ML17312A763

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Calculation Summary of Radiological Doses for SG Tube Rupture W/Loss of Offsite Power & Stuck Open Adv.
ML17312A763
Person / Time
Site: Palo Verde  Arizona Public Service icon.png
Issue date: 05/08/1996
From: Golbabi M
ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:
Shared Package
ML17312A761 List:
References
NUDOCS 9605210182
Download: ML17312A763 (11)
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Enclosure to ST-96%287 Page 1of9 Enclosure to ST-96-0287 Calculation Summary of Radiological Doses for Steam Generator Tube Rupture with Loss of Offsite Power and Stuck Open ADV Prepared By ehran lbabai NamelSignaturtJD ate VERIFICATIONSTATUS'OMPLETE The Safety-Related design information contained in this document has been verified to bc correct by means of Design Rcvicw using the Other Design Document Checklist of QPM 101.

Name < ~ ~~<<<>><KiSignaturc~ ~~ Date~~ii Independent Reviewer 9g0SZ<oasa 9 o PDR ADOCK 05000528 II PDR

Enclosure to ST-96-0287 Page 2 of9

Subject:

Calculation of the Radiological Doses for the Steam Generator Tube Rupture with LOSS OF OFFSITE POWER and Stuck Open ADV event at 2% Stretch Power

SUMMARY

The following discussion summarizes selected methods and assumptions used in calculation of the radiological releases for the Steam Generator Tube Rupture with LOSS OF OFFSITE POWER and Stuck open ADV at 2% stretch power. In summary:

1) The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> primary to secondary leakages and secondary steam releases remain the same as the original calculation. The primary to secondary leakage is a function of the RCS and steam generators pressures which are mostly controlled by the operator during the event and are not impacted by the 2% power increase or lower AFAS setpoint.

For the erst two hours, the secondary steam releases are driven by the stuck open ADV on the affected steam generator which removes much more heat than the core generates as decay heat. For the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, the total integrated decay heat generated by the stretch power core using the 1979 ANS decay heat curve is less than that assumed in the original analysis.

2) The additional radiological doses are due to the increased period of tube uncovery during the event. The assumed period of uncovery was increased &om 887 seconds in the original calculation to 1230 seconds (effectively assuming a tube uncovery period &om 460 seconds to 1690 seconds) for the stretch power analysis.
3) A copy of the chronology of the event was marked up and is attached. The only markup is the time that the level in the faulted steam generator rises above the top of the U-tubes. This value has been increased &om 1385 seconds (in the copy) to 1690 seconds. Even though the changes considered in the stretch power calculation would slightly impact the timing of the event, the increase in time of level recovery above the U-tubes is the only change needed to represent the conservative approach used for radiological dose calculation.

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Enclosure to ST-96-0287 Page 3 of 9 DISCUSSION The radiological doses for this event are primarily due to the releases &om the stuck open ADV on the affected steam generator. Most of the releases occur between the time that the affected steam generator ADV becomes stuck and the time that the level in the affected steam generator rises above the top of the tubes. During this period, the entire primary to secondary leakage is assumed to flash immediately and be released to atmosphere with a DF of 1.0. The original calculation of this event established this time to be &om 460 seconds to 1347 seconds, for a total of 887 seconds (14.8 minutes). The 2%

stretch power calculation increased this time period to 1230 seconds (20.5 minutes),

effectively assuming a tube uncovery period from 460 seconds to 1690 seconds. The additional 343 seconds are due to the following considerations:

1) 212 seconds resulting &om reduction of the AFW rate &om 750 GPM to 650 GPM.
2) 60 seconds to account for lower steam generator masses at initiation of the event and the time of AFAS initiation.
3) 71 seconds resulting &om the reduction of AFAS initiation analytical setpoint &om 25% to 21% of wide range span.

The 2% higher power does not affect the two hour or 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> steam releases and radiological doses. For the first two hours, the secondary steam releases are driven by the stuck open ADV on the affected steam generator which removes much more heat than the core generates as decay heat. For the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, the total integrated decay heat generated by the stretch power core using the 1979 ANS decay heat curve is less than that assumed in the original analysis.

Based on the above discussion, the two hour and 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> PIS and GIS radiological doses were calculated as follows:

2 Hour PIS From the original calculation, the radiological doses during the tube uncovery period (460 to 1347 seconds) were 185.2 REM. The total two hour radiological doses were 206.6 REM.

The two hour PIS doses for stretch power are calculated by ratioing the original dose release in proportion to the increased time of tube uncov'ery, as follows:

K20.5/14.8)

  • 185.2] + (206.6- 185.2) = 278 REM.

Enclosure to ST-96-0287 Page 4 of 9 The radiological doses were conservatively increased by another 5% to account for the expanded MSSV setpoint tolerances &om 1% to 3%. This brings the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> PIS dose to 292 REM (rounded up).

