ML20095L208
| ML20095L208 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/06/1992 |
| From: | Abdelghany J, Distefano J, Salvo J YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20095L191 | List: |
| References | |
| YAEC-1847, NUDOCS 9205060276 | |
| Download: ML20095L208 (32) | |
Text
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, YANKEE ATOMIC ELECTRIC: COMPANY
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, n Seabrook Station Code Safety Valve Setpoint Tolerance Relaxation February 26. 1992 Major Contributors: H. D. Fulcher D. A Prelewicz R19\15
Preparea By: /, 3 - [- N fimal M. Abdelgh6ny # [ (Date)
Prepared By: e w M' , 3 - d - 9 2.,
onn DiStefafio { ( D.i t e )
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Prepared By. h3.-
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- b'9I astpnt-[.Saivo (Date)
Reviewed Ey: YM, *d ML 0 b/ 0 '
Alan E. Ladieu (Date) i l
Reviewed By: [ tu<3# 3 4 [$ 2 Robert C. Harvey /(D(te)
Reviewed By: c > h[ 3 /( [9 L Cha rles D. Thord s (pate)
ADproved By: / O $ f2 Ro ert E. White # ~ /
(Dne)
Yankee At0mic Electric Company Nuclear Services Division 580 Main Street Bolton, Massachusetts 01740 R19\15 ii.
DISCLAIMER OF RESPONSIBillTY 4
This document was prepared by Yankee Atomic Electric Company (" Yankee").
The use of information contained in this document by anyone other than Yankee, or the Organization for which this document was prepared, is not authorized and with respect to any unauthorized use, neither Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility or liability or makes any warranty or representation concerning the contents of this document or its accuracy or completeness, e
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TABLE OF CONTENTS Pace DISCLAIMER OF RESPONSIBILITY .
. . . . . . . . . . . . . . . . . iii TABLE OF CONTENTS . . . . ...... . . . . . . . . . . . . . . . iv LIST OF FIGURES . . . . . ...... . . . . . . . . . . . . . . . v
1.0 INTRODUCTION
. . . . . ... .... . . . . . . . . . . . . . . . 1 2.0 iECHNICAL APPROACH , . ....... . . . . . . . . . . . . . . . 2 3.0 DETAILED EVALUATION . . ...... . . . . . . . . . . . . . . . . 3 3.1 Transient Analysis ...... . . . . . . . . . . . . . . . 3 3.2 LOCA Analysis . . . . . . . . . . . . . . . . . . . . . . . 20 3.3 SGTR Mass Release . . . . . . . . . . . . . . . . . . . . . . 21
. 3.4 ASME Code Review ...... . . . . . . . . . . . . . . . . 22 4.0
SUMMARY
AND CONCLUSIONS . ..... . . . . . . . . . . . . . . . . 24
-iv-
LIST OF FIGURES Noiber _T i t l e Paae 3-1 RETRAN Noding Diagram 6 32 Turbine Trip Without Pressurizer Control, Minimum Reactivity feedback Pressurizer Pressure Versus Time UFSAR Benchmark 9
-3 Turbine Trip Without Pressurizer Control, Minimum Reactivity Feedback Normalized Power Versus Time UFSAR Benchmark 10 34 Turbine Trip Without Pressurizer Control, Minimum Reactivity feedback Pressurizer Liquid Volume UFSAR Benchmark 11 3-5 Turbine Trip Without Pressurizer Control, Minimum Reactivity feedback Core inlet Temperature Versus Time UFSAR Benchmark 12 36 Turbine Trip Without Pressurizer Control, Minimum Reactivity feedback Pressurizer Pressure - High Safety Valve Setpoint Analysis 15
.y.
l l.0 INTRODUCTION The code safety valves on the Seabrook station pressurizer and main steam lines currently are tested and verified to have their setpoints within a tolerance of ill of nominal. This tolerance is more limiting than required by the ASME Code and imposes unnecessary valve testing requirements. By increasing the setpoint tolerance, the amount of testing and also the man-rem exposure can be reduced without impacting plant safety.
