ML17254A586

Proposed Tech Specs Re Containment Internal Pressure Limitations

ML17254A586
Person / Time
Site:	Ginna
Issue date:	10/16/1985
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17254A584	List: ML17254A582 ML17254A585 ML17254A586
References
NUDOCS 8510210134
Download: ML17254A586 (42)

Text

3.6 Containment, S stem A licabilit :

Applies to the integrity of reactor containment.

~b'o define the operating status of the reactor containment.

for plant operation.

S ecification:

'ontainment Inte" rit 3,.6.1

'. 'xc'ept as al'lowed by 3'.6.3 containment shall not be violated unless the integrity reactor is in the cold shutdown condition.

b. The containment. integrity shall not be violated when the reactor vessel head is removed unless the boron concentration is greater than 2000 ppm.

c Positive reactivity changes shall not be made by rod drive motion or boron dilution whenever the containment integrity is not intact unless the boron concentration is greater than 2000 ppm.

3.6.2 Internal Pressure If the internal pressure exceeds 1 psig or the internal vacuum exceeds 2.0 psig, the condition shall be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the reactor rendered subcritical.

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Basis:

The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the reactor coolant system ruptures.

The shutdown margins are selected based on the type of activities I

that are being carried out. The (2000 ppm) boron concentration provides shutdown margin which precludes criticality under any circumstances. When the reactor head is not to be removed, a cold shutdown margin of 1/Pk/k precludes criticality in any occurrence.

Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pg~sure before a major steam break accident, were as much as 1 psig.

The contaj~ent is designed to withstand an internal vacuum of 2.5 psig. The 2.0 psig vacuum is specified as an operating limit to avoid any difficulties with motor cooling.

References:

M (1) Westinghouse Analysis, "Report for, the BAST Concentration Reduction"', August, 1985" (2) FSAR Section 5.5 3.6-3 Proposed

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Attachment B The original licensing basis for containment integrity at Ginna was the Loss of Coolant Accident (LOCA). During the Systemmatic Evaluation Program (SEP) this licensing basis was reviewed. The results of the review concluded that the large steam break inside containment was more limiting than the LOCA for containment integrity. Therefore, the licensing basis for containment integrity became the large steam break supported by analysis done for the Staff by Lawrence Livermore National Laboratory and by Rochester Gas and Electric Corp. (RG&E)

Recently RGSE has contracted Westinghouse Electric Corp. to perform analysis to evaluate the possibility of reducing boron concentration in the Boric Acid Storage Tanks (BAST). A byproduct of this evaluation is a new containment integrity analysis (Enclosure 1).

This analysis does not invalidate the previously approved SEP analysis. The new analysi's uses a different methodology, different assumptions, different codes, and is better documented "than the SEP analysis. It is proposed -that the new analysi's become the design basis containment integrity analy'sis. Since the SEP analysis used different.,assumptions, substituting, the, new analysis as the licensing basi.s" necessitates revising the Technical Specifications. Tobe"consistant with the in'itial'onditi'ons assumed in the new'nalysis, containment pressure should be to 1 psig. 'imited In accordance with 10 CFR 50.91, this change to the Technical Specifications has been evaluated against'hree criteria to determine if the operation of the facility in accordance with the proposed amendment would:

1. involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. involve a significant reduction in a margin of safety.

The proposed change would decrease the initial containment pressure before a major steam break inside containment and therefore does not increase the probability or consequences of a previously evaluated accident or create the possibility of a new or different kind of accident or involve a significant reduction in a safety margin. Therefore, a no significant hazards finding is warranted for the proposed change.

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ENCLOSURE 1 REPORT FOR THE BAST CONCENTRATION REDUCTION FOR R. E. GINNA August 1985 8709 Q: 10/051 585

INTRODUCTION Westinghouse has developed improved analytical techniques which allow a reduction in the Boric Acid Storage Tank (BAST) concentration. This report provides background information on the BAST design basis, reasons why boron reduction may be desirable, plant design features which allow the change to be proposed, as well as a sutreary of analytical results which demonstrate the feasibility of this option on the BAST system for Ginna.

BACKGROUND The two BASTs are components of the Chemical and Volume Control System which also pro'vides concentrated boric acid to the reactor coolant to mitigate the consequences of postulated steamline break accidents. In this function, they act as part of the Safety Injection System. Although the BASTs act to

. mitigate steamline break of various sizes occurring from any power level, the cases which serve as the Westinghouse steamline break licensing basis, and which define the existing requirements on the minimum BAST boron concentration, are as follows:

For the "hypothetical" steamline break, i.e., double ended rupture of a main steamline, the radiation releases must remain within the requirements of 10CFR Part 100. This is the ANSI N18.2 criterion for Condition IV events, "Limiting Faults." Westinghouse conservatively meets this for Ginna by demonstrating that the DNB design basis is met, the criterion typically used for Condition II events.

