose equivalent I-131).

The majority of the dose from a SGTR is from the release of flashed break flow directly to the atmosphere, and from the release of activity collecting in the ruptured steam generator. The release of activity collecting in the intact steam generators accounts for only a small fraction of the total dose.

The analysis shows that the break flow stops flashing at 1970 seconds. After this time break flow is mixed in the steam generator and subject to partitioning prior to release to the atmosphere. This is before the spike was terminated (at 5076 seconds). Break flow is terminated at 5024 seconds. This is also before the spike is terminated. Thus, extending the duration of the accident initiated iodine spike would have no impact on the contribution of the ruptured steam generator to the dose for this case.

Intact steam generator leakage continues for the duration of the calculation (36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />). Extending the spike beyond 5076 seconds would have smallimpact on the site boundary (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) doses by increasing the concentration ofiodine in the intact steam generator leakage. This would increase the activity in the intact steam generators available for release. This impact is small since all releases are subject to partitioning prior to release, and the activity increase is limited to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The impact on the low population, control room and technical support center (TSC) doses would be higher, since the leakage is continuing and the reactor coolant system concentration is allowed to increase. [By 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the I-131 concentration would be about 17 times the pre-accident iodine spike initial concentration.] Still, the impact on the total dose would be small.

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NOC-AE- 000529 Page 7 of 15 A quick, informal, rerun with a non-terminated spike confirmed the small impact of this assumption.

The increase in thyroid dose at the different locations is about:

• EAB: ~0.001 rem

. LPZ: ~0.180 rem, o CR: ~0.030 rem I e TSC: ~0.040 rem i

The impact would be smaller if the spike is terminated at some reasonable value/ time. l l

11)Please provide a Table (s) with break flow and flashing fraction as a function of time for the SGTR. !

Table 11-1, "SGTR Break Flow Data" (Attached), contains the flashed break flow, calculated flashing fraction, and break flow (in that order), from the start of the event until break flow termination.

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NOC-AE- 000529 Page 8 of 15 References

1. Letter from T. H. Cloninger, STPNOC, to the U.S. Nuclear Regulatory Commission (ST-NOC-AE-000140), " South Texas Project, Units 1 & 2, Docket Nos. STN 50-498, STN 50-499, Proposed License Amendment Concerning Radiological Aspects of Operation at Reduced Feedwater Temperature and of Operation with Replacement Steam Generators," dated 30 September 1998.

2. " Assessment of the Use of Extended Burnup Fuelin Light Water Power Reactors", NUREG/CR-5009, February 1988.

3. Letter from T. H. Cloninger, South Texas Project, to the U.S. Nuclear Regulatory Commission,

" Proposed Amendment to Incorporate Voltage-Based Repair Criteria into Technical Specifications 3.4.5" dated 16 February 1998. (STP submittalfor use of Voltage-Based Repair Criteria in Unit 2)

4. 12tter from T. H. Cloninger, STPNOC, to the U.S. Nuclear Regulatory Commission (ST-NOC-AE-000097), " Proposed Amendment to Incorporate Voltage-Based Repair Criteria into Technical Specification 3.4.5," dated 2 April 1998. (Supplement to the STP submittalfor use of Voltage-Based Repair Criteria in Unit 2)

5. Letter from T. W. Alexion, U. S. Nuclear Regulatory Commission, to W. T. Cottle, STPNOC (ST- AE-NOC-000155)," Request for AdditionalInfonnation on Proposed Amendment to Incorporate Voltage-Based Repair Criteria, South Texas Project, Units 1 and 2 (STP) (TAC NOS. MA0967 and MA0968)," dated 18 May 1998.

6. Letter from D. A. Leazar, STPNOC, to U. S. Nuclear Regulatory Commission (ST-NOC-AE-000228),

" South Texas Project, Units 1 and 2, Docket Nos. STN 50 498, STN 50 499, Response to NRC's Request for Additional Information (TAC Nos. MA0967 and MA0968)," dated 15 July 1998.

