ML17264A341

Low Temp Overpressure Analysis Summary Rept

ML17264A341
Person / Time
Site:	Ginna
Issue date:	03/15/1995
From:	Merritt Baker, Cudlin J, Parece M BABCOCK & WILCOX CO.
To:
Shared Package
ML17264A337	List: ML17264A336 ML17264A338 ML17264A340 ML17264A341
References
86-1234820, 86-1234820-00, NUDOCS 9602130442
Download: ML17264A341 (46)

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Attachment IV Evaluation ofLTOP Limits Using RGAE Proposed Methodology 9602i30442 960209 PDR ADOCK 05000244 P

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BWNT NON-PROPRlETARY 86-1234820-00 Low Temperature Overpressure Analysis Summary Report Prepared for Rochester Gas & Electric Corporation Prepared by:

Reviewed by:

M.B. Baker, Engineer lll M. V. Parece, Supervisory Engineer Date:

Oat:

Approved by:

J. J. Cudiin, Manager, Analysis Services Unit Date:

B&WNuclear Technologies Lynchburg, Va

BWNT NON-PROPRIETARY

1.0 INTRODUCTION

86-I234820-00 BRW Nuclear Technologies (BWNT) updated the analysis of the low temperature overpressure (LTOP) events for the Rochester Gas and Electric (RGE) R.E. Ginna Nuclear Power Station (hereafter referred to as the Ginna plant).

Because the steam generators at the Ginna plant will be replaced in 1996, it was decided that the analysis should bound the Ginna plant with either'he existing or the replacement steam generators. Therefore, the replacement steam generators manufactured by Babcock and Wilcox International (BWI) were modelled in this analysis.

This analysis becomes the new analysis of record for the Ginna Station.

The results of the analyses of the limiting LTOP events were compared with 10CFR50 Appendix G and RHR overpressure limits.

In all cases, the peak reactor vessel and RHR system pressures were within the applicable limits.

2.0 DlSCUSSION OF LTOP EVENTS The United States Nuclear Regulatory Commission (USNRC) Regulatory Guide 1.99, Revision 2, dated May 19 discusses the effects of neutron irradiation embrittlement of low alloy steels used in the reactor vessel.

Appendix G of Chapter 10, Part 50 of the Code of Federal Regulations gives the fracture toughness requirements for the reactor vessel under low temperature conditions. During LTOP events, the reactor vessel temperatures and pressures approach the Appendix G limits. The LTOP system is designed to ensure that the reactor vessel embrittlement limits are not exceeded.

The LTOP events can occur during cold shutdown or plant heatup and cooldown.

To provide protection against exceeding the Appendix G limits, the Power Operated Relief Valve (PORV) on the pressurizer is reset to a low setpoint, which is appropriate for the plant conditions.

Two types of overpressurization events that can occur while the plant is operating with low pressure and temperatures are analyzed:

The first type of event is a mass addition event and the second type of event is a heat addition event.

2.1 MASS ADDITION LTOP EVENTS The mass addition events are characterized by addition of mass to a water-solid primary system.

This can occur during a shutdown situation, if the charging pumps or if the safety injection(SI) pumps are started inadvertently.

Technical Specification limits on Sl pump operability and discharge valve position eliminate the mass injection case due to a high head Sl pump start.

With three Sl pumps inoperable, an inadvertent Sl signal will not cause a pump start.

With the Sl system discharge paths

isolated, no single inadvertent operator action will result in safety injection. Therefore, the startup of three charging pumps with letdown isolated was analyzed as the bounding mass-addition event.

BWNT NON-PROPRIETARY 2.2 HEAT ADDITION LTOP EVENTS 86-1234820-00 The heat addition events are characterized by an addition of heat to a water-solid primary system. Heat can be added to the primary system by the actuation of pressurizer heaters, loss of the residual heat removal system (RHR) cooling, or two types of reactor coolant (RC) pump startups while a temperature asymmetry exists in the RC loops.

The inadvertent actuation of the pressurizer heaters when the pressurizer is water solid will cause a slow rise in the water temperature and increase in pressure, ifthe installed automatic pressure control equipment is not in service.

Since this pressure transient is very slow, the operator should recognize and terminate the transient before an unacceptable pressure is reached.

