ML19331A042
| ML19331A042 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/28/1978 |
| From: | Early P POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| References | |
| JNRC-78-47, NUDOCS 7810040122 | |
| Download: ML19331A042 (18) | |
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4 POWER AUTHORITY OF THE STATE OF NEW YORK
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Director of Nuclear Reactor Regulation 39
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71 Attention:
Mr. Thomas A. Ippolito j
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Operating Reactors Branch No. 3 Division of Operating Reactors ggi m
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Subject:
. Tames A. FitzPatrick Nuclear Power Plant i
y Mark I Containment Short Term Program
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Multiple Sub t Safety / Relief Valve Actuations y, 'J3 Docket No/ 50-333
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Dear Sir:
In accordance with the U. S. Nuclear Regulatory Commission's letter of March 20,1978-to the Povier Authority, and as committed to in the Authority's letter dated August 18,1978 to the Commission, a plant-unique assessment of the effect of multiple subsequent actua-tions of safety / relief valves following an isolation transient event has
..been performed.
Enclosed with this letter is Attachment A, " Multiple Safety / Relief Valve Actuation Evaluation for the Tames A. FitzPatrick Nuclear Power Plant", summarizing the results of our analysis and describingthe meth-ods and assumptions used. The analysis supports upcoming operation of the FitzPatrick Plant during Cycle 3 by assuring that the structural cap-ability of the torus satisfies acceptance criteria established by the Short Term Mark I Containment Program.
The most severe Boiling Water Reactor pressurization event with respect to multiple subsequent safety / relief valve actuations occurs as a result of the closure of the main steamline isolation valves. Best estimate plant parameter values were used whenever sufficient data was available to substantiate their usage. In those cases in which the data
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base was not adequate, the specified conservative safety analysis parameters were assumed. The calculational models used for the analysis are current General Electric Boiling Water Reactor trans-ient safety analysis models, considered by the Commission to be acceptable.
In the analysis the safety / relief valves are divided into three setpoint groups. Setpoints analyzed in the analysis are 1090 psig (2 valves),1105 psig (2 valves), and 1140 psig (7 valves). Target Rock valves with new two-stage topworks are used at the lowest setpressure while existing Target Rock valves with three-stage t
topworks are used at the other cetpressures. The two new two-stage valves will be installed during the refueling outage currently underway, and an of the safety / relief valves will be reset to the new setpoints.
Valve setpoint tolerance is +1%, so the setpressures win be distributed within the tolerance band. The expected distribution-of actual SRV setpressures around specified setpressures can be calculated from the expected value of the ordered statistics for sman samples selected from a parent normal population of actual variations from nameplate values.
The transient analysis completed by General Electric Company indicates that two (2) safety / relief valves experience simultaneous subsequent reopening following closure of all main steam line isola-tion valves at full power operation based on the new valve groupings.
All eleven (11) safety / relief valves open for the initial pressure increase, however, the two lowest set valves (two-stage Target Rock valves at 1090 psig) remain open when the other valves close.
The safety / relief valves predicted to reopen on the subsequent actuation fonowing the second pressure rise are the two valves with the 1105 psig setpressure.
The effect of the two subsequent actuations has been analyzed by Teledyne Engineering Services. The valves experiencing sub-sequent actuation (1105 psig setpressure) have discharge points l
0 located approximately 180 apart in the torus. This structural analysis verifies that the highest strength ratio for the torus sheu and torus support system under this condition is 0. 29.
This calcu-lated strength ratio is well below the Short Term Program acceptance criterion of 0.5.
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,3, An analysis was also performed confirming that, even with subse-quent multiple actuations involving the four (4) lowest set safety / relief valves (2 valves at 1090 psig, 2 valves at 1105 psig, and the four dis-charge points located approximately 90 apart in the torus), the calcu-lated maximum strength ratio for the torus shell and torus support system is 0.34. Again, this is less than the acceptance criterion of a limiting strength ratio of 0. 5.
Since startup of the James A. Fitzpatrick Plant for Cycle 3 is scheduled to take place on November 13, 19 78, timely review by the Commission of our evaluation would be greatly appreciated.
Very truly yours,
/ J.~!
/ A Pa J. Ear y Ase tant C ef Engineer-Prdjects Att.
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- ACTUATION EVALUATION i
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FOR JAMES A FITZPATRICK NUCLEAR POWER PLANT t
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TABLE OF CONTENTS Page 1.
Introduction 1
2.
Analysis Basis 1
L 3.
Summary of Results 2
4.
Conclusions 3
i L.
j-TABLES I
i Table 1:
Important Plant Parameters 4
Table 2:
Major Analysis Assumptions 5
. Table 3:
Summary of Results 6
FIGURES i
Figure 1: S/R Valve Statistical Groups 7
Figures 2A to 2E: 4.5 Second MSIV Closure W/ATWS-DMT. Statistica1 S/RV Set Pressures,12296 Nam =plata Capasity
+
A, Pressure vs. Time 8
B.