7. Letter from T.W. Alexion, U.S. Nuclear Regulatory Commission, to W. T. Cottle, South Texas Project )

(ST-AE-NOC-000258), " South Texas Project, Units 1 And 2 - Amendment Nos. 96 And 83 To Facility Operating License Nos. NPF-76 and NPF-80 (TAC NOS. MA0967 and MA0968), " dated 24 September 1998.1996 (Safety Evaluationfor the Voltage-Based Repair Criteria in Unit 2).

8. Letter from L. E. Martin, STPNOC, to the U.S. Nuclear Regulatory Commission (ST-NOC-AE-000159), " South Texas Project, Units 1 and 2, Docket Nos. STN 50-498 and STN 50-490, Licensing Methodology for the Replacement Steam Generator Project," dated 7 May 1998.

9.- I2tter from S. L. Rosen, STPNOC, to the U.S. Nuclear Regulatory Commission (ST-HL-AE-3906), "

South Texas Project Electric Generating Station, Units 1 & 2, Docket Nos. STN 50-498, STN 50-499, Proposed Revision to Updated Final Safety Analysis Report for Extended Burnup Fuel South Texas Project Electric Generating Station (STPEGS), " dated 30 October 1991. (Submittalfor Extended STI: 30870065

NOC-AE- 000529 Page 9 of 15 Burnupfuel)

10. Letter from G. F. Dick, Jr., U.S. Nuclear Regulatory Commission. to D. P. Hall. STPNOC (ST-AE-HL-93097), " Issuance Of Amendment Nos. 38 And 29 To Facility Operating License Nos. NPF-76 and NPF South Texas Project. Units 1 and 2 (TAC NOS. M82128 and M82129)," dated 8 June 1992.

(NRC SERfor Extended Burnupfuel)

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NOC-AE- 000529 Page 10 of 15 Table 11 1 SGTR Break Flow Data Time Integrated Break Flow Break Flow (sec) Flashed Flashing (lbm/sec)

Break Flow Fraction (lbm) >

0.000000 0.000000 0.0000000 0.000000 0.150000 6.496903 0.1498166 43.36570 30.1500 200.2647 0.1499328 42.97000 60.1500 393.0055 0.1501338 42.68090 90.1500 455.4249 0.0286636 37.50500 120.150 499.9442 0.0462745 39.25850 150.150 560.9958 0.0559740 39.12570 180.150 629.8814 0.0619140 38.63740 210.150 703.4112 0.0660111 38.06140 4 240.150 780.5166 0.0689662 37.42280 270.150 857.4511 0.0694740 36.72410 300.150 934.0452 0.0702768 36.39620 330.150 1011.000 0.0709028 36.19900 360.150 1087.909 0.0711380 35.94200 390.150 1164.212 0.0708513 35.68810 420.150 1239.476 0.0703702 35.38400 450.150 1313.528 0.0696594 35.07660 480.150 1385.962 0.0685930 34.73870 510.150 1456.235 0.0671893 34.34380 540.150 1524.447 0.0659719 33.95810 570.150 1590.689 0.0649007 33.57500 600.050 1657.229 0.0660252 33.39340 630.050 1738.864 0.0885547 35.37510 660.050 1845.160 0.1052695 36.81050 690.050 1971.044 0.1177270 37.83050 720.050 2111.899 0.1263981 38.58220 750.050 2263.015 0.1320456 39.16000 780.050 2422.680 0.1375876 39.62800 810.050 2590.489 0.1430397 39.94790 840.050 2765.734 0.1481096 40.15050 870.050 2947.220 0.1523437 40.24510 900.050 3133.371 0.1555937 40.24390 STI: 30870065

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Table 11-1 '

SGTR Break Flow Data  !