Ifthe operator does not terminate the transient, the pressure willincrease and be terminated by the PORV with little or no over shoot above the PORV setpoint.

This case is not significant to the design of the LTOP system.

The loss of RHR cooling when the RCS is water solid could be caused by a loss of flow malfunction in the component cooling water or service water systems, or the closure of the RHRs inlet isolation valves.

This would cause a slow rise in temperature and pressure since there would be a continual release of core residual heat into the reactor coolant with no heat removal.

This transient is also very slow and the operator has sufficient time to mitigate the event.

The first type of. temperature asymmetry can occur ifthe reactor coolant is at a relatively warm temperature with little or no natural circulation and cold reactor coolant pump seal injection water continues to enter the system.

The cooler injection water will settle in a pool in the loop seal.

The pressure transient is initiated by starting one reactor coolant pump. As the pump comes up to speed, the reactor coolant flowrate slowly increases in the active loop and the pool of cold water willbe drawn up into the pump and discharged out the cold leg piping.

Simultaneously, the pool of cold water in the inactive loop will flowbackward through the steam generator at a flowrate significantly less than that of the active loop.

As this pool of cold water flows through the steam generator, the temperature will increase due to heat transfer from the secondary side of the steam generator.

This causes expansion of the primary side water and an increasing pressure transient.

The second type of temperature asymmetry occurs when the RCS has been cooled without sufficient circulation. This could occur when the RHR system is used to cool the RCS without use of any reactor coolant pumps. Under these conditions, the water in the steam generator secondary side and the primary side will be in thermal equilibrium at a temperature higher than that of the reactor coolant. If one RC pump is inadvertently started under these conditions, the RCS flowrate increases and the cold water from the

'CS enters the SG tubes.

This results in the transfer of heat from the secondary to the

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BWNT NON-PROPRIETARY 86-1234820-00 primary system, causing the primary system liquid to expand and the primary system to pressurize.

This is a relatively fast event and, because of the transfer of heat from the secondary system to the primary system, this event is the most limiting heat addition transient.

3.0 EVENTS ANALYZED The limiting mass addition case, the inadvertent startup of three charging pumps, was analyzed at 85 F. This temperature is used because it is the lower temperature limitof the Appendix G limits, and, at the lower limit, the fluid has the minimum compressibility.

The limiting heat addition case is the inadvertent start of a reactor coolant pump following RCS cooldown solely with the RHR system. This event is analyzed at RCS temperatures of 85 F and 320 F with the SG liquid temperature 50 degrees hotter than the RCS. The transient is analyzed at 85 F since it is the lower limitof the Appendix G limits and has the lowest pressure limitfor the acceptance criterion.

The event is analyzed at 320 F because this is the maximum credible temperature at which a secondary-to-primary temperature difference of 50 F can be achieved.

Specifically, the reactor coolant pumps may be tripped at 350 F.

With instrument uncertainties, the temperature could be as high as 370 F.

If the RCS is subsequently cooled to obtain the maximum allowed temperature difference (50 F), the RC pump start could occur at 320 F.

This heat addition event is the most limiting for the RHR overpressurization.

4.0 ACCEPTANCE CRlTERIA The acceptance criteria for the LTOP events are:

1.

The pressure and temperature of the reactor vessel can not exceed the Appendix G limits, which are depicted in Figure 1.

This figure is obtained from Reference 2.

'I 2.

The pressure in the RHR system can not exceed 110 percent of the design pressure of 600 psig, or 660 psig.

5.0 METHODOLOGY The LTOP transient analyses were performed using the RELAP5/MOD2 B&WVersion 20 (Reference 5) computer

code, which has received full BWNT certification.

RELAP5/MOD2-B&Wis a two-fluid, six equation, nonhomogeneous, nonequilibrium

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BWNT NON-PROPRIETARY 86-1234820-oo RCS PRRSSURX (PSXG) 2500

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1500 UNACCEPTABLE OPERATION 1250 0 F 750 ACCEPTABLE OPEULTZOH 329'F 250 0

0 50 150 200 250 300 350 400 450 5(

RCS TEMPERATURE (DEC F)

Figure 3.3.-1:

Ginna Reactor Vessel Heatup Limitations hpplicable for the first 21 EPPY using Reg Guide 1.99, Rev.

