ML19269D608

Forwards Results of Analysis Performed by GE Re Suppression Pool Temp Transients.Will Revise Station Operating Procedure by 790615 to Depressurize Reactor Vessel

ML19269D608
Person / Time
Site:	Cooper
Issue date:	05/29/1979
From:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
To:	Ippolito T Office of Nuclear Reactor Regulation
References
NUDOCS 7906050272
Download: ML19269D608 (44)

Text

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CENER AL OFFIC E P. O. BOX (99. COLUMBUS. NEBR ASKA 68601 Nebraska Publ.ic Power D.

is tn.

ct rete ~o~ E non s......i May 29, 1979 Director, Nuclear Reactor Regulation Attention: Mr. Thomas A. Ippolito, Chief Operating Reactors Branch No. 3 Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Suppression Pool Temperature Transients Cooper Nuclear Station NRC Docket No. 50-298, DPR-46

References:

1. Letter, G. Lear (NRC) to Nebraska Public Power District,

" Suppression Pool Temperature Transients",

dated December 13, 1977.

2. Letter, J. Pilant (NPPD) to T. Ippolito (NRC),

Same Subject, dated August 7, 1978.

3. Letter, J. Pilant (NPPD) to T. Ippolito (NRC),

Same Subject, dated January 5, 1979.

4. Letter, J. Pilant (NPPD) to V. Stello (NRC),

" Schedule for the Implementation and Resolution of the Mark I Containment Long Term Erogram",

dated May 7, 1979.

Dear Mr. Ippolito:

Reference 1 requested information related to Suppression Pool Temperature Transients at Cooper Nuclear Station. The District's response to this request (Refererxes 2 and 3) coprained a commitment to supply the results of a Plant Unique Analysis of specific parameters as a function of time for certain depressurization transients. Attached please find the results of this analysis performed by General Electric Company entitled " Cooper Nuclear Station Suppression Pool Temperature Response".

Per the justifications discussed in section 5.2 of the attached, Nebraska Public Power District is not requesting a change to Technical Specification 3.7.A.l.f at this time since the ramshead discharge devices will be replaced with T-quencher discharge assemblies during the next refueling outage (Reference 4). However, the District will revise Station Operating Pro-cedures by June 15, 1979 to indicate that during reactor isolation con-ditions, if the pool temperature reaches 1200F, the reactor pressure vessel will be depressurized at a rate such that the reactor pressure is less than 150 psig when the pool temperature reaches 150 F.

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Thomas A. Ippolito May 29,1979 Page 2 If you have any questions, or require any additional information, please do not hesitate to contact me.

Very truly youts,

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1. ,4%

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6. W y

/ ,/JYy L'Pilant Director of Licensing and Quality Assurance JDW/cmk Attachment

COOPER NUCLEAR STATICN SUPPRESSION PCOL TEMPERATURE RESPONSE 2261 224

TABLE OF CONTENTS Pace

1.0 INTRODUCTION

1 2.0

SUMMARY

2 3.0 MODEL DESCRIPTION, INITIAL CONDITIONS AND ASSUMPTIONS 3 3.1 Model Description 3 3.2 Initial Conditions 4 3.3 Assumptions 5 4.0 ANALYSIS 7 4.1 Safety / Relief Valve Discharge During Non-LOCA Events 7 4.1.1 Event 1: Stuck-Open Relief Valve From Power 7 4.1.1.1 Event Description 7 4.1.1.2 Analysis Results 8 4.1.2 Event 2: Stuck-Open Relief Valve From Hot Standby 10 4.1.2.1 Event Description 10 4.1.2.1 Analysis Results 11 4.1.3 Event 3: Controlled RPV Depressurization 14 From Hot Stancby 4.1.3.1 Event Description 14 4.1.3.2 Analysis Results 15 4.2 Safety / Relief Valve Discharge During LOCA Event 17 4.2.1 Small Line Break with ADS 17 4.2.1.1 Event Description 17 4.2.1.2 Analysis Results is