, This evaluation supports a change in Seabrook Technical Specifications (3.4.2.1 and 3.4.2.2 Safety Valves and 3.7.1.1 Turbine Cycle Safety Valves) to increase the setpoint tolerance from ilt to 13%. With the increased tolerance the upper and lowc? limits on safety valve setpoints will be 2575 psia and 2425 psia for the pressurizer. Each main steam line has five valves with
. differant setpoints. The valves with the lowest nominal setpoint will have an opening pressure between 1164 psia and 1236 psia.
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1
2.0 TrrH41 rat APPROArH The evaluation which follows cemonstrates that the proposed changes in Technical Specifications represcnts no significant hazards consideration.
Each of the transients considered in the Upuated Final Safety Analysis Report (UFSAR. Ref. 1) is evaluated to determine the effects of increased safety valve setpoint tolerance, in most cases it is apparent that a change in setpoint tolerance cannot have an effect on the existing calculation. as for example when the opening setpoint is not challenged. -
UFSAR DNBR evaluations take credit for operation of the pressurizer PORVs. which have a setpoint pressure of 2400 psia. Since this is lower than the proposed lower limit of 2425 psia on pressurizer safety valve settoint, revising the safety valve setpoint tolerance does not affect DNBR. Increasing pressure yields less limiting values of DNBR. Therefore, all further evaluations address peak pressure response.
Each event was then evaluated to determine the effects of revised safety valve setpoint tolerance on peak pressure. The most detailed evaluation was ___
performed for the limiting pressurization transient. the turbine trip. This event was simulated using the RETRAN02 MODS (Ref. 2) computer code. The Safety Evaluation Report for this code (Ref. 3) has recently been issued by the NRC. This code is acceptable for the type of system transient analyzed in this report.
Comparisons were performed with the licensing basis analysis of record -
to demonstrate that the RETRAk model provides comparable results. The ,
limiting pe;k pressure case with minimum reactivity feedback and without pressure control was simulated using RElRAN. Agreement with UFSAR results was very close with the RETRAN analysis results slightly overpredicting the peak pressure compared to the UFSAR. Evaluations with revised setpoints were then performed with the RETRAN model to demonstrate that the peak pressure remains well below the Condition 11 limit of 110% of design pressure, or 2750 psia for the primary system and 1320 psia for the secondary system. --
M 2-
3.0 DETAltfD [ VALUATION 3.1 Transient Analysis A lower safety valve setpuint will increase steam flow from the system and hence decrease the pressurization, This will decrease the severity of a pressurization event. The important consideration at the lower value of the setpoint is whetner the decreased setpnint will result in increased challenges to the safety valve, or interfere with other pressure dependent functions, such as the high pressure reactor trip.
The proposed lower limit of the pressurizer safety valve setpoint (-31 tolerance) is 2425 psia. This is sufficient 1/ above the PORV setpoint of 2400 psia to prevent unnecessary challenges to the safety valve. The high pressure reactor trip setpoint is also set at 2400 psia, so the trip will occur prior to opening of the safety valves.
Group 1 main steam safety valves (MSSVs), which have the lowest opening setpoint, have a proposed lower limit setpoint ( 31 tolerance) of 1164 psia.
This is sufficiently above the atmospheric steam dump valve opening setpoint of 1140 psia to prevent challenges to the safety valves. It is also well above the no load steam pressure of 1106 psia.
In summar c , no increase is expected in the frequency of challenges to either the pressurizer or main steam safety valves due to the increased setpoint tolerance. A margin on the order of 25 psi remains between the PORV setpoints and the pressurizer saiaty valve (PSV) setpoints. Similarly, the high pressure reactor trip will also occur at 25 psi below the minimum PSV lift point.
Mass and energy release for the containment performance analyses in UFSAR Chapter 6 will not be affected by the change in setpoint tolerance.
Each of the events analyzed in the Seabrook UFSAR Chapter 15 (Ref. 1) was evaluated to determine the effects of increased safety valve setpoint tolerance. Following is a summary of the results of the evaluations:
3-
3.1.1 Introase in Heat Removal by the Secondary System knduction in feedwator Tenperature This event is bounded by the Excessive increase in Steam Flow event discussed below. For this event RCS pressure and MSS pressure do not reach the reduced PSV and MSSV setpoints. Thus the results are unaffected by increasing the setpoint tolerance to 31.