2 Hour GIS From the original calculation, the radiological doses between 1337 and 1690 seconds (the additional period of tube uncoveiy for stretch power) were recalculated, assuming that the entire primary to secondary leakage Qashes immediately and becomes airborne with a DF of 1.0, The additional doses during this period were calculated to be 30.6 REM. The additional doses during this period were added to the 40.4 REM, the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses &om the original calculation. This resulted in a two hour GIS dose of 71 REM.

The increased MSSV tolerances do not significantly impact the overall GIS doses since the GIS spiking is small during the MSSV opening this period.

8 Hour PIS and GIS The additional dose increases determined above for two hour PIS and GIS, resulting &om the increased tube uncovery period for stretch power were subsequently recalculated using the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to two hour dispersion factor ratio and then added to the respective 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> PIS and GIS doses of the original calculation.

t Enclosure to ST-96-0287 Page 5 of9 a valve Vruasru I'Oars DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 1 of 5)

Time Setpoint Success Path (sec) Event or Value or Comment 0.0 Tube rupture occurs 40 Third charging pump -0.75 Primary system started, feet'below integrity program level 40 Letdown control valve -0.75 Primary system throttled back to integrity minimum flow, feet below pxogram level 47 CPC hot leg saturation Reactivity trip signal generated control 47.15 Trip breakers open Reactivity control Turbine/generator trip Secondary system integrity 51 Loss of offsite power 52 LH main steam safety 1) 265 Secondary system valves open, psia integrity 52 RH main steam safety 1,265 Secondary system valves open, psia integrity 56 Maximum steam generator 1,330 pressures both steam generators, psia (1)

March 1990 15.6-40 Revision 2

PVNGS UPDATED FSAR I ~

DECREASE IN REACTOR Enclosure to ST-96-0287 COOLANT INVENTORY Page6of9 Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULIY STUCK OPEN ADV (Sh'eet 2 of 5)

Time Event Setpoint Success Path (sec) or Value or Comment 121 Steam generator water 25+ Secondary system level reaches auxiliary integrity feedwater actuation signal (AFAS) analysis setpoint .in unaffected generator, percent wide range level 122 AFAS generated 131 Steam generator water 25+ Primary system level reaches AFAS integrity analysis setpoint in the affected generator, percent wide range level 132 AFAS generated 167. 0 Auxiliary feedwater Secondary system initiated to unaffected integrity steam generator 177. 0 Auxiliary feedwater Secondary system initiated to affected integrity steam generator

  • The analysis used a setpoint of 21%. Even though this change would slightly impact the timing of the event, the only change needed to represent the conservative approach of the analysis is the time of level recovery above the u-tubes.

L (1)

Revision March 1990 15. 6-41 2

CREASE IN REACTOR COOLANT INVENTORY Enclosure to ST-96-0287 Page7of9 Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FUI LY STUCK OPEN ADV (Sheet 3 of 5)

Time Event Setpoint Success Path.

(sec) or Value or Comment 460 Operator initiates Reactor heat plant cooldown by removal opening one ADV on each SG ADV of the affected SG instantane-ously opens fully 484 Pressurizer empties 513 MSIS actuation 919 Secondary system secondary pressure. integrity psia 535 Automated isolation of 185 Secondary system AFW to affected SG, integrity hP SGs, psi 581 Pressurizer pressure 1, 578 Reactivity control reaches safety injection actuation signal (SIAS) analysis setpoint, psia 581 Safety injection actuation signal generated (1)

March 1990 15.6-42 Revision 2

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4 CREASE IN REACTOR COOLANT INVENTORY Enclosure to ST-96-02&7 Page 8 of 9 Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 4 of 5)

Time Event Setpoint Success Path (sec) or Value or Comment 581 Safety injection flow Reactivity control initiated 655 Operator overrides the AFW isolation signal and starts feeding the affected SG with AFW 775 Operator takes manual control of the AFW system, feeds affected SG with both AFW pumps 895 Operator shuts the ADV of the unaffected steam generator 1015 Operator initiates auxiliary spray to the pressurizer Level in the affected 71. 5 16 fg SG above the top of U-tubes, percent wide range (1)

March 1990 15.6-43 Revision 2

CREASE IN REACTOR COOLANT INVENTORY Enclosure to ST-96-0287 Page 9 of 9 Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 5 of 5)

Time Event Setpoint Success Path (sec) or Value or Comment 2040 Pressurizer level, 50 percent 2400 Operator controls HPSI 20 flow, backup pressur-izer heater output, and auxiliary spray flow to control RCS pressure and subcooling, F 28.800 Shutdown cooling entry 400/350 conditions are reached; RCS pressure, psia/

temp. F 28,800 Operator activates shutdown cooling system (1)

March 1990 15.6-44 Revision 2

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