Time Integrated Dreak Flow Break Flow )

(sec) Flashed Flashing (lbm/sec)

Break Flow Fraction (1bm) 930.050 3322.720 0.1580006 40.18920 '

960.050 3514.053 0.1593177 40.12630 990.050 3705.701 0.1591412 40.07300 1020.05 3896.421 0.1582881 40.04710 1050.05 4085.911 0.1571360 40.05400 1080.05 4273.602 0.1552043 40.08940 1110.05 4459.144 0.1531686 40.18660 1140.05 4643.057 0.1515863 40.32530 1170.05 4826.244 0.1506461 40.49760 1200.05 5009.203 0.1499205 40.66440 1230.05 5192.007 0.1492402 40.79590 1260.05 5374.349 0.1483070 40.86990 1290.05 5555.692 0.1473035 40.90570 1320.05 5735.659 0.1458514 40.95330 1350.05 5913.989 0.1440364 41.04630 1380.05 6090.692 0.1424033 41.19960 1410.05 6266.173 0.1409777 41.37100 1440.05 6440.987 0.1399538 41.63540 1470.05 6615.977 0.1392381 41.92330 1500.05 6791.484 0.1374408 42.10500 1530.0.5 6952.795 0.1223696 41.28060 1560.05 7094.609 0.1101520 40.56350 1590.05 7220.586 0.0998423 39.93730 1620.05 7334.298 0.0922958 39.37840 1650.05 7439.428 0.0876345 38.86890 1680.05 7539.865 0.0860842 38.38660 1710.05 7637.750 0.0853253 37.89880 1740.05 7733.535 0.0845830 37.39040 1770.05 7826.201 0.0829047 36.07400 1800.05 7912.406 0.0806670 34.28220 1830.05 7990.654 0.0749654 32.57800 1860.05 8056.443 0.0639107 31.00160 1890.05 8108.466 0.0508495 29.30220 STI: 30870065

NOC-AE- 000529 Page 12 of 15 Table 11 1 SGTR Break Flow Data Time Integrated Break Flow Break Flow l (sec) Flashed Flashing (Ibm /sec)

Break Flow Fraction 1 (lbm) 1920.05 8144.290 0.0329773 27.94130 1950.05 8162.485 0.0127711 26.83720 1980.05 8165.649 0.0000000 26.43240 2010.05 8165.649 0.0000000 26.24640 '

2040.05 8165.649 0.0000000 26.02240 2070.05 8165.649 0.0000000 26.17300 2100.05 8165.649 0.0000000 26.60080 2130.05 8165.649 0.0000000 26.96670 2160.05 8165.649 0.0000000 27.31390 2190.05 8165.649 0.0000000 27.63820 2220.05 8165.649 0.0000000 27.93700 2250.05 8165.649 0.0000000 28.22910 2280.05 8165.649 0.0000000 28.49040 2310.05 8165.649 0.0000000 28.72640 2340.05 8165.649 0.0000000 28.96580 2370.05 8165.649 0.0000000 29.09510 2400.05 8165.649 0.0000000 29.30570 2430.05 8165.649 0.0000000 29.51100 2460.05 8165.649 0.0000000 29.66320 2490.05 8165.649 0.0000000 29.76080 2520.05 8165.649 0.0000000 29.97480 2550.05 8165.649 0.0000000 30.06080 2580.05 8165.649 0.0000000 30.11750 2610.05 8165.649 0.0000000 30.46110 2640.05 8165.649 0.0000000 31.19120 2670.05 8165.649 0.0000000 31.76980 2700.05 8165.649 0.0000000 32.31680 2730.05 8165.649 0.0000000 32.69970 2760.05 8165.649 0.0000000 33.00860 2790.05 8165.649 0.0000000 30.84230 2820.05 8165.649 0.0000000 25.37420 2850.05 8165.649 0.0000000 21.06950 2880.05 8165.649 0.0000000 16.63150 STI: 30870065 m

. \

NOC-AE- 000529 Page 13 of 15 Table 11-1 SGTR Break Flow Data Time Integrated Break Flow Break Flow .

Flashed I (sec) Flashing (lbm/sec)

Break Flow Fraction (lbm) 2910.05 8165.649 0.0000000 11.41600 l 2940.05 8165.649 0.0000000 1.325770 2970.05 8165.649 0.0000000 18.92290

)

3000.05 8165.649 0.0000000 26.38710 3030.05 8165.649 0.0000000 30.65010 3060.05 8165.649 0.0000000 32.80420 3090.05 8165.649 0.0000000 33.80200  !