Increase in Feedwater Flow The peak pressurizer pressure does not reach the roduced setpoint pressure of 2425 psia (UFSAR Figure 15.1-1. Sheet 3) and therefore, the relaxed tolerance does not affect the existing analysis for this event.
Secondary side pressure decreases for this event 50 the relaxed tolerance setpoint cannot affect the existing analysis, fwressive increa'.e in Steam Finw As shown in UFSAR Figures 15,1-2 through 15.1 5 pressure decreases for this event so the relaxation in pressurizer safety valve setpoint tolerance does not affect the existing analysis.
Inadvertent Opening of a Steam Generatnr Relief or Safetv Valve As shown in UFSAR Figure 15.1-7 pressure decreases for this event and therefore the relaxed safety valve setpoint tolerance does not affect the existing analysis for this case.
Steam line Brea6 As shown in UFSAR Figures 15.1-10 and 15.1 11 the pressure does not rise above the initial system pressure. Therefore, the relaxed safety valve setpoint tolerance does not affect the existing analysis for this case, 3.1.2 Decrease in Heat Removal by the Secondarv System Steam Pre (sure Renulator Malfunction The Seabrook plant does not have steam pressure regulators, therefore this event need not be considered.
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t 0 <. 5 nf Ixternal l05d The pressurization during this event is bounded by the turbine trip which is evaluated in the next section.
Tortsine Trip This is the limiting RCS pressurization event. To demonstrate that the increased tolerance on safety valve opening setpoints does not compromise safety limits, a detailed evaluation was performed including simulation of the event with revised setpoints.
Since both the pressurizer and steam generator pressures increase for this event the increased opening setpoints are considered in the reanalysis.
The RETRAN02 MOD 5 (Ref. 2) computer code was used to simulate the system response for this limiting event. Figure 31 is a nodal diagram of the RETRAN model for the Seabrook plant. The model is a 1 x 3 representation, with one loop to which the pressurizer is connected, modeled as a single : cop. and the remairing three loops combined as a single loop. A total of 53 nodes and 71 junctions are used to model the plant.
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l Technical Specification opening setpoints were used for both the , ,
pressurizer and main steam safety valves. Accumulation of 31 was assumed, i.e.. valves are fully open at a pressure 31 above the openin" setpoint.
Blowdonn pressure was taken as 75 psi below the opening setpoints for the main steam safety valves based on manufacturer's tett data summarized in Reference 4 Pressurizer safety valve blowdown pressure was assumed as 96% of the lift setpoint, which is equivalent to 100 psi below the setpoint. Since the relief lines for the Seabrook plant do not have loop seals, no delay is required to account for loop seul clearing.
RETRAN Benchmark t
To demonstrate that the RETRAN model is suitable for this analysis, the model was first benchma;ked against the calculated results for the event presented in the UFSAR. Bounding initial conditions and a conservative control system operability are assumed for the UFSAR analysis, The direction of the uncertainty included on each parameter (high or low) is consistent with th> conservative UFSAR analysis. The following initial conditions were nsumed:
Parameter Valua Comments Core Power Level 3479 MWt (102%) rated + uncertainty Average Temperature 594.3'F nominal + uncertainty Pressurizer Pressure 2220 psia nominal - uncertainty RCS Loop Flow 97,700 gpm nominal Steam Flow 15.14 Mib/hr from heat balance Secondary Pressure 1000 psia nominal Ftedwater Temperature 440*F nominal Pressurizer Liq Vol 1110 ft 3 nominal + uncertainty l
The point reactor kinetics model was used with Beginning of Cycle (B0C) reutronics parameters and minimum reactivity feedback, as in the UFSAR.
Additionally, control system operability was set to be the same Cs the UFSAR. The following Table summarizes assumed system status:
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Syt t en Op,Pr$hility ComTOntt Pressurizer Safety Valves operable nominal setpoint 31 accumulation Main Stecm Safety Valves operable nominal setpoint 31 accumulation Pressurizer PORVs inoperable Pressurizer Sprays inoperable Atmospheric Dump Valves inoperable Steam Dump to Condenser inoperable i
feedwater flow isolate on turtine trip Emergency feed flow inoperable Direct Reactor Trip on inoperable Turbine Trip High Pressure Reactor Trip operable Overtemperature AT Trip operable Pressurizer Heaters inoperable CVCS inoperable Figures 3-2 through 3-5 show system responses for the turbine trip event for both the RETRAN calculations and the UfSAR results, The RETRAN predicted peak pressurizer pressure (2543 psia) shows excellent agreement with the UfSAR result (2540 psia). The peak pressure in the primary system occur 5 at the bottom of the reactor vessel lower plenum. A peak pressure of 2639 psi, well below the 2750 limit, is reached momentarily during the event, Maximum steam generator pressure is calculated to be 1279 paia at the bottom of the downtomer, weli below the 1320 psia limit. Steam generator pressure is not reported in the UfSAR.