Water Leval vs. Time 9
C.
Flow vs. Time 10 D.
Flow vs. Time 11 E.
Flow vs. Time 12 1
Figure 3: Relief Valve Cycling 13 Q --
1.
Introdhetion L
f Multiple Safety / Relief Valve (SRV) operation following pressur-ization transients may impose significant loads on the Mark I containmants. Safety /ralief valve discharge load tests ware made at the Monticello Nuclear Generating Plant Unit I during 1976 and 1977. After reviewing the results of the Monticello SRV discharge tests, the NRC staff required that each utility submit a plant unique analysis of the "most probable" effect of multiple SRV actuations following an isolation transient avant, as requasted in the USNRC's March 20,19781atter to PASNY.
An analysis was performed for James A. FitzPatrick Nuclaar Power Plant (JAFNPP) based on Cycle 3 SRV groupings of 2 SRV's at 1090 psig setpressure and 9 SRV's at 1115 psig set-pressure. The enalysis, transmittad'to the Commission in an August 18,19781stter, predictad 11 SRV's with multiple actua-tions. Subsequently, as committed to in the August 18, 1978 letter, General Electric Co,has performed a raanalysis of i
multiple SRV actuations using SRV groupings of 2 at 1090 psig, 2 at 1105 psig and 7 at 1140 psig. Target Rock SRV's with 2-stage topworks are used at the lowast satpressure, while Target Rock SRV's with 3-stage topworks are used at tha other
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setpressures. The salaction of the SRV satpressuras was based on a study performed by Genaral Electric Company.
2.
Analysis Basis The most severe Boiling Water Reactor (BWR) pressurization event with respect to multiple subsequent SRV actuations occurs as a result of the closure of the Main Steam Line Isolation Valvas (MS1V). The basis for the plant parameters usad in this analysis is given in Table 1. The current General Electric BWR transient safety analysis models, which as stated in the Commission's March 20,1978 letter are acceptable, consider the maximum SRV capacity to be the ASME nameplate setpressure. For safety analysis, the SRV's are divided into setpoint groups with the behavior of all valves in a group assumed to be idantical. The safety analysis assumes that the SRV's open at their nameplata setprassure plus a 196 tolerance as indicated in the plant technical specifications.
For the purposes of the analysis describad harein, the NRC requira-ment of a "most probable" analysis has been met by using best estimate plant parameter values whenever sufficiant data was avail-able to substantiate their usage. In those casas in which the data base was not adaquate, the specified conservativa safety analysis pacameters ware assumad. -
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The parametars with sufficient data to support a "btst ostimato" value are i,
1.
3-Stage SRV Setpressure Distribution 2.
2-Stage SRV Setpressura Distribution 3,
2-Stage SRV Response Timas 4.
2-Stage SRV Blowdown (oponing pressure minus closing prassure) 5.
MSIV Closure Time 6.
Recirculation Flow Coastdown Time The SRV satpoint tolerance is i %, so the satpressuras will ba l
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distributed within the tolerance band. The expected distribution of actual SRV setpressures around nameplate satpressures can be calculated from the expacted value of the ordered statistics for small samples selectad from a parent normal population of actual variations from nameplate values. The distribution of SRV setpressures used in these analysas is shown in Figure 1.
The dalay and responsa times for the 2-Staga valve were obtainad from steam test data taken on pre-production and production valva samples. Thk values used in this analysis are shown in Table 2.
Tha blowdown of the 2-Stage valva as a function of back prassure was also determined from steam test data. Tha data shows the blowdown is greater than 45 psi for the calculated JAFNPP SRV discharge line back pressuras. A conservativa valua of 3% (valva specification value) was used in this analysis.
The test data from the periodic MSIV closura tasts at JAFNPP shows the average MSIV closure time is 4.5 saconds. This value was used in the analysis. Tha recirculation flow coastdown tima constant was determined by using actual.TAFNPP MG-set and pump-motor inertia values in the transient analysis model and determining the resulting flow coastdown time constant from a simulated drive motor trip transient. The resulting value is given in Table 2.-
3.
Summary of Results Table 3 summarizes the predicted SRV response to the isolation event. The results for the first pressure rise are assantially th=
same for all cases. The paak pressure and time to reach peak pressure,are nearly the same for all cases, and all eleven valvas open for each case. For all cases the two lowest sat valves romain open during the first pressure transiant and the three lowast sat valves remahl opan for Case 2.
Figure 3 shows the r=11af valva cycling which occurs for each casa.
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Case 1, the basa casa, r:prrcnts tha "best estimata" of SRV parformance. The results of this case are plotted in Figures 2A through 2E. The remaining cases are included to show the effect of varying selected parameters.