For the "credible" steamline break, i.e., the failure open of a single steam generator relief, safety, or turbine bypass valve, that radiation releases must remain within the requirements of 10CFR Part 20. This is the ANSI N18.2 criterion For Condition II events, "Faults of Hoderate Frequency." Westinghouse conservatively meets this criterion by showing that the DNB design basis is met.

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In order to assure the dity of the safety analyses p ormed to verify that the evaluation criteria are met, Technical Specifications have been applied to the BAST and associated equipment. Specifically, these assure that the boric acid concentration is maintained in excess of 20,000 ppm, approximately a 12 weight percent solution. In order to maintain this high concentration, heat tracing of the tanks and associated piping is required.

Furthermore, the safety-related nature of the boric acid system requires that the heating systems be redundant.

The required solubility temperature imposes a continuous load on the heaters, and low-temperature alarm actuation and heater burnout have occurred in some operating plants. Violation of the Technical Specification on concentration in the BAST poses availability problems in that recovery is required within a very short time. If the concentration is not restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the plant must be taken to the hot shutdown condition. Thus,,this requirement has a potentially serious impact on plant availability.

These potential difficulties unfavorably affecting plant availability, operability, and maintainability can be drastically reduced in severity or eliminated by reducing the boron concentration to a minimum level at which heat tracing would no longer be required. The effect of this change is discussed in the following section.

DESCRIPTION OF THE ANALYSES The only accident analyses which are significantly affected by boron, concentration reduction are the steamline break transients. Since the steam break affects the core and the containment responses, both of these were considered in the boron concentration reduction analysis. The following analysis consists of a core analysis and a containment mass-energy analysis.

CORE ANALYSIS The following cases must be considered For the BAST boron concentration reduction with respect to the core analysis.

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a. Complete severance pipe inside the containment he outlet of the steam generator at initial no-load conditions with outside power available and two loops in service. The equivalent break area is 4.37 sq. ft.

b. Case (a) above with loss of outside power simultaneous with the steamline break.

c. A break equivalent.to steam release through one steam generator safety valve with outside power available and two loops in service.

d. Case (a) above with only one loop in service.

e. Case (c) above with only one loop in service.

The severance of a pipe downstream of the steam flow measuring nozzle is not analyzed. The equivalent break area (1.4 sq. ft.) is less than that of case (a) and would result in a less severe cooldown. Thus, this break is bounded by cases (a) and (b).

Of these cases, cases (a) through (c) were analyzed with the BAST concentration at 2000 ppm in the Reload Transition Safety Report (RTSR) and approved by issuance of a Technical Specification change . The results of these analyses in the RTSR show that the DNB design is met. Thus, only cases (d) and (e)- need be considered here.

1. NRC Letter, R. M. Kober (NRC) to D. H. Crutchfield (RGIIE), Application for Amendment to Technical Specifications, December 20, 1983.

2. Amendment No. 61 to the R. E. Ginna Technical Specifications dated Hay 1, 1984 ~

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Anal sis method As in the Ginna RTSR steamline break analysis, the system transient parameters, i.e., RCS pressure, temperatures, steam flow, core boron (3) system concentration and core power are calculated using the LOFTRAN transient analysis computer code. This computer code includes models of the reactor core,'team generators, pressurizer, primary piping, protection systems and engineered safeguards systems.

The results presented are a conservative indication of the events which would occur assuming a steamline rupture. The worst case assumes that all of the following occur simultaneously.

1. Hinimum shutdown reactivity margin equal to 2.45 percent (1 loop in service).

2. The most negative moderator temperature coefficient for the rodded core at end-of-life.

3. The rod having the most reactivity stuck in its fully withdrawn position.

4. One safety injection pump fails to function as designed.

The plant is initially assumed to be at hot zero power at the minimum required shutdown margin. Following the break, the RCS temperatures and pressures decrease rapidly, and in the presence of a large End-of-Life (EOL) moderator coefficient of reactivity, the reactor returns critical with the rods inserted, assuming the most reactive RCCA in the fully withdrawn position.

The reactor power increases at a decreasing rate until boron from the safety injection system reaches the core and begins to offset the positive reactivity insertion caused by the cooldown. The core is subsequently brought subcritical with boron injection, aided by the abatement and eventual termination of steam flow from the broken steam generator.