5.0 CONCLUSION

S 19 5.1 Conclusions 19 5.2 Justification for Exceptions to the Technical 20 Specifications 5.3 Justification for Exceptions to Reference (1) 22 Analysis Assumptions

6.0 REFERENCES

25 Appendix -

Event Descriptions 38 2261 225

LIST OF TABLES Page Table 1. Summary of Results 26 2261 226 iii

LIST OF ILLUSTRATIONS Ficure Title Pace

1. Model Schematic of Reactor and Containment 27 System - Non-LOCA Event

2. Stuck-Open Relief Valve at Power - 2 RHR Loops 28

3. Stuck-Open Relief Valve at Power - 1 RHR Loop 29 4 Stuck-Open Relief Valve From Isolated Hot 30 Standby - 2 RHR Loops

5. Stuck-Open Relief Vavle From Isolated Hot 31 Stancby - 1 RHR Loop

6. Controlled Depressurization from Hot Standby at 32 200 F/hr - 2 RHR Loops

7. Controlled Depressurization From Hot Standby at 33 300 F/hr - 2 RHR Loops

8. Controlled Depressurization From Hot Standby at 34 300 F/hr - 1 RHR Loop

9. Controlled Depressuri:ation From Hot Standby at 35 400 F/hr - 1 RHR Loop

10. Liquid SBA With A05 36

11. Ramshead Discharge Mass Flux vs. Reactor Vessel 37 Pressure 2261 227 iv

1.0. INTRODUCTION The Cooper Nuclear Station (CNS) takes advantage of the large thermal capacitance of the suppression pool during plant transients requiring safety / relief valve (SRV) actuation. The discharged steam is piped from the reactor pressure vessel (RPV) through the main steam and SRV discharge lines to the suppression pool, where it condenses, resulting in a tempera-ture increase of the pool water, and an increase in the containment pres-sure. Most transients that result in relief valve actuations are of very short duration and have a small effect on the suppression pool temperature.

However, certain single and multiple failure events can be postulated wnich have the potential to discharge _ steam into the sucpression pool for an extended period of time, significantly increasing the pool temperature.

This may result in a situation where the suppression pool temperature and the ramshead discharge mass flux are such that the condensation stability limit may be approached. Stable condensation is expected for ramsheads if the suppression pool bulk temperature does not exceed 150 F when the ramshead mass flux (G) is greater than 40 lbm/sec-ft2 . The condensation phenomena is determined by the local temperature in the vicinity of the discnarge device, whereas the calculations assume a bulk temperature.

The bases for the 10 F temperature difference between bulk and local temperature and the condensation phenomena are given in Reference 2.

The Nuclear Regulatory Commission requested (Reference 1) that four postulated events, which have the potential to discharge steam into the suppression pool for an extended period of time, be analyzed to demon-strate that the condensation stability limit wil' not be exceeded at the Cooper Nuclear Station. These four events are:

1. Stuck open safety-relief valve (SRV) during power operation.

2. Stuck-open SRV during isolated hot standby.

2261 228 LMZ:bjw:cas/53I 1

3. Controlled RPV depressurization from isolated hot standby.

4. Automatic Depressurization System (A05) activation following a small line break.

This document presents the results of the analysis of these four events for the Cooper Nuclear Station.

2.0

SUMMARY

The events mentioned in Section 1.0 were analyzed using licensing basis safety analysis values, and assume that the CNS Technical Specifications are not violated except as indicated in Assumption 16 in Section 3.3.

Section 3.0 describes the calculational model, initial conditions, and assumptions that were used in the analyses. Section 4.0 presents the event descriptions and the complete analysis results. Section 5.0 presents the conclusions of the analysis.

The results of the analyses are summarized in Table 1. For each of the four events analyzed, values of the maximum suppression pool temperatures when the ramshead discharge mass flux (G) is greater than the critical value of 40 lbm/ft:-sec are given. Values of the discharge mass flux when the suppression pool temperature reaches the critical value of 150 F are also given. The table also shows the number of RHR loop (s) that were assumed operaole and the number of SRV's that were opened to depressurize the reactor. The analysis results show that the condensation stability limit will not be exceeded for these events at the Cooper Nuclear Station.