Case 2 shows the effect of reduced SRV capacity on the predic-tions. The results are less severe than th9 base case b*causa 3 valves remain open during the first pressure decrease, rasult-ing in a slower rate on the sacond pressure risa.
Cases 3 and 4 show the.effect of assuming the valves opan at their nameplate setpressure. The results for both cases are similar to the cases with statistical satpressures excapt AT is shorter.
2,3 Case 5 shows the effect without the Anticipated Transient Without Scram Drive Motor Trip. This case has a much lowar paak pres-sure on the second pressure rise, and AT is shortar than the 2,3 base case.
4.
Conclusions a.
For the SRV groupings and the analysis basis usad, the maximum predicted number of safety r=11ef valves with multiple actuation is two.
b.
A conservative 30 psi blowdown was used for the 2-Stage valve instead of the "best estimate" 45 psi.
Tharefore, the analysis results are indepandent of SRV discharge line backpressure so the installed location of the 2-Stage valves will not affect the analysis results, c.
The predicted time between the first and second actuations, AT 2, is short (one second or less),
d.
The SRV's predicted to open on the second actuation ara those with the 1105 psig setpressura. Subsequent to the second actuation, one of the 2-Staga valves will cycle on and off until the pressure transient is terminatad.
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L TABLE 1 IMPORTANT PLANT PARAMETERS g
r JAMES A. FITZPATRICK NUCLEAR POWER PLANT i
SAFETY / RELIEF VALVE MULTIPLE ACTUATION STUDY-1.
Reactor Operating Conditions:
Licensing Basis Values in Table 5-6 of NEDE-24011-P-A and NEDO-24011-2, " Generic Raload Fuel Application",
March 1978.
2.
SRV Setpoints and Capacity; Set ASME Capacity Pressure
@ 103% Setpressttre No.,of Valms psig Ib/hr per valva 2
1090 818,000*
2 1105 829,000 7
1140 S55,000 l
6 3.
Main Steam Isolation Valve Closure:
Stroke Time:
- 4. 5 sec.
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T M D L t. 4 MAJOR ANALYSIS ASSUMPTIONS
-1.
Isolation Event:
Closure of all MSIV's, (Position Scram) 2.
Initial Core Power:
104%
3.
Core Power Decay:
3 full powar saconds followed by l
ANS + 20% decay heat curve (Infinite Exposace) 4.
ATWS MG-Set Drive Motor Trip-High Vessel Pressure Vessel Pressure Trip Setpoint:
1135 psig Recire System Inertial Tim
- Constant:
- 8. 5 seconds Mechanization Delay:
0.3 seconds 5.
SRV Setpoint Tolerance (Satpoints based on staam test dsta)
- a. Target Rock, 3-Stage Standard Dev: 5.0 psi Mean: Approx. O psi
- b. Target Rock, 2-Stage Standard Dav: 3.0 psi Mean:
- 1. 7 psi 6.
SRV Capacity:
Maximum:
122.5% ASME Rated 7.
SRV Response Times:
- a. Target Rock, 3-Stage Delay Time: 0.4 seconds Stroke Time: 0.15 seconds
- b. Target Rock, 2-Stage Delay Time:.3 seconds Stroke Time:.1 seconds 8.
Reactor Vessel Blowdown Response:
- a. Target Rock, 3-Stage Closure Setpoint: Opening Setpoint minus 25 psi Dalay Time: 1.5 seconds Stroke Tim?: 0.3 seconds
- b. Target Rock, 2-Stage Closure Satpoint: Opening dotpoint minas 30 psi Delay Tim?: 0.3 s>conds Stroke Tima: 0.1 s3 condo.
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TABLE 3 SuramRT OF RESULTS FOR JAfi[5 A. FITZPATRICK NUCLEAR POWP. Plasti SAFETT/ RELIEF VALVE 111LTIPLE ACTilATION STbTY First $RY Second SRY g,gg
$gy Atuadon Actuatim To Nest Cap.
SRV 1
Peak Time To Peak Time To Menber A T' '
OT Valve On 2*
Setpressure A5ME ATW5 Press.
Peak Press.
Press.
r+4k Press.
Of 2nd l'op' r.ase SNute Distribution Pated Dili (P5tG)
(Secs)
(PSIG) f, Secs)
Valves (Secs)
(Sets)
(PSI)
I 7681T Statistical 122 Yes 1162 3.6 1129 10.9 2
1.0 11.1 4
8 4
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1061T Statistical lor)
Yes 1165 3.6 1120 13.6 1
.8 12.3 13 3
1065T Nominal 100 Yes 1164 3.6 1125 12.9 2
.8 5.3 15 4
1069T Nominal 122 Yes 1160 1.5 1131 11.2 2
.5 3.9 9
5 1077T Statistical 122 No 1161 3.6 1112 9.9 2
I.0 4.3 20 I
6T.2 Time between last valve reciosure on first openin9 and reopening os secoms actuation.
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