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BWNT NON-PROPRIETARY 86-1234820-00 thermal-hydraulic code developed forbest-estimate transient analysis ofpressurized water reactors and associated systems.

The code has options to consider equilibrium, homogeneous hydrodynamic control volumes and a limited ability to calculate conditions for co-existing noncondensibles.

The numerical solution technique is semi-implicit finite differencing.

RELAP5 is a highly flexible code that, in addition to calculating NSS behavior, can be used for simulation of a wide variety of thermal-hydraulic transients.

RELAP5/MOD2-BBWhas special process models that are not available in the industry version of the code.

The only such process model used in these LTOP analyses is the Henry-Fauske extended subcooled critical flow model.

For those instances when the pressurizer PORV. experienced critical flow, the extended Henry-Fauske critical flow model was used rather than the Ransom-Trapp model.

The extended Henry-Fauske model was used because it is widely accepted for use over the range of conditions experienced in these analyses and because the Ransom-Trapp model overpredicts the.

test data using a discharge coefficient of 1.0 (Reference 3).

The plant model that was employed for the LTOP analyses included two complete reactor coolant loops including RC pumps and steam generators.

The secondary side included steam lines, main steam safety valves (MSSVs), main steam isolation valves (MSIVs),

and turbine stop valves.

A noding diagram of the RELAP5/MOD2 model is shown in Figure 2. The steam generator model used for the analyses is a simulation of the U-tube replacement steam generator designed by BWI. This steam generator design has 21.3 percent higher heat transfer surface area than the current steam generators used in the R.E Ginna Nuclear Power Station.

Use of this model in the LTOP analyses accommodates the steam generator replacement, while bounding the current steam generators.

The feedwater systems and the auxiliary feedwater systems are not modelled since these are not functioning during the LTOP events.

The primary system has a reactor vessel model with two equal and parallel core paths for adjusting the mixing of loop flows in the lower plenum. This feature is not used in the LTOP analyses as this is not required.

The core has six axial nodes and a core bypass with three nodes.

The upper and the lower plenum volumes are common.to both the loops, whereas the downcomer is split into two parallel set of volumes. A noding diagram of the reactor vessel is shown in Figure 3.

The pressurizer was modelled as a ten node vertical pipe component and was initialized liquid solid. One PORV is attached to the top node of the pressurizer.

Only one PORV is modelled because the other PORV is assumed to fail closed.

The PORV was set to liftwhen the pump suction pressure on the loop with the pressurizer exceeded 430 psig, consistent with the location of the pressure transmitters and instrument error.

Instrument error is not normally considered for Appendix G protection but is included in this analysis for RHR overpressure concerns.

The PORV model was sized to deliver 49.722 Ibm/s

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$55 900 9&5 970 975 MSIV 831 MSSV MSIV 87S 8?8 To thh model, added

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185 170 175 180 RC Pump 8 LEFT LOOP REACTOR VESSEL RIGHT LOOP RC 280 Pump A 150 1&5

Figure 3

G)NNA REACTOR VESSEL AND CORE MODEL hot leg nozzle 350 OO C9 302 364 360

)eakage path upper p)enum 354 352 upper p)enum 368 hot feg nozzle 351 349 ZR 0z I

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cold leg nozzle Ot Ch 326 324 320 338 316 338 336 332 330 328 o

C9 cold leg nozzle 312 310 375 380 Nots: Thhb a teocorechared modsr -~~ cores C) t Q

BWNT NON-PROPRIFTARY 86-1234820-00 saturated steam at 2335 psig.

The opening stroke time was 1.0 second using the C characteristics in Table 1.

The model contains the piping from the PORV to the pressurizer relief tank (PRT) as well as the PRT with a rupture disc. The nitrogen blanket on the PRT is modelled.

The RC pumps were modelled as centrifugal pumps with the homologous curves representing the performance under various conditions. The pump performance curves shown in the UFSAR were used as the basis for the active octants in this pump model.

The passive metal of the whole system was modelled for the LTOP analyses.

The passive metal included the reactor vessel walls, the reactor internals, the fuel end fittings, the hot and cold leg pipe walls, pressurizer walls, the steam generator primary side metal and the steam generator secondary side metal.

The steam generator tube metal was modelled as part ofthe active heat structures.

The steam line metal and the RHR system passive metal were not modelled.