3120.05 8165.649 0.0000000 34.66670 3150.05 8165.649 0.0000000 34.09180 3180.05 8165.649 0.0000000 33.46550 3210.05 8165.649 0.0000000 32.65560 3240.05 8165.649 0 0000000 31.81220 3270.05 8165.649 0.0000000 30.86370 3300.05 8165.649 0.0000000 29.88370 3330.05 8165.649 0.0000000 28.85440 3360.05 8165.649 0.0000000 27.83930 3390.05 8165.649 0.0000000 26.84930 3420.05 8165.649 0.0000000 25.91820 3450.05 8165.649 0.0000000 25.04200 ,

3480.05 8165.649 0.0000000 24.19920 1 3510.05 8165.649 0.0000000 23.38590 I 3540.05 8165.649 0.0000000 22.61340 3570.05 8165.649 0.0000000 21.83670 j 3600.05 8165.649 0.0000000 21.08110 3630.05 8165.649 0.0000000 20.30990 3660.05 8165.649 0.0000000 19.54550 3690.05 8165.649 0.0000000 18.82900 l

3720.05 8165.649 0.0000000 18.19380 l 3750.05 8165.649 0.0000000 17.66170 i 3780.05 8165.649 0.0000000 17.25140  !

3810.05 8165.649 0.0000000 16.96700 3840.05 8165.649 0.0000000 16.74780 3870.05 8165.649 0.0000000 16.55280 STI: 30870065 j I

a

r NOC-AE- 000529 Page 14 of 15 Table 11-1 SGTR Break Flow Data Time Integrated Break Flow Break Flow (sec) Flashed Flashing (Ibm /sec)

Break Flow Fraction (Ibm) 3900.05 8165.649 0.0000000 16.38810 3930.05 8165.649 0.0000000 16.25770 3960.05 8165.649 0.0000000 16.15900 3990.05 8165.649 0.0000000 16.06020 4020.05 8165.649 0.0000000 15.97270 4050.05 8165.649 0.0000000 15.89230 4080.05 8165.649 0.0000000 15.81680 4110.05 8165.649 0.0000000 15.73630 4140.05 8165.649 0.0000000 15.45930 4170.05 8165.649 0.0000000 15.07790 4200.05 8165.649 0.0000000 14.65480 4230.05 8165.649 0.0000000 14.19690 4260.05 8165.649 0.0000000 13.71090 4290.05 8165.649 0.0000000 13.19420 4320.05 8165.649 0.0000000 12.64080 4350.05 8165.649 0.0000000 12.07830 4380.05 8165.649 0.0000000 11.53210 4410.05 8165.649 0.0000000 10.98840 4440.05 8165.649 0.0000000 10.45360 4470.05 8165.649 0.0000000 9.928950 4500.05 8165.649 0.0000000 9.414760 4530.05 8165.649 0.0000000 8.913870 4560.05 8165.649 0.0000000 8.316050 '

4590.05 8165.649 0.0000000 7.735980 4620.05 8165.649 0.0000000 7.027770 4650.05 8165.649 0.0000000 6.498990 4680.05 8165.649 0.0000000 5.995700 4710.05 8165.649 0.0000000 5.513180 4740.05 8165.649 0.0000000 5.054270 4770.05 8165.649 0.0000000 4.180800 4800.05 8165.649 0.0000000 3.992040 4830.05 8165.649 0.0000000 3.308150 4860.05 8165.649 0.0000000 2.659420 STI: 30870065

r NOC-AE- 000529 Page 15 of 15 Table 11-1 SGTR Break Flow Data Time Integrated Break Flow Break Flow (sec) Flashed Flashing (lbm/sec)

Break Flow Fraction (lbm) 4890.05 8165.649 0.0000000 1.610790 4920.05 8165.649 0.0000000 1.208510 4950.05 8165.649 0.0000000 0.985733 4980.05 8165.649 0.0000000 0.912981 5010.05 8165.649 0.0000000 0.935347 5020.05 8165.649 0.0000000 0.404052 4950.05 8165.649 0.0000000 0.985733 4980.05 8165.649 0.0000000 0.912981 5010.05 8165.649 0.0000000 0.935347 5020.05 8165.649 0.0000000 0.404052 l

STI: 30870065 e