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3.0 MODEL DESCRIPTION, INITIAL CONDITIONS AND ASSUMPTIONS 3.1 Model Description 3.1.1 Non-LOCA Events To solve the transient response of the reactor vessel and suppression pool temperature due to the postulated events, a coupled reactor vessel and suppression pool thermodynamics model was used. The model is based on the principles of conservation of mass and energy and accounts for any possible flow to and from the reactor vessel and the suppression pool as shown in Figure 1.

The model incorporates a control volume, which includes the reactor pressure vessel (RPV) and the suppression pool. The RPV model is capable of tracking the reactor vessel water level and having a rate of change of temperature or pressure imposed on it. The various modes of operation of the residual heat removal (RHR) system have been sb;'at ed as well as the relief valves, HPCI, liCIC and f aadner functions. % *0 del also simulates system setpoints (automatic ano manual), operator actions and accepts as input the specific plant geometry and equipment capability.

3.1.2 Small Break Accident Model In the Small Break Accident analysis, the mass and energy conservation laws are applied to a control volume which includes all of the reactor vessel contents and its walls.

This control volume is subjected to the boundary conditions of decay heat input. The break and the safety-relief valve flow rates and the associated fluid enthalpies are derived from the state of fluid in the control volume undergoing the transient and the specified flow areas and location.

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The time dependent break and safety-relief valve mass and energy flows are then input to another control volume containing the suppression pool. The pool temperature transient is obtained using the energy and mass balance equations on the suppression pool.

3.2 Initial Conditions The following initial conditions were assumed in all of the analyses:

1. Operation at licensing bases safety analysis steam flow conditions.

(CNS = 105% NBR steam flow.)

2. Maximum condensate storage tank water temperature. (CNS =

90*F.)

3. Maximum RHR heat exchanger service water temperature. (CNS =

90 F.) .

4. Suppression pool temperature at normal power operation Technical Specification limit (Top). (CNS = 90*F.)

5. Minimum Tech Spec suppression pool water volume. (CNS = 87,650 cu. ft.)

6. Drywell air pressure at maximum of normal operating band. (CNS

= 1.1 psig.)

7. Drywell air temperature at normal drywell average. (CNS =

135 F,)

8. Wetwell air pressure at maximum of normal operating band. (CNS

= -0.4 psig.)

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3.3 Assumptions The following assumptions were used in all of the analyses, except as noted in the individual event descriptions in Section 4.0 and in the Appendix.

1. Normal auxiliary power is available.

2. Normal automatic operation of the plant auxiliary systems (HPCI, RCIC, Core Spray, LPCI and ADS).

3. Control Rod Drive (CRD) flow maintained constant.

4. Outy of RHR heat exchangers based on maximum observed equilibrium crud buildup.

5. SRV capacities at 122.5% of ASME rated.

6. Licensed decay heat curve (May-Witt) for containment analysis (adjusted to account for delay between scram and isolation).

9. Turbine-driven feedwater pumps were assumed to be available until MSIV closure occurs. Additional coast down flow was included.

10. Linear reduction of main steam flow rate from its initial flow rate at the start of reactor isolation to zero flow in 3.5 seconds (MSIV closure time).

11. In calculating the overall heat transfer coefficient of the vessel wall and internal structures, it is assumed that the heat transfer is dominated by conduction.

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12. The heat transfer area of the reactor internals is obtained by assuming that they have the same metal thickness as that of the vessel wall, which is assumed to be 0.333 ft. uniformly.

13. The control volume of the reactor includes the reactor vessel, the recirculation lines, and the steam lines from the vessel to the inboard main steam isolation valves (MSIV).

14. The initial water level in the reactor vessel is calcu-lated based on the assumption that the voids in the two phase region collapse. Therefore, the ECCS volumes are based on the total liquid volume of the reactor vessel, the feedwater lines and the recircula-tion lines combined.