The RHR system was modelled as two parallel trains with two separate pumps and cross connects.

Two heat exchangers were modelled as control volumes with no heat removal since the heat exchangers were assumed to maintain a constant temperature in the RCS during the LTOP analyses.

The RHR relief valve was attached to the RHR system near the cold leg connection.

The RHR relief valve was benchmarked for flow under the design conditions.

A noding diagram of the RHR system is shown in Figure 4.

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BWNT NON-PROPRIETARY 86-1234820-00 TABLE 1 Cversus position Copes Vulcan Valve - Model Number D-100-160 Stroke %

0.0 0.0 Cv normalized 1,.9 7.9 14.0 20.0 26.1 32.2 38.2 44.3 50.3 56.4 62.5 68.5 73.6 78.3

'4.5 91.6 98.6 100.000 0.016 0.067 0.143 0.231 0.346 0.474 0.626 0.734 0.823 0.878 0.924 0.957 0.970 0.977 0.985 0.992 0.999 1.0

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RHR SYSTEM MODEL Primary Loop (LTOP Active Loop)

Hot l eg Primary (Inactive Loop)

Cold Leg 100 280 450 451 453 RHR pump B 462 Heat exchanger B 463 473 460 457 469 sink RHR Relief Valve 468 452 472 455 461 454 RHR pump A Heat exchanger A 475 456 CO CJl I

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BWNT NON-PROPRIETARY 86-1234820-00 For the mass addition case, the primary system was initialized at 85 F and 315 psig with no reactor coolant flow. The primary and secondary systems were decoupled since there was no heat transfer in this case.

The event was initiated by starting three charging pumps with a total capacity of 180 gpm. The analysis was terminated after 10 minutes when the operator was assumed to secure charging flow. The peak RCS pressure was compared with the acceptance criteria.

For the heat addition cases, the primary system was initialized to isothermal conditions with no reactor coolant fiow. The secondary and primary fluid in the steam generators were initialized at a temperature 50 degrees above the primary system. The RHR system was running with a capacity of 1700 gpm, as specified in Attachment C of Reference 4.

The capacity obtained in our model was 1701 gpm.

This is an obtained value, not a specified one, and this result was considered adequate.

The transient was initiated by starting a reactor coolant pump in the loop that contains the pressurizer.

The pump start-up characteristics of Table 3 were used to bring the pump to full speed in 17.4 seconds.

The analysis was run until the peak pressure was obtained.

The peak pressures in the reactor vessel and the RHR system were compared with the acceptance criteria.

6.0 ANALYSIS

-The sections following describe the initial and boundary conditions for each event analyzed, as well as the results.

All values were taken from Reference 4.

6.1 Mass Addition Case The mass addition case was initialized at a primary temperature of 85 F and a primary pressure of 315 psig.

Using the initial pressure of 315 psig assures that the. transient is well defined by the time the PORV is actuated.

There was no flow in the reactor coolant system and the pressurizer was water solid.

It was assumed that the RHR system was removing decay 13

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The event was initiated by starting three pump charging flow (180 gpm or 25 Ib/s). The analysis was run for ten minutes. The sequence of events for this case is shown in Table 2.

Plots of Loop B hot leg pressure, Loop A hot leg

pressure, and reactor vessel pressure are shown on Figures 5 - 7, respectively.

The peak reactor vessel pressure was 480.2 psia. The allowable pressure, according to the Appendix G limit at 85 F, is 540 psig or 554.7 psia.

Therefore, there is 74.5 psi margin to the Appendix G acceptance criterion.

To compare the peak pressure in the RHR system with the acceptance criterion, the pressure drop from the hot leg to the RHR pump discharge (128.1 psi, from Reference

4) was added to the peak hot leg pressure.

This case yielded a peak RHR pressure of 598.4psia.

The peakallowable pressure in the RHR systemis674.7 psia. Thisresults in a.76.3 psi margin to the acceptance criterion.

TABLE 2 SEQUENCE OF EVENTS - MASS ADDITIONCASE EVENT Charging pumps started.