The nuclear power, pressurizer liquid volume and core inlet temperature responses are also comparable. There is a somewhat lower pressur,Zer pressure predicted by RETRAN later in the event since the steam line and header model in RETRAN is believed to be more detailed (additional volume is modeled) than
- the fuel vendor model presented in the UfSAR. The close comparison of the RETRAN and UfSAR results, particularly the peak pressure, demonstrates
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FIGURE 3-2 .
TURBINE TRIP WITHOUT PRilSSURIZER CONTROL, MINIMUM REACTIV}TY FEEDBAC PRESSURIZEG IPESSURE VS. TIME - UFSAR BENCilMARK 2600 - ItETRAl202 0
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TUR31NE TRIP WITHOU T PRESEURIZER CONTROL, MINIMUM REACTIVITY FEEDBACK NORMAllZED POWER VS. TIME - UFSAR BENCHMARK -
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l that the Seabrook RETRAN model is acceptable for predicting system responses for the turbine trip event, and in particular the peak primary and secondary system pressures.
Turbine Trip ReanT1ysis The RETRAN model was run with the pressurizer and steam generator safety valve opening setpoints increased by 3%. The only differences between the UFSAR benchmark RETRAN model and the model used for this analysis, in addition to the setpoint increases, was a pressurizer nomin,1 liquid volume of 1080 f t 3 with an uncertainty of 170 ft 3and a pressurizer pressure of 2200 psia. The revised liquid volume uncertainty represents current information, Initial pressurizer pressure is based on the nomin;' 2250 psia minus a new value of uncertainty of 50. psia.
Figure 3-6 shows the pressurizer pressure response versus time for the increased setpoint (+3%) case. Peak pressurizer pressure is 2620 psia. The peak primary system pressura reached is 2718 psia which is below the limit of 275 ' psie. The peak secondary side pressure is 1311 psia, which is also below the limit of 1320 psia. Therefore, increasing the PSV and MSSV setpoints bv
+3% will still maintain the peak pressure in both the primary and secondary ,
systems below their respective limiting values.
To assure that the bounding case has been considered, three sensitivity study cases were run with 1) initial pressurizer pressure at-the high limit of 2250 + 50 - 2300 psia, 2) initial pressurizer liquid volume at the low initial value of 1050 - 170 - 910 ft3 and 3) initial pressurizer-liquid volume at the high initial value of 1080 + 170 - 1250 ft3 . The most recent value for pr essurizer liquid volume uncertainty is 170 f t ,3 the value used above.
Each of the above sensitivity runs produced peak primary system and steam ger.erator pressures which were comparable to or less than the base case.
These sensitivity studies assure that the peak pressures determined above are bounding values. The following Table summarizes the results of the sensitivity study.
4 3
INITI AL PRESSURIZER PEAK PRE 3SURE CASE PRESSURE L10tl10 V0lllME PRIMARY SECONDARY (psia) 3 (ft ) (psia) (psia)
Base 2200 1080 2718 -1311 1 2300 1080 2712 1304-2 2200 910 2716 1312 3 2200 1250 2717 1309
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Inadvertent Closure of MSIVs This event is bounded by the turbine trip discussed above.-
The faster acting turbine control valves produce a more limiting overpressure t r e .1s i e n t ,
toss of Condenser Vacuum This event is bounded by the turbine trip discussed above.
Loss of Off-Site Power While this event results in a pressure increase, the event proceeds in_a much slower manner than the turbine trip. The pressure increase in both the primary and secc Jary systems is slow enough that the peak pressure is only a few psi above the opening pressure of the safety valves. Figures 15.2 5 and= <
15 -9 of the UFSAR show that the peak pressures in both the primary and secondary are less-than those calculated for the turbine trip event, i.e..