15. The specific heat of th'e reactor vessel and the internals is assumed to be 0.123 btu /lbm/ F. The metal density is assumed to be 490 lbm/ft3.

16. The Cooper Nuclear Station Technical Specifications are not violated during the events, except that:

a) depressurization to avoid condensation instablity is not limited to the normal cooldown rate (100 F/hr) specified in Technical Specification 3.7.A.1.f; and b) since the SRV capacities are assumed to be 122.5%

of ASME rated, the reactor pressure corresponding to the ramshead critical SRV discharge mass flux of 40 lbm/sec-ft2 is 150 psig, rather than the value of 200 psig given in Technical Specification 3.7.A.1.f.

These assumptions are justified in Section 5.2.

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17. Operator actions are based on normal operator action times during the given event. ,

I

18. A stuck-open relief valve can be detected and the corresponding ramshead within the torus identified. I I

19. Additional safety / relief valves are manually opened, as necessary, to depressurize the reactor.

.0 ANALYSIS A detailed description of the event sequences and analysis results for the four events which have been analyzed is presented below. A chrono-logical summary of the event secuences is presented in the Appendix.

4.1 Safety Relief Valve Discharge During Non-LOCA Events Tne first classification of SRV discharge events considered are those whicn occur under non-LOCA (i.e., non-accicent), conditions.

These events consider a stuck-open relief valve (50RV) condition and the manual operation of additional SRV's to depressurize tne vessel.

4.1.1 Event 1: Stuck-Open Relief Valve ' rom Power 4.1.1.1 Event Description For this event, the reactor is initially operating at the power level corresponding to 105% of rated steam flow (2487 Nt) and the suppression pool temperature is at 90 F. An SRV is postulated to stick fully open at this point in time (time zero of the transient analyzed). The plant operator is I alerted that an SRV has opened by a drop in power output, as well as other plant parameter indications. ,

The plant operator attempts to close the SRV but is 7

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unsuccessful. Therefore, the operator initiates a reactor scram at the time when the pool temperature has reached the maximum value permitted during power operation (110 F). As a result of the scram, the voids in the reactor collapse and the reactor water level drops momentarily below Level 2 of the reactor protection system causing the main steam isolation valves (MSIV) to close. At 10.5 seconds after scram the MSIV's are assumed to be fully closed. During this time period (i.e., from time O to the scram time + 10.5 seconds) full feedwater flow into the reactor vessel is assumed. At the time the MSIV's are fully closed, the feedwater turbine-driven pump is conservatively assumed to coastdown and drop to zero flow linearly in 7 seconds.

The pool temperature rises due to the steam discharge from the 50RV into the suppression pool. Cooling of the suppression pool, utilizing the RHR heat exchangers in the pool cooling mode, is initiated three (3) minutes after the pool temperature reaches the maxi-mum value permitted during normal power operation (90 F). At 10 minutes after the occurrence of the 50RV the plant operator is assumed to open addi-tional SRV's to depressurize the reactor vessel such that, should the bulk pool temperature reach 150 F, the ramshead mass flux at the discharge point is below 40 lbm/sec ft 2, 4.1.1.2 Analysis Results The analysis was performed assuming:

1) Both RHR loops are available - Event 1(a).

2) Only one RHR loop is available - Event 1(b).

8 2261 235

The results are shown in Figures 2 and 3, and are summarized in Table 1.

The vessel depressurization and the resulting sup-pression pool temperature increase with both RHR loops available is shown in Figure 2. Time zero in the figure corresponds to the time of reactor pres-sure vessel isolation. The stuck open relief valve initially opened approximately 400 seconds prior to the time of isolation.