3. WCAP-7907, T.W.T. Burnett, et. al., "LOFTRAN Code Description," October, 1972.

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Figures 1 through 5 show the transient behavior for the 4.37 sq. ft.

Hypothetical Break with one loop in service with the BAST concentration equal to 2000 ppm (case d). A comparison of the RTSR cases (fig. 14.2.5-18) with Figures 1-5 reveals that the reactor coolant system transients are similar, with the single exception of core power, which is understandably higher for the case with reduced boron concentration in the BAST. The effect of the boron on the total reactivity is both delayed and damped in Figure 1 because the boron source is both colder and of a lower boron concentration. This causes the heat flux to initially rise to a higher peak (33$ of 1520 NMth) and to subsequently decay at a slower rate after the boron reaches the core. A ONB analysis for 'this transient shows that the minimum DNBR is above the limit value, thus no fuel failure is predicted due to ONB.

Figures 6 through 8 depict transient parameters for the Condition II steamline

'reak, assuming 2000 ppm in the BAST (case e). In the RTSR, the reactivity plot in Figure 14.2.5-25 shows that the reactor remains subcritical. This assures that the ONB design basis is met in a very conservative manner. The reactor, also remains subcritical when the BAST is at 2000 ppm. Boron enters the core while the reactor is still significantly shut down, as 'can be seen in Figure 8. Since the reactor remains,subcritical, the DNB design basis is met.

The sequence of events is presented in the attached table.

In conclusion, calculations have been performed for Ginna which show that from the ONB standpoint BAST concentration can be reduced to 2000 ppm since the DNB design basis is met. For 2 loop operation, this analysis is contained in the RTSR. The analyses presented here show that the results are acceptable for operation with one loop in service.

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HASS AND ENERGY ANALYSIS Steamline ruptures occurring inside a reactor containment structure may result 1n s1gnificant releases of high-energy fluid to the containment environment, possibly resulting in highcontainment temperatures and pressures. The quantitati.ve nature of the releases following a steamline rupture is dependent upon the many possible configurat1ons of the plant steam system and containment designs as well as the plant operating conditions and the size of the rupture.

The following cases have been reanalyzed for the BAST boron concentration reduction.

Large or Full Double-Ended Steamline Ruptures Small Double-Ended Steamline Ruptures Split Steamline Ruptures The large breaks analyzed are listed in Table 2; the small break analyzed are listed in Table 3; and the split breaks analyzed are 'listed in Table 4. These break sizes were chosen because the 4.37 sq. ft. is the largest break that can occur. The 1.4 sq. ft. break is the largest break that can occur downstream of the flow restrictor. The split break is chosen to be the largest break which can occur such that protection is actuated by the containment signals, rather than the primary signals (low steam pressure, high steam flow, etc.). =

I The hot zero power and hot full power cases have been analyzed since these have been previously defined by the NRC to be the steamline break mass and (4) for R. E. 61nna.

energy release inside containment 11cens1ng requ1rement

4. NRC Letter, 0. H. Crutchfield (NRC) to John E. Haier (RGhE), "Evaluation Report on SEP Topic VI-2.D and VI-3," November 3, 1981.

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Anal sis Hethod The mass and energy analysis 1s initiated by using the LOFTRAN (3) code to determine the mass and energy released to the containment during a steamline break. The mass and energy data is then used by the COCO (5) code to determine temperature and pressure response in the containment following a steaml1ne break accident. The bas1c initial conditions, heat sink model and fan cooler parameters employed 1n the containment response calculation are outlined in Tables 5 through 7. The following conservat1ve assumptions are.

made for each mass and'energy release analysis:

l. Haximum decay heat equivalent to 120$ of , ANS finite model.

2. Ho credit is taken for water entrainment in the blowdown results.

3. Conservatively high values for reverse steam generator heat transfer.

4. The most negative moderator temperature coefficient for the rodded core at end-of -l i fe.

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One containment spray pump fails to function.

6. Offsite power 1s ava1lable throughout the trans1ent.

Figure I 9 provides the pressure and temperature curves for the limiting large

'reak case prov1d1ng the highest peak containment pressure and temperature of those cases listed in Table 2. Figure 10 provides the pressure and temperature curves for- the limiting small break. case providing the highest peak containment pressure and temperature of those cases listed in Table 3.