2543 psia for the cressurizer and 1279 psia for the secondary. It is noted that the turbine trip simulation assumes feedwater isolation' on turbine trip ,
and also the unavailability of emergency feedwater flow. These assumptions make the turbine trip a more limiting event in terms of the peak pressure.
Loss of Feedwater Flow Figures 15.2-11 and 15.2-14 of the UFSAR show that the -peak pressurizer and steam generator pressures are below those calculated for the turbine trip event. Therefore' this event is less limiting than the turbine trip event discussed above.
_Feedwater line Break Figures 15.2-17 and 15.2-20 of the UFSAR show that the-peak pressurizer and steam generator pressure are below those calculated for the turbine trip event. Therefore, this event is less limiting than the turbine trip event discussed above.
, _ _ _ _ _ _ _ _ - 1_.______ ._ _
3.1.3 Decrease in Reactor Coolant System Flow Rate Partial loss of Flow For this event the peak pressurizer pressure remains below the reduced safety valve setpoint of 2425 psia as shown in Figure 15.3-1. Sh.- 2 of-the UFSAR. Increase of secondary side pressure is not significant for this event.
Therefore, increasing 'Na safety valve setpoint tolerance will' not affect this event.
Complete loss of Forced Reactor Coolant Flow As with the partial loss of flow, pressurizer pressure remains well
' below the reduced safety valve setpoint as shown in UFSAR Figure 15.3 4 Sh. 1.. and secondary side pressure increase is not significant. Therefore, the increased safety valve tolerance will not affect this event.
Reactor Coolanti Pump Shaft Seizure (Locked Rotor)
As shown in Figure 15.3-7 of the UFSAR, the peak pressurizer pressure (2531 psia. Table 15,3-2 of tl,e UFSAR) remains below that for the turbine trip event discussed earlier (2543 psia). Therefore, this event-is less limiting than the turbine trip. Increase in secondary side pressure is not significant.
Reactor Coolant Pump Shaft Break The consequences of this. event are limited by the locked rotor event j described above.
3.1.4 Reactivity and Power Distribution Anomalies.
Uncontrolled RCCA Withdrawal from Subcritical Condition When the safety valve opening setpoints-are increased the peak pressure for this event will continue to .be bounded-by the turbine trip event discussed earlier. Decreasing the opening setpoint will cause the: event.to be less-severe. Therefore, increasing the safety-valve opening setpoint tol6r~.nce will not cause this event to exceed established pressure limits.
_ _ = _ _. _
)
Uncontrolled RCCA Withdrawal at Power As shown in-UFSAR Figures 15.4-4. Sh. 2 and 15.4-5. Sh. 2 the peak pressbrizer pressure for this event does not reach the reduced opening setpoint. Therefore, increasing the setpoint tolerance will not affect this event.
RLCA Misoperation This category includes several events all of which can reduce DNBR. Non of the events causes the pressure to increase above the reduced safety valve setpoints. Therefore, increasing the setpoint tolerance will have no affect on thes? events.
Startuo of an inactive Reactor Coolant pump As shown in UFSAR Figure 15.4-10. Sh. 4. the maximum reacter coolant pressure remains well below the decreased safety valve setpoint. Therefore, increasing the safety valve setpoint tolerance will not affect 1the results-for this event.
Baron Dilution This event is analyzed in all six modes of operation. During power operation, the consequences of this event are bounded by the RCCA withdrawal.
For shutdown operation, the operator has adequate time to take action-prior to_
criticality. There is no pressure increase and therefore the increased setpoint tolerance has no affect for this event.
Fuel loadino Error
?!o pressurization results from this event. Therefore.=the decreased safety valve setpoints cannot affect the results.
RCCA Eiection This is a Condition IV event and therefore emergency conditionIstress limits apply. From Reference 5 the emergency condition primary system 1 pressure limit is 3200 psig.