Following isolation, the vessel pressure increases slightly to a maximum value of 1050 psia at 15 seconcs and then begins to gradually decrease due to the 50RV. At 95 seconds the vessel pressure begins to drop sharply due to the automatic initiation of the HPCI and RCIC systems. This large pressure drco during make up water injection results from the fact that these systems not only extract large vciumes of steam from the RPV to power their operation but they also quench the vessel fluic through their injection of cool water. At approximately 200 seconds after the time of RPV isolation one SRV is manually opened by the operator. The flow from tnis manually ope.1ed SRV further increases the already rapid depressuri-zation of the RPV due to the stuck open valve. At about 435 seconds the HPCI and RCIC are shut off and the vessel experiences a gracual repressurization.

At about 662 seconds the plant operator manually initiates the HPCI and RCIC system, to make c the vessel fluid being lost through the two open SRVs, and the vessel pressure again drops rapidly. The vessel depressurization transient subsequently continues with periods of vessel pressure decreases l

during make up water inflow, and slight repressuri- t zation between these injections of make up water.

At about 1299 seconds the vessel pressure drops 9

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, below, and subsequently remains below, 165 psia, which corresponds to the critical SRV discharge mass

, flux of 40 lbm/sec-ft 2, The manually opened SRV is closed by the operator after the vessel pressure drops below 165 psia and

' before it reaches 100 psia.

This is to reduce the vessel cooldown rate ( F/nr.).

The vessel will continue to depressurize due to the one stuck-open relief valve. However, it is possible that the one SORV may close at this lower vessel pres-sure if the operator attempts to close the valve.

The analysis results for the SORV at full power event with only one RHR loop available (Figure 3) are identical to those of the case with two RHR loop available up to the time that the RHR loops are turned on in the pool cooling mode. This occurs at 3 minutes after the time that the stuck open relief valve initially opened.

With only one RHR heat exchanger available, the pool temperature increases more rapidli than when both RHR loops are available.

With two RHR luops available the pool temperature reaches 140 F at the time that the vesse'l pressure drops below the critical value of 165 psia. At the critical vessel pressure, the pool temperature is approximately 2 F higher when only one RHR loop is available than when both loops are available (Table 1).

Condensation pool instability is avoided by a pool temperature margin of 8 to 10 F, depending on whether one or two RHR loops are assumed to be available.

Additional pool temperature margin could be obtained

. by manually opening more than one SRV to depressurize the reactor vessel. However, the analysis results show that opening only one SRV is sufficient to avoid condensation instability, and also minimizes the con-tainment loadings and the thermal duty on the vessel due to the resulting depressurization rate.

1g 2261 237

4.1.2 Event 2: Stuck-Open Relief Valve From Hot Standby 4.1.2.1 Event Descriotion In Event 2 the initial pool temperature is 90 F and the reactor power level is initially at 105% of rated steam flow (2487 Mwt) at the time of reactor scram. A transient occurs which results in reactor scram and isolation. The reactor is held in an extended hot standby condition during which the reactor pressure is maintained at 920 psig by operating the SRV's intermittently. The steam flow from these SRV actuations causes the suppression pool temperature to rise. Cooling of the suppression pool, utilizing the RHR heat exchangers in the pool cooling mode, is initiated three (3) minutes after the pool temperature reaches the maximum value permitted during normal power operation (90*F). When the pool temperature is at 120*F (the maximum permissible pool temperature when the reactor is isolated from the main heat sink), the plant operator begins the required reactor depressurization. At this point in the transient, a relief valve is postulated to stick open. Additional SRVs are therefore opened and maintained opened to depressurize the reactor vessel so that the ramshead mass flux is below the critical value of 40 lbm/sec-ft2 when the bulk pool temperature reaches 150 F.

4.1.3.2 Analysis Results For Event 2, the analysis was performed assuming:

1) Both RHR loops are available - Event 2(a).

2) Only one RHR loop is available - Event 2(b).

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The results are shown in Figures 4 and 5, and are summarized in Table 1.

The vessel pressure transient for Events 2(a) and 2(b) are very similar, the major difference being the time that it takes for the pool temperature to increase to 120 F. When the pcol temperature reaches 120 F, the operator begins the vessel depressurization as required by the CNS Technical Specification. The suppression pool temperature responses are also similar, varying primarily by the difference in pool cooling capability assumed during the events.

