ML20079Q659
| ML20079Q659 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 05/31/1983 |
| From: | Cornwell K, Hong F GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20079Q644 | List: |
| References | |
| 82NED105R1, DRF-E52-000012, DRF-E52-12, NEDO-22243-1, NUDOCS 8402010287 | |
| Download: ML20079Q659 (17) | |
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. DRF E52-00012 82NED105R1 f Class I May 1983 f
Revision 1 f..
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SAFETY EVALUATION OF MSIV LOW TURBINE INLET PRESSURE ISOLATION SETPOINT CHANGE FOR VERMONT YANKEE NVCLEAR POWER STATION Prepared by:
K. F. Cornwell F. S. Hong I
Approved: / 4h Approved: 4'//f[8 T.~Lf Gridley.(Manager '
~
? R.'J. 8/andon, Manager Fuel and Services Licensing Nuclear Services Engineering I
I I
NUCLEAR POWER SYSTEM 5 DIYiSION
- GENERAL ELECTRIC COMPANY 6
5AN JOSE, CALIFORNIA 95125 7
GENER AL $ ELECTRIC P
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f.ren 4LJii?Eb.i8N' - E.EYdNN8h$IbUk'A.sii@lwM 4 ' CAAuA. mij.,3l.;j s . ,- ; m NEDO-22243-1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT l
Please Read Carefully I i
The only undertakings of General Electric Company respecting information in this document are contained in the contract between Vermont Yankee Nuclear Power Corporation and General Electric Company (Purchase Order No.18905 dated July 14, 1982) and nothing contained in this document ,
shall be construed as changing the contract. The use of this informa- l tion by anyone other than Vermont Yankee Nuclear Power Corporation, or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
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N EDO-22243-1 CONTENTS Page l-1
- 1. INTRODUCTION 2-1
- 2.
SUMMARY
AND CONCLUSIONS 3 ^. VERMONT YANKEE MSIV LOW PRESSURE ISOLATION SETPOINT 3-1 CNANGE ANALYSIS 3.1 Description of MSIV Low Pressure Isolation Setpoint 3-1 Function 3.2 Methods of Analysis - Simulation of the Pressure Regulator 3-2 Failure (Open) Event 3-3 3.3 Results of Pressure Regulator Failure (Open) Calculation
- 4. VERMONT YANKEE FATIGUE ASSESSMENT FOR 750 PSIG 4-1 MSIV ISOLA ANALYTICAL LIMIT 4.1 Peak Stress of Pressure Regulator Failure Transient with 4-1 750 psig MSIV Isolation Analytical Limit 4-2 4.2 Fatigue Usage Under Pressure Regulator Failure Transient 5-1
- 5. OTHER CONSIDERATIONS FOR LOWERING MSIV ISOLATION 5-1 SETPOINT 5.1 Radiological Consideration 5-1 5.2 MCPR and Future Fuel Cycle Considerations 5-2 5.3 Impact on Other Systems 6-1
- 6. REFERENCES
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..rw.w - u ~ ~ s. . .mm -~ m :z..a.w _.. A.u ~ aaa u a m "-=+ n NEDO-22243-1 The analysis performed simulates the Pressure Regulator Failure (opan) set at the event described above, with the MSlv low pressure isolation limit proposed 750 psig limit. The GE thennal-hydraulic and nuclear kinetics coupled transient code REDY 2 is used to evaluate the dynamic system response to the Pressure Regulator Failure (open) event previously discussed. The calculations are performed with the following basic assumptions and initial conditions.
- 1. The initial reactor power is at 1664 MWt, corresponding to 105% of rated steam flow.
- 2. Initial dome pressure is 1036 psia.
- 3. Initial core flow is at 100% of rated core flow.
- 4. End-of-cycle (EOC) scram, void and doppler reactivity curves are used, based on Cycle 9 fuel loading conditions.
- 5. The MSlv closure time is assumed to be 5 seconds.
- 6. The turbine bypass valves with 110% full capacity arc open for faster depressurization, providing conservative results.
- 7. The pressure regulator upper limit is set at 115% of steam flow demand.