Charging pump reach full flow Peak pressure oF.'fgo2.I,'s~ reached in the bottom of the RV Peak pressure of 470.3 psia reached in the hot leg connection to RHR 0.0 1.0 534.0 534.0 TIME, SECONDS

LTOP EVENT-MASS'DDlTIONCASE CASE I Primary temperature 85 < F, Secondary temperature 136 o F Three charging pumps operating 450 "-"-"--""

CD CD ii 350 300 250 200 0

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LTOP EVENT-MASS ADDlTlONCASE CASE 'l Primary temperature 86 0 F, Secondary temperature 136 < F Three charging pumps operating 450 400 350 300 200 0

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The RHR system was operating with a capacity of 1701 gpm, or 236.47 Ibs/sec.

The pressurizer was water solid. There was no charging flow for this event.

The event was initiated by starting the RC pump in the loop that contained the pressurizer.

The pump start-up profile that was used in the analysis is shown in Table 3. The transient was analyzed for 40 seconds The sequence of events for this case is shown in Table 4.

Figures 8 - 16 show plots of RV pressure, RHR pressure, PORV flow, secondary system temperature in the tube region of Loop 8, secondary system temperature in the tube region of Loop A, flow.in RC loop 8, flow in RC loop A, hot leg temperature in the Loop 8, and hot leg temperature in Loop A, respectively.

Figure 8 depicts the first pressure peak, which is the highest pressure, as illustrated by the decreasing slope.

The peak pressure in the reactor vessel for this case was 546.8 psia.

The allowable pressure for the Appendix G limitat this temperature is 554.7 psia. This yields a 7.9 psi margin.

This case is the most limiting for Appendix G.

It is stated in Reference 6 that it is acceptable to exclude the pressure instrument error for the LTOP evaluation because the low temperature overpressure protection system is considered to be a mitigation system, as opposed to a protection system.

This analysis assumed a 20 psi instrument error for the PORV setpoint.

In a water solid system, it can be expected that a 20 psi reduction in the setpoint can yield at least a 20 psi reduction in the peak pressure.

Ifadditional margin to the Appendix G limitis needed in the future, this instrument error could be credited.

The peak pressure in the RHR system was 640.8 psia as compared with an acceptance criterion of 674.7 psia, for a margin of 33.9 psi.

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~ I BWNT NON-PROPRIETARY TABLE 3 PUMP STARTUP PROFILE 86-1234820-00 0.0 3.5 6.6 9.7 13.3 15.8 Time,sec 0

240 480.

720 960 1080

Speed, rpm 17.4 1189 (FULL SPEED)

TABLE 4 HEAT ADDITIONat 85 F - SEQUENCE OF EVENTS EVENT RC pump started in loop that contains the pressurizer.

RC pump reaches full flow 0.0 17.4 TIME, SECONDS PORV opening signal for the first time 23.2 Peak pressure reached in the RV Peak pre'ssure reached in RHR 23.2 23.2 19

560 LTOP EVENT-HEAT ADDiTIONCASE CASE 2 Primary temperature 85 0 F, Secondary temperature

'i 35 0 F One RC pump starting 520 480 440 400 360 320 10 15 20 TIMEIN SECONDS 30 35

LTOP EVENT-HEATADDITIONCASE CASE 2 Primary'emperature 86 o F, Secondary'emperature i36 0 F One RC pump starting 680 640 600 560 520 480 440 0

10 15 20

'PIME IN SECONDS 1-o nitro 0

30 35

120 LTOP EVENT-HEATADDmON CASE CASE 2 Primary temperature 86 0 F, Secondary temperature 136 0 F One RC pump starting 100 80 60 40 20 CO Ch I

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BWNT NON-PROPRIETARY 6.3 Heat Addition at 320 F 86-1234820-00 The second heat addition case was analyzed with steam generator secondary system temperatures of 370 F.

This temperature is the maximum temperature, including instrument uncertainty, at which both reactor coolant pumps can be stopped. The primary system was assumed to'be 50 degrees colder than the secondary

system, so the temperature in the primary system was 320 F. Initially, the RC pumps were not running and cooling was assumed to be provided by the RHR system.

The RHR system was operating with a capacity of 1701 gpm, or 215.75 Ibs/sec.

The pressurizer was water solid. There was no charging flowfor this event. The transient was initiated by starting the RC pump in the loop that contained the pressurizer.

The pump start-up profile that was used is shown on Table 3.

The sequence of events for this case is shown in Table 5.