Generic analyses for this event are discussed in WCAP 7588. Rev. 1 (Ref. 6). Very conservative analyses of this event show that ,4ith a 18-
pressurizer safety valve setpoint of 2500 psia, the peak pressure will not exceed 2800 psia. The increase in peak pressure due to increased safety valve setpoint tolerance will be equal or less than the 75 psi increase in the setpoints. Maximum pressure will therefore remain below 2900 psia and the emergency condition stress limits will rot be exceeded.
less conservative three-dimensional studies reported in Reference 6 show that pressurizer surge rate is such that the pressurizer steam pressure will not rise above 2600 psia with a 2500 psia safety valve setpoint. It is noted that no credit is taken for flow through the PORVs. With an increase of-75 psi (+3%) in the safety valve setpoint the peak pressure will increase by no more than 75 psi and remain below 2750 psia. With the increased setpoint-the safety valves will open later during the event. However, the flow after opening will be greater due to the higher pressurizer pressure.
Increasing the safety valve setpoints will therefore not result in exceeding the applicable stress limits for this event. .<
3.1.5 Increase in Reactor Coolant Inventerv Inadvertent Operation of ECCS durino Power Operation The pressure does not reach the reduced safety valve setpoint during thia event and therefore the relaxed safety valve setpoint tolerance cannot affect this event.
CVCS M31 function that Increases Reactor Coolant Inventory The pressurizer PORVs are assumed to operate for this event. Since the pressure does not rise sijnificantly above the PORV setpoint. the' increased; safety valve setpoint tolerance will not affect this event.
3.1.6 Decrease in Reactor Coolant inventorv (Non-LOCA)
All of the events in this category result in pressure decrease. The safety valves are not required to lift and the setpoint pressure is not approached. Therefore, this class of events cannot be affected by the increased safety valve setpoint tolerance.
3.2 LOCA-Analysis 3.2.1 Larqt Break LOCA The postulated LBLOCA events in the Seabrook UFSAR do not challenge the, MSSVs because the primary system pressure draws down the secondary pressure almost immediately after initiation. The postulated SBLOCA events. however, by virtue of their break sizes (< 1 ft'), can challenge the MSSVs because the secondary side plays the role of heat sink early in the transiert.
3.2.2 Small Break LOCA In the UFSAR SBLOCA analysis, the limiting event was defined as a cold leg pipe rupture, at the CCCS injection location, with an equivalent diameter of 4 inches. This evGluation is performed for this limiting event since it yielded the highest peak cladding temperature -(?CT) of 1973.2 *F.
The UFSAR SBLOCA analysis assumes loss-of off-site power coincident with reactor scram. This causes closure of the main stearl isolation valves (MSIVs) and as a result an increase in secondary pressure'to the. MSSV setpoint, At the sar..> time, the initial subcooled break flow causes- the primary- system to depressurize to a pressure slightly above the MSSV setpoint where the portion of the de iy heat not removed by the break gets transferred to the secondary system. The secondary pressure is controlled by the safety valves as long as the break energy removal capacity is insufficient to remove decay heat. This time period is characterized by two-phase flow at the break. The primary liquid mass is distributed in the vessel and the loops with the pump suction loop seals filled with liquid. This condition remains until pump suction loop seal clearing occurs'ano steam produced in the core can be vented out the break. Vapor venting causes the energy removal at the break to exceed decay heat. Thus depressurizing the primary and secondary systems.
The MSSV setpoint assumption influences the primary heat transfer rate -
to the secondary and consequently.'the primary system presture beforE initiation of vapor venting. Therefore, the core mass inventory and by implicat on the extent of core heatup and the FCT are also affected.
The proposed MSSV setpoint tolerance rel'xation trar. slates into about 25 psia wider band. A higher MSSV setpoint ossumption causes a decrease in
.the primary heat transfer rate to the secca.ary resulting in a higher primary system pressure and thereby higher break.*iow. This, in turn results in decreased core mass inventory. In addition, the-higLer primary system
)
pressure causes decreased ECCS flow and, thus, decreased core mass inventory.
Correspondingly, a lower MSSV setpoint assumption results in increased core mass inventory due to both a decrease in break flow and an increase in ECCS.
flow.
Table 3-1 shows the results of the safety evaluation. A 25 psia increase in MSSV setpoint was assumed in order to bound the impact of the proposed Technical Specification change on the SBLOCA PCT. The decrease in core mass inventory due to lower ECCS injection and higher 'reak s flow was evaluated as described in Table 3-1. This resulted in a maximum increase in PCT of about 2.5 'F over the reported PCT of 1973.2 'F for the limiting SBLOCA event (4-inch 10).