3.3 RESULTS OF PRESSURE REGULATOR FAILURE (OPEN) CALCULATION The results of the Pressure Regulator Failure (open) transient calcula-tion are presented in Figure 3-1. The figure includes the time history of vessel steam flow, vessel dcme pressure response and the corresponding satu-rated steam temperature, initially, the vessel steam flow increases rapidly as the turbine control valves open'up because of the pressure regulator failure in the open direction. With the high steam outflow, the vessel pressure decreases which results in vessel water level swell as the bulk fluid void 3-3 E- ' Csr*"3 9 M )b C.Wf.-E,.#f.pyy@eiggt;{.{. )]. f: ;% M .m:= h: k$ T* 5* .b hin$d.hi% b i bbkhb$. NEDO-22243-1 At 4.8 seconds, the water level reaches the high water volume is increased. Level 8 setpoint. 'This initiates MTSVC and feedwater pump trip, followed by a reactor scram on MTSV position. At this time the turbine bypase valves are fully opened, resulting in further depressurization of the vessel until the 750 psig turbine inlet pressure limit is reached which initiates MSIV closure. Following isolation, the system depressurization is terminated and the vessel The reactor, [ water level begins to fall as the system pressure increases. being scransned and isolated can now be brought to a controlled shutdown. Initially, the vessel done has a saturated steam temperature corresponding to 549'F (Figure 3-1). The done saturation temperature then drops, as a result of system depressurization, to a minimum value of 510.6*F at approximately 17.5 seconds. This maximum change in the steam dome saturation temperature provides the basis for the thermal stress evaluation, which is discussed in the next section. 6 9 3-4 , +1 k u I 1 # A C# m -L - L 1 A E g y NEDO-22243-1 l Misys CLOSE ON HWL.8, RE ACTOR SCRAMS i SYPASS VALVES FULLY OPEN g 100 - E MSIVs BEGIN TO CLOSE g N 3 .0 - l g m 1 1 E- l 5 MSIVs FULLY CLOSED ' I I ' ) 0 I 30 30 40 50 40 70 O 10 1 1.2 s I n 1.1 - 1 l g 1.0 - 3 - g 0. - 0.s . .a l Fw% = 790 peg g 0.7 - SA '8 8 8 8 8 8 0 to 20 30 40 to 00 70 ( 76 nim *#88 F 000 I i w ~ E 530 - e iW s20 - 0 a
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500 at 0 10 20 30 4J SO 90 70 TIME (sect Figure 3-1. Press.are Regulator Failure (Open) Transient Vermont Yankee 3-5/3-6
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4.1 PEAK STRESS OF PRESSURE REGULATOR FAILURE TRANSIENT WITH 750 PSIG MSIV ISOLATION ANALYTICAL LIMIT Lowering the MSIV low pressure isolation limit to 750 psig can allow the reactor vessel to depressurize further, imposing a larger vessel temperature gradient during the transient. The impact of this temperature change on the vessel component fatigue life can be evaluated based on conservative thermal stress calculations. The most severe thermal stress experienced by the component material, which is bounded is undergoing a temperature change within a finite amount of time, by the peak skin stress experienced by the material under a sudden temperature change of the same amount. This skin surface peak stress resulting from the instantaneous temperature change can be expressed as: Ac = ATaE (1-v) where An = peak stress induced at surface (psi) AT = temperature change from thermal shock (*F) a = coefficient of thermal expansion (in./in./*F) l E = modulus of elasticity (psi) v = Poisson's ratio 4-1 1 f . ~- -- m _. n g m . ~ J &a wW% 4 fwd 5ES ~ :5@'$b b %55L Uik br4.%L%-U & $ NEDO-22243-1 Based on the peak temperature change of 38'F, which the vessel component has experienced under the Pressure Regulator Failure (open) transient (Figure 3-1), the bounding thermal stress is calculated to be 750 psig MSIVC Analytical limit = 13127 psi 21s is a very conservative and bounding thermal stress value for the Pressure Regulator Failure transient. Tliis peak type stress is classified in the ASME Pressure Vessel Code (Section III) as the stress for fatigue life evaluation, where an allowable stress limit is imposed for f atigue cycling considerations. 4.2 FATIGUE USAGE UNDER PRESSURE REGULATOR FAILURE TRANSIENT Fatigue cycling is classically described by a " usage factor," which is defined as I ng/Ngusing a linear damage relationship; assuming that, if Ng cycles would produce failure'at a stress level Sg , then ng cycles at the same stress level would use up the fraction n /N of the total life. Failure g g is postulated to occur when the cumulative usage factor becomes 1.0. he original Vermont Yankee Stress Report has treated reactor vessel trensients in a hiahly conservative manner by " enveloping" the various life-time cycling events with a " worst case" transient where the total number of major cycles are assumed to have occurred under this transient. He fatigue usage change of the Pressure Regulator Failure Transient under the 750 psig MSIV isolation limit was evaluated based on the results of this orignal stress report. To assure that the transient with the 750 psig limit condition is treated conservatively, the entire 13127 psi stress range was included for the postulated fatigue effect calculation. The 13127 psi was accounted for by using scaling techniques or direct addition to modify the maximum stress ranges previously calculated in the stress report at each vessel region. l 4-2