Figure ll provides the pressure and temperature curves for the limiting split break case providing the hi'ghest peak containment pressure and temperature of those cases listed in Table 4. These latter two curves are not representative of the split break accident with a single failure assumed since all three

5. Bordelon, F. M., and Murphy, E. T., Containment Pressure Analysis Code (COCO), NCAP-8326, Dune 1974.

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fa1lures were included. Because of the margin observed in the peak pressures and temperatures of the large and small steamline breaks when a single fa1lure was assumed, all those s1ngle fa1lures: conta1nment spray pump, NSIV and FIV were assumed 1n the split break analyses. Thus, the peak values illustrated 1n Figure ll are conservative due to the multiple failures.

Figure 9 contains the containment pressure and temperature response for the 4.37 sq. ft. HZP double-ended rupture. Note that the HZP case proved to be more 11m1ting than the HFP case analyzed. This is primarily due to the large mass of water in the steam generator under HZP conditions which 1s available for discharge through the break. For this particular case analyses were performed which examined the consequences of two single failures: a single containment spray pump fa1lure, and an auxiliary feedwater runout failure.

The case presented in Figure 9 represents the containment spray pump failure.

This case was analyzed assuming a BAST boron concentration of 20000 ppm.

Figure 12 contains the mass and energy release rates for this case.

Figure 10 shows the containment pressure and temperature response for the 1.4 sq. ft. HFP double-ended rupture. Note that the peak pressure and temperature are significantly lower for the l.4 sq. ft. break than for the 4.37 sq. ft.

break. This -is due to the smaller break area which reduces the blowdown mass and energy release rate, this in turn results in a lower peak containment pressure and. temperature than the 4 .37 sq. ft. case. Due to the significant margin available to the containment pressure design limit only the containment spray pump failure was considered for the 1.4 sq. ft. cases. This case was analyzed assuming a BAST boron concentration of 6000 ppm.

Figure ll contains the containment pressure and temperature response profiles for the 0.6 sq. ft. HFP spl1t rupture. As discussed above this case contains three single failures: a containment spray pump failure, a main steam 1solation valve failure, and a feedwater isolation valve fa1lure. This case was analyzed assuming a BAST boron concentration of 6000 ppm.

The large break mass and energy calculations were proven to be the limiting cases because of the higher pressures reached. The temperatures and pressures reached 1n the large breaks with the assumed BAST concentration of 20000 ppm fall below the containment design limits.

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~ 0 Therefore, from a mass and energy point of view for the cases analyzed, it does not appear possible to reduce the BAST boron concentration below the current value of 20000 ppm due to the lack of significant available margin to the peak containment pressure limit of 60 psig.

A sensitivity study was performed to determine the impact of superheat for the RGE steamline break containment analysis. This sensitivity was performed on the limiting pressure case, 4.37 sq. ft. double-ended rupture at hot zero power, utilizing updated mass and energy releases modeling superheat characteristics. The results from this case revealed no diversion from the results of the non-superheat case.

CONCLUSIONS Plant specific analyses have been performed for the R. E. Ginna steamline break transients and have shown that while the current boron concentration of 20000 ppm will ensure that the peak containment pressure limit of 60 psig is not exceeded, there is not a sufficient amount .of margin to the containment pressure limit to allow a reduction in the Boric Acid Storage Tank boron concentration requirement.

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TABLE 1 TIME SEQUENCE OF EVENTS Case Event Time (seconds)

Steamline ruptures Pressur izer empt1es Critical1ty atta1ned 22 Boron enters core 45 Safety valve fails open Pressurizer empties 93 Low pressurizer pressure SI setpoint reached 99 Boron enters core 183 8709Q:10/051585

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4.37 FT FULL-OOUBLE-ENDED BREAK Power level Sin le Failure Offsite Power 102Ã Containment Spray Pump Avai1abl e Containment Spr ay Pump Availabl e 0$ Auxiliary Water Runout Availabl e 8709Q:10/051585 12

TABLE 3 4.37 FT OOUBLE-ENDED BREAK Power Level Sin le Failure -Offsite Power 102$ Containment Spray Pump Available 0$ Containment Spray Pump Available 8709Q:10/051585 13

TABLE 4 SPLIT BREAK THAT WILL NEITHER GENERATE A PRIHARY STEAHLINE ISOLATION SIGNAL NOR RESULT IN ENTRAINHENT S~ll i1 Offsite Power U

102'A 0.6 ft Containment Spray Pump Avai 1 abl e HSIV (Hain Steam Isolation Valve)

FIV (Feedwater Isolation Valve) 0$ 0.3 ft Containment Spray Pump Avail abl e HSIV FIV 8709 Q: 1 D/051 585 14

TABLE 5 ASSUHPTIONS FOR CONTAINMENT ANALYSIS Refueling water temperature ( F) 80 Initial Containment Temperature ('F) 120 Initial pressure (psia) 15.7 Initial relative humidity (X) 30 6

Net free volume (ft ) 1.00 x 10 Saf eguard System Number of fan coolers Pressure set point (psig)

Delay time (sec) 32 Number of spray pumps Maximum spray flow (gpm) 1200.