3.3 SGTR Mass Release An evaluation of the effect on steam mass flow to the atmosphere was performed, An overall increase in steam generator steam mass release will
- result due to increased MSSV "open time" from 20 seconds +a 11.5 seconds. The MSSV "open time" however, occurs immediately following the reactor / turbine trip which is early on in the event at t- 904 seconds, and lasts only approximately 30 seconds. The incremental dose increase of approximately 2%
during this time period due to slightly longer MSSV open time, is small and 'Is within round-off error of the total dose-for either the 0-2 hour EAB or.the 0-8 hour LPZ doses.
Table 3-1 Effect of MSSV Setpoint Tolerance Relaxation or Seabrook Station SBLOCA Analys:s Interval of Increase in Total Equivalent Maximum Time After MSSV Setpoint Decrease In Vessel Level Increase in Pipe Rupture l Pressure 2 Vessel Mass Decrease 4
PCT 5 (*F)
(sec) (psia) Inventory 3 (in.)-
(lbm)80-385 25 2288 5.5 2,5 I
Notes:
- 1. Time interval of interest during the SBLOCA scenario when pressure of Drimary and secondary systems are controlled by the MSSVs (UFSAR Figure 15.6-31).
- 2. Equivalent to bounding increase in tolerance from +1% to +3% relative to the nominal MSSV setpoint pressure (Technical Specification Table 3.7-2).
- 3. Due to both an increase in break flow and a decrease in ECCS flow.
Saturated liquid was assumed for calculation of break flow > increase for the limiting break (4-inch 10) during the time interval of interest.
HEM tables were used to calculate critical flows at the break. The . ate of break flow increase was calculated to be about 7 lbm/sec for a 25 psia primary system pressure increase from 1212 psia to 1236 psia. 'This-result was also verified by using Moody tables to calculate critical flows at the break as required by Appendix K.
UFSAR Figure 15.6-43 was used to calculate the rate of FCCS flow decrease at 1200-1300 psia during the time interval of' interest. The rate of ECCS flow decrease was calculated to be about 0,5 lbm/sec for a 25 psia primary system pressure increase.
- 4. Calculation based on vessel flow area of 106 ft'.
- 5. In order to maximize the PCT increase, the slowest rate of vessel level decrease of 'aut 2.2 in./sec (UFSAR Figure 15.6-32) and"the may' mum rate of clad Heatup of about'1.*F/sec (UFSAR Figure 15.6-33) were used.
3.4 ASME Code Review The Seabrook Station PSVs and MSSVs were' designed and manufactu' red to meet the 1971 Edition including the Winter 1972 Addenda.and the ~1974 Edition including the Summer 1975 Addenda.respectively of_the-ASME.Co'e. d Section'111 which required the PSVs and MSSVs to be designed to open within- 1% of thefset pressure. The current Technical Specifications (TS) also impose a tolerance of 11% on the set p'ressure in the LC0 for the PSVs and MSSVs. However. the Surveillance Requirements of these TS require testing the PSVs and MSSVs under Section XI of the ASME Code. The in-service test program at Seabrook-is based' on the requirements of Paragraph IWV of the ASME Code,Section XI,-1983 Edition through the Summer 1983 Addenda. This Edition of Section XI does not specify a tolerance to be applied to lift pressure verification: _therefore, the tolerance prescribed in the LCO (11%).is used as the acceptance criteria for Section XI testing.Section XI also requires-that when any valve in a system f ails the setpoint criteria, additional valves in the system shall be-tested, and o valve failing to function during a test st.all be repaired or renlaced.
The 1989 Edition of the ASME Code,Section XI, requires that the PSVs and the MSSVs be tested per the standard ASME/ ANSI OM 1987, Part 1. This standacd allows the tested lift pressure to exceed.the stamped set pressure by up to 3% before declaring a test failure. It also provides a guideline for testing additional valves when a valve exceeds the i3% tolerance. .Therefore,.
increasing the PSV and MSSV setpoint_ tolerance to i3% for testing acceptance criteria is in compliance with the later Code requirements.
The PSVs and MSSVs have been reconciled to be in comnliance with the 1989 ASME Code Section III, Subarticle NB-7410/NC-7410, vdlich states that "The set pressure of at least one of the pressure relief devices connected to the-system not be greater than the Design Pressure of any component within the' pressure retaining boundary of the protected system" (in this case 2485 psig for the PSV and 1135 psig for the MSSV). The licensing basis analysis has bten reviewed / evaluated and it shows that the licensing basis criteria is still met when the increased +3% tolerance is applied to the relief pressures discussed above.
Seabrook Station will use the 3% tolerance for the "as found" acceptance criteria during valve testing and has committed to. reset the valve to within 1% prior to declaring the valve operable.
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4.0
SUMMARY
AND CONCLUSIONS An evaluation has been performed which demonstrates that no significant hazards consideration results when the Seabrook PSV and MSSV setpoint tolerance is incceased from 1% to 3%.
The evaluation centered on satisfying peak pressure limits in the primary and secondary systems. DNBR evaluations take credit for the PORV opening since lower pressure is more limiting. The PORV opens at a pressure 25 psi below the minimum pressurizer safety valve opening pressure.
Therefore, DNBR will not be affected by the revised safety valve setpoint tolerances.
For the lvwer limit of the setpoint (-3% tolerance) the PSV setpoint remains on the order of 25 psi above the power operated relief valve opening pressure and the high pressure reactor trip setpoint. The high pressure tr'p will therefore not be compromised, and challenges to the safety valve will not be 1 creased. Similarly, the team generator safety valve lower limit setpoint remains 24 psi above the atmospheric dump valve opening pressure and 58 psi above the no load operating steam pressure. These margins are sufficient to assure that challenges to the MSSVs will also not be increased.
The evaluation considered each of the events documulted in Chapter 15 of the UFSAR. Detailed calculations were performed for the limiting pressurization event, the turbine trip, utilizing the RETRAN02 M005 code.
These calculations demonstrated that peak pressure limits are not exceeded due to the increase in PSV or MSSV setpoint tolerances.
The impact of the proposed MSSV tolerance relaxation on the UFSAR design basis LOCA events were evaluated. The limiting LOCA analysis for Seabrook is a large break LOCA event with a PCT of 2041.2 'F. This is not affected by the proposed change to MSSV tolerance. The LOCA analysis event affected by this change is a SBLOCA. The proposed modification may result in a SBLOCA PCT increase of about 2.5 *F. A conservative PCT penalty of 5' F will be applied to the SBLOCA PCT result. The 5 'F PCT penalty should be tracked in accordance with 10CFR50.46 reporting requirements. This net PCT increase is less than 50 *F and, hence, is not a significant change to the Seabrook UFSAR SBLOCA analysis per 10CFR50.46 requirements. Also the revised SBLOCA PCT value (1978.2 *F) remains belcw the 2200 'F limit as well as below the large break LOCA results.
increasing the M55V setpoint tolerance-from i1% to 13% will not impact-the design bases SGTR calculation. Incremental ~ dose increases due to the slightly increased steam mass releases during-the time that MSSV are actually-open, will not impact the total dose beyonc ound-off error.
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5.0 RFFERENCES
- 1. Seabrook Station Updated Final Safety Analysis Report.
2 J. H. McFadden, et. al. *RETRAN-02 A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," EPRI Report .
l NP-1850-CCM A, Vol. 3. Rev, 4. November 1988. ',
- 3. United States Nuclear Ragulatory Commission, " Safety Eval"Ad$ **p
RETRAN02 M00005.0, RETRANQ2 Maintenance Groul,* transmitf>' In 'f from A. C. Thadani, USNRC to W. J. Boatwright Texas Ut9'i' d }[P - U Co., dated November 1, 1991.
4, D. E. Tuttle, Crosby Valve and Gauge Co., letter to W.LCloutier, Yankee Atomic Electric Co. " Valve Performance Data,", dated June 7,1985.
- 5. Letter NS-TMA-2182 Anderson, T. M. (Westinghouse Electric Corporation) to Hanauer, S. H. (USNRC), ATWS Submittal, dated December 30, 1979.
- 6. D. H. Risher, Jr. "An Evaluation of The Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatici Kinetics Methods "
WCAP 7588, Rev. 1, December 1971.
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