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-...,.m, y W .k ** h w ~--r~.~.'.w n -m p z.-x.m ,m n NEDO-22243-1 The reactor vessel component with the largest fatigue usage as shown in The dominating stress in the ,, the previous report was the feedwater nozzle. feedwater nozzle is due to the injection of 100*F cold inlet water during This peak startup when the reactor is at an operating temperature of 546*F. thermal stress was combined with steady-state and pipe stresses, and assumed The to occur in all of the 1500 postulated lifetime major design cycles. resulting thermal stress has a far greater stress range than that found during a Pressure Regulator Failure event. Also, the Pressure Regulator Failure transient is limited to only a few cycles during the lifetime of the reactor (eight times in 40 years). Consequently, based on the above stress results the and cycles of occurrence of the Pressure Regulator Failure transient, additional usage factor introduced to the feedwater nozzle by the adoption of 2 the 750 psig MSIV isolation limit is insignificant. The A similar result is obtained for the other vessel components. l remaining regions of the reactor vessel, which all have similar fatigue usage f actors, were examined in a similar fashion. It was found that their lifetime fatigue usages would not increase appreciably and all would remain at less than unity. that lowering the It is concluded, based on the preceding assessment, Vermont Yankee MSIV isolation analytical limit to 750 psig will not have a significant effect on the reactor vessel's lifetime fatigue usage. l I 4-3/4-4 7 3M W 44 M NEDO-22243-1 o T
- 5. OTHER CONSIDERATIONS r0R LOWERING MSIV ISOLATION S 5.1 RADIOLOGICAL CONSIDERATION bine The impact on the radiological release due The to reducing radiological the low tur inlet pressure isolaticn setpoint has been considered. t inment will release caused by the design basis steamline break l i n isoutside assumed the con a not be af fected, because for this large a break, MSIV iso pressure.
bine inlet at o to occur as a result of high steamline flow, not tur h high steam Steamline breaks of a size small enough not to be detected byl t e flow signal are isolated either by temperature sensors in the steam tunneThe or area radiation monitors in the turbine building. d i g the low turbine that the radiological release will not be af fected by re uc n inlet pressure isolation setpoint to 750 psig. 5.2 MCPR AND FUTURE FUEL CYCLE CONSIDERATIONS its current The reduction of the MSIV low pressure isolation l thermalsetpoint margins.from This value to 750 psig will have no impact upon MCPR or fue Regulator Failure can be demonstrated by comparing the scenario of two Pressure 1V isolation f (open) transients, with one utilizing the current low pressure MSIn both case and the other with the proposed value of 750 psig. decreasing vessel pres-control valves open, increasing vessel steam flow and f sure. The vessel depressurization causes the water level to swell because o a turbine When the water level reachas the high water level setpoint, voiding. t p valve position. trip occurs which initiates a reactor scram on turbine s o depressurize As a result of the large bypass capacity, the vessel continues is reached. to By then, until the current low pressure MSIV isolation setpointflux of less than 1%. the reactor is essentially shut down with a neutron hd the two I Up until the time that the current MSIV isolation h e setpoint in MCPR is is reac e , transient responses are identical and, therefore, no c ang Af ter this time, the dif ferences in the responses will have no possible. impact on MCPR because of the very low reactor power. l tion Therefore, it is concluded that reducing the MSIV low pressure iso a setpoint has no impact upon MCPR or fuel thermal limits. y 5-1 % MOC1 6, ~ x . a_-J:'.C -4)iMN:TrL'&Na s[yia$$~ 'e :an . . _.N;kNyjdU.$G;M&. , NEDO-22243-1
- 'Ihe present analysis is applicable to future fuel cycles since the MCPR and fuel thermal limits, are not a concern, and the reactor response to this event is not sensitive to'the cycle dependent core nuclear characteristics.
5.3 IMPACT ON OTHER SYSTEMS The isolation signal from the steam turbine inlet pressure will initiate isolation of the following valves:
- 2. Main steam drain valve (MSDV)
- 3. Reactor water sample valve (RWSV)
The isolation of the MSIV has been discussed earlier and is the main objective of this safety evaluation. The isolation of the RWSV at a lower steam turbine inlet pressure is acceptable because the RWSVs and MSDVs can
- be opened from the contrel room. The isolation setpoint of MSDV and RWSV will not affect any of the design bases. Therefore, it is coacluded that the 750 psig isolation setpoint is acceptable.
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- 6. REFERENCES I'
- l. " Vermont Yankee Stress Report, Reactor Vessel." CB&I Contract 9-6201, GE PO #205-55565-I ' Reactor, Rev. 3, March 20, 1970.
- 2. " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor " General Electric Company February 1983 (NEDO-10802) .
- 3. Formulas for Stress and Strain, Roark and Young, 5th Edition,1975 McGraw-Hill.
- 4. " Data for National Reliability Evaluation Program" (NREP). A. J. Oswald et al. , EG&G Report, .Nne,1982 (EG&G-EA-5887) .
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