Pressure set point (psig) 33.5 Delay time (sec) 35.5 8709Q:1D/051 585 15

TABLE 7 FAN COOLER HEAT REMOVAL Containment RCFC Heat Removal/Fan Cooler Temp ('F) Btu/sec 200. 4416. 67 210, 4833.3 220. 5750.0 230. 7166.67 240. 8500.0 250. 9583.3 260. 10583.2 270. 11583.3 280. 12500.0 290. 13416.7 300. 14083.3 8709(}:10/051585 16

TABLE 6 PASS I V) HEAT SINKS Heat Transfer Thickness Wall Description Area (ft2

) Hater ial (ft)

l. Insulated portion of dome 36181.0 Insulation 0.1042 and containment wall steel 0.03125 Concrete 2.5

2. Uninsulated portion of dome 12474.0 Concrete 2.5 steel 0.03125

3. Basement floor 7955.0 Concrete 2.0 Steel 0.03125 Concrete 2.0

4. Walls of sump in basement 2342.0 Concrete 5.0 floor Steel 0.03125 Concrete 3.5

5. Floor of sump 297 ' Concrete 2.0 Steel 0.03125 Concrete 2.0

6. Inside of refueling cavity 3800.0 Stainless Steel 0.020833 Concrete 2.5

TABLE 6 (Continued)

PASSIVE HEAT SINKS Heat Transfer Thickness Wall Description Area (ft2 ) Material (ft)

7. Bottom of refueling cavity 1117.0 Stainless Steel 0.020833 Concrete 2.5

8. Area on outside of refueling 5952.0 Concrete 2.5 cavity walls

9. Area inside of loop and steam 12463.0 Concrete 2.5 generator compartment"

10. Floor area intermediate level* 6170.0 Concrete 0.5

11. Operating floor* 6540.0 Concrete 2.0

12. 1 1/2" thick I-beam"* 3151.0 Steel 0.125

13. 1" thick I-beam** 5016.0 Steel 0.0833

14. 1/2" thick I-beam 8138.0 Steel 0.04167

15. Cylindrical supports for 430.0 Steel 0.04167 S.G. and HCP's

TABLE 6 (Continued)

PASSIVE HEAT SINKS Heat Transfer Thickness Wall Oescription Area (ft2 ) Haterial (ft)

16. Plant crane rectangular 5756.0 Steel 0.0625 support columns

17. Beams used for crane 6023.0 Steel 0.125 structure**

18. Structure on operating 2622.0 Concrete 2.0 floor

19. Grating, stairs, 7000.0 Steel 0.0104 misc. steels

• Both sides exposed, valve represents area for one side.

• • Both sides exposed, valve represents area for both sides.

Thermo h sical Pro erties of Containment Heat Sinks Thermal Conductivity Volumetric Heat Capacity Haterial Btu/hr-ft-'F Btu/ft3-'F Insulation 0.0208 2.0

.Steel 28.0 58.8 Concrete 0.9 32.9

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R. E. GINNA BAST CONCENTRATION Figure 1 4.37ft2 Steamline Break One Loop in Service REDUCTION STUDY

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R.'. GINNA BAST CONCENTRATION Figure 2 REDUCTION STUDY 4.37ft Steamline Break One Loop in Service

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R. E. GINNA BAST CONCENTRATION Figure 6 Failed Safety Valve One Loop in Service REDUCTION STUDY

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R. E. GINNA BAST CONCENTRATION Figure 8 Failed Safety Valve One Loop in Service REDUCTION STUDY

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R. E. GINNA Figure 9 BAST CONCENTRATION REDUCTION STUDY 2

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R. f. GINNA Figure 10 BAST CONCf NTRATION 2 1.4 Ft Per Steamline Break 102~ po><<r Rf DUCTION STUDY

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R. E. GINNA Figure 11 BAST CONCENTRATION REDUCTION STUDY 0.6 Ft 2 Split Steamline Break 102% Power

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R. E. GINNA Figure 12 BAST CONCENTRATION 4,37 Ft 2 Per Steambreak 0% Power REDUCTION STUDY Mass and Energy Release Flow Rates

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