Figures 17 - 25 show plots of RV pressure, RHR pressure, PORV flow, secondary system temperature in the tube region of Loop B, secondary system temperature in the tube region of Loop A, flow in RC Loop B, flow in RC Loop A, hot leg temperature in Loop B, and hot leg temperature in Loop A, respectively.

The peak pressure in the reactor vessel for this case was 563.8 psia.

The allowable pressure for the Appendix G limitat this temperature is 1664.7 psia. This yields a 1100.9 psi margin.

The peak pressure in the RHR system was 655.7 psia as compared with an acceptance criterion of 674.7 psia, for a margin of 19.0 psia. This case is the most limiting for peak RHR pressure.

TABLE 5 HEAT ADDITIONat 320 F - SEQUENCE OF EVENTS EVENT RC pump started in loop that contains the pressurizer.

PORV opening signal for the first time Peak pressure reached at RHR pump outlet RC pump reaches full flow Peak pressure reached in the RV 0.0 8.81 10.5 17.4 21.3 TIME, SECONDS 29

LTOP EVENT-HEAT ADDITIONCASE CASE 3 Primary temperature 320 o F, Secondary temperature 370 0 F One RC pump starting 550 500 450 350 300 0

10 15 20 TIMEIN SECONDS Cinee~n 1 I 30 35

LTOP EVENT-HEATADDITION CASE CASE 3 Primary temperature 320 0 F, Secondary temperature 370 0 F One RC pump starting 650 600 550 500 450 I

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BWNT NON-PROPRIETARY 7.0

SUMMARY

AND CONCLUSIONS 86-1234820-00 B&WNuclear Technologies updated the analysis of the low temperature overpressure events forthe Rochester Gas and Electric (RGE) R.E. Ginna Nuclear Power Station. This analysis will become the new analysis of record for RGE.

In this effort, the most limiting LTOP events were analyzed and compared with the acceptance criteria of the Appendix G limits for embrittlement and the RHR overpressure limit. This analysis was performed with the BWI steam generators, to bound the plant as it exists now and to bound the plant when the steam generators are replaced.

The most limiting mass addition and heat addition cases were analyzed and the results were compared with the acceptance criteria.

The mass addition case, which was analyzed at a primary temperature of 85 F, had a peak pressure in the reactor vessel of 480.2 psia.

This is 74.5 psi below the acceptance criterion for Appendix G. The peak pressure in the RHR system was 598.4 psia, which is 76.3 psi below the acceptance criterion.

The heat addition case that was performed at a primary system temperature of 85 F yielded a peak reactor vessel pressure of 546.8 psia.

The allowable pressure for the Appendix G limit at this temperature is 554.7 psia.

This yields a 7.9 psi margin.

The peak pressure at the RHR pump discharge forthis case was 640.8 psia as compared with an acceptance criterion of 674.7 psia, for a margin of 33.9 psi.

The heat addition case that was performed at a primary system temperature of 320 F yielded a peak reactor vessel pressure of 563.8 psia.

The allowable pressure for the Appendix G limit at this temperature is 1664.7 psia.

This yields a 1100.9 psi margin.

The peak pressure at the RHR pump discharge for this case was 655.7 psia as compared with an acceptance criterion of 674.7 psia, for a margin of 19.0 psia. Therefore, all of the LTOP analyses that were performed with the BWI steam generators were within the acceptance criteria for both Appendix G and RHR overpressure.

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BWNT NON-PROPRIETARY

8.0 REFERENCES

86-1234820-00 2.

4 6.

US Nuclear Regulatory Commission Regulatory Guide 1.99, "Embrittlement of Reactor Vessel Materials", Revision 2, May 1988.

R.E. Ginna Nuclear Power Plant Technical Specifications Amendment 57, March 1994.

NUREG/CR-5194 EGG-2531 R4, "RELAP5/MOD2 Models and Correlations",

August 1988.

BWNT Document 32-1232650-00, "LowTemperature Overpressure Analysis for RGE - Ginna Plant", February 1995.

BAW-10164P, "RELAP5/MOD2-BRW-An Advanced Computer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis," Code Topical Report, Revision 2, August 1992.

"Trojan Unit LowTemperature Overpressure Protection System (LTOPS) Setpoint Evaluation", June 1990, pg. J-i (akba.eked).

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