With both RHR loops available, the 50RV from Hot Standby Event is as follows. At time zero the reactor is scrammed and isolated. Following the scram and isolation, the vessel pressure (Figure 4) increases to the lowest relief valve set point and is maintained at this value by the automatic opening of the SRVs. At about 32 seconds into the transient the HPCI and RCIC are initiated automatically at the low water Level 2 setpoint. This injection of cooler water from the condensate storage tank into the vessel causes the vessel pressure to drop sharply to a minimum value of about 690 psia at the time the plant operator shuts off the make-up flow. Once the high pressure injection flow is stopped, the RPV repressurizes due to decay heating.

At about 520 seconds the plant operator begins manually operating the SRVs to maintain the vessel pressure at 920 psig. As a result of these manual actuations of the SRVs fluid is lost from the RPV, and the plant operator must intermittently operate 12 2261 239

the HPCI and RCIC to maintain vessel water level.

During these periods of make-up water injection the RPV pressure drops but again repressurizes to 920 psig after the flow is discontinued.

As shown in Figure 4, the pool temperature increases rapidly to about 96 F during the period of automatic relief valve operation. Following this rapid increase, the pool temperature gradually increases up to 120 F (i.e., the limit for isolated hot standby operation) as a result of the steam flow from the manually operated SRV and also from discharge steam from the HPCI and RCIC turbines.

When the suppression pool temperature has reacned 120*F (at about 3050 seconds after isolation), the plant operator begins a controlled depressurization of the reactor pressure vessel (RPV). A safety-relief valve is assumed to stick open at this time. The plant operator immediately opens, and maintains open, one additional SRV. Therefore, starting at 3050 seconds the RPV experiences a rapid depressuri-zation due to the steam flow through the two open SRVs (i.e., the 50RV plus one (1) additional valve).

At about 3080 seconds the automatic initiation of the HPCI and RCIC systems further increase the rate of depressurization. At about 3260 seconds the .

make-up is discontinued and the vessel depressuriza-tion decreases. From this time on the vessel pres-sure gradually decreases due to the continuous flow through the two open SRVs and the intermittent operation of the HPCI anc' RCIC systems. At approxi-mately 3935 seconds the vessel pressure drops below, and subsequently remains below, the critical vessel pressure of 165 psia. The pool temperature at the 13 2261 240

time the vessel pressure drops below the critical

. value (165 psia) is 143 F.

The assumption that only one RHR loop is available during an SORV from Hot Standby (Event 2(b) ) 'does not conform to the plant- licensing basis. Also, the operator would not go into hot shutdown

• with only one RHR loop available. The following analysis of this event is presented for information only.

The above discussion of Event 2(a) also applies to Event 2(b), with the following exceptions. As shown in Figure 5, if only one RHR loop is available various events occur earlier in time. The vessel depressurization, which begins when the pool tempera-ture reaches 120 F, is initiated at 2903 seconds after isalation. The critical vessel pressure of 165 psia is reached at 3770 seconds. The suppressien pool temperature at the time the vessel pressure drops below 155 psia is 144*F. The manually opened SRV is closed by the operator after the vessel pressure drops below 165 psia and before the pressure reaches 100 psia.

Additional pool temperature margin could be obtained by manually opening more than one SR'l to depressurize the reactor vessel. However, the analysis results show that opening only one SRV is sufficient to avoid the condensation instability threshold during this event.

• Hot shutdown is similar to het standby, except that the reactor is scrarred, instead of at approximately 5". power.

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3 Event 3: Controlled RPV Depressurization from Hot Standby 4.1.3.1 Event Descriotion i

In this event, the reactor is operating at 105% of rated steam flow when an isolation and scram occurs.

The poci temperature is assumed to De at the maximu:r technical specifications value (90 F). The reactor is put in Hot Standby. The operator initiates actions to place the RHR iceps in the pcci cooling mode, as required by the technical specifications.

The RHR loops are on for pool cooling three (3) minutes later. The operator actuates SRV's to maintain the reactor pressure at 920 psig. One of the actuated SRV's is assumed to stick open when the pool temperature reaches 120 F. When the pool temperature reacnes 120*F (the maximum permissible peel temperature when the reacter is isolated from the main heat sink),

the plant c::erator begins tne recuired reacter depressurization. Again, the rate of depressuri:ation required is determined by the criteria that the ramshead discnarge mass flux must be below the critical value of 40 lbm/sec-ft2 when the bulk suppression pool temperature reaches 150 F.

4.1.3.2 Analysis Results The analysis was performed assuming:

1) Scth RHR loops are available - Event 3(a).

2) Only one RHR loop is available - Event 3(b).

The results are shown in Figures 5 through 9, and are summarized in Tacle 1. Figures 6 and 7 show the -

i vessel pressure and pecl temperature for Event 3(a) 2261 242 15

for operator controlled cooldown rates of 200 F/hr and 300*F/hr, respectively. Similarly, Figures 8 and 9 show the vessel pressure and pool temperature for Event 3(b) for operator controlled cocidown rates of 300*F/hr and 400*F/hr, respectively. The assumption that only one RHR loop is available (Event 3(b)) does not conform to the plant licensing basis. Also, the operator would not go into hot shutdown with only one RHR loop available. The analysis of this event is presented for information only.

The four transients shown in :ne figures all share common general characteristics. The particular differences between the individual transient responses are in timing and magnituce, which are expected, given the different rates of cocidown and the number of RHR heat exchangers which are assumed availacle.

Therefore, the following discussion is considered to be applicable to all four transients.

For tne initial period of isolated het standby, the RPV pressure and suppression pool temperature transients are similar to those previously discussed for the 50RV at Isolated Hot Standby transient (Event 2).

However, for this transient the depressurization is conducted at a controlled rate and takes longer to depressurize the vessel to below the critical 165 psia pressure. The saw-tooth c.ppearance of the figures during the period of controllec depressurization result from the intermittent operation of the HPCI and RCIC systems by the plant operator. During i 5

these periods of make-up water injection, the vessel pressure drops faster than the specified cocidown rate and, consequently, no depressurization through ,

additional SRVs is required.

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Referring to Table 1, it is seen that with both RHR heat exchangers available (Event 3(a)), the vessel cooldown rate required to meet the 150 F pool tempera-ture limit is between 200 F/hr and 300*F/hr.

The maximum pool temperature when the vessel pressure drops belcw the critical vessel pressure of 165 psia is 151*F for a 200*F/hr cooldown and 148*F for a 300 F/hr coolJown. With only one RHR heat exchanger available, (Event 3(b)), a vessel cooldown rate slightly less than 400 F/hr is required. The maximum suppression pool temperature when the vessel pressure drops below 165 psia is 151 F for the 300*F/hr cooldown and 149 F for the 400*F/hr cooldown. For both Events 3(a) and 3(b) the operator reduces the vessel cooldown rate to 100*F/hr after the vessel pressure drops to 165 psia and before it reaches 100 psia.

4.2 Safety Relief Valve Discharge During LOCA Event "

4.2.1 Event 4: Small Line Break with ADS

4. 2.1.1 Event Description This event assumes that a small break occurs in the recirculation system inlet line. The size of this postulated liquid break is chosen such that the highest suppression pool temperature results when the ramshead discharge mass flux is 2 40 lbm/sec-ft2 ,

In this event, the reactor is assumed to be auto-matically scranred a high drywell pressure signal (2 psig), which occ .s at approximately 6 seconds af ter the small line breaks. Isolation is subsequently initiated automatically on icw water Level 2. The assumed single active failure is the failure of the

' High Pressure Coolant Injection (HPCI) system.

Therefore, the vessel pressure after the small' liquid break remains early ccastant at the lower 2261 244 17

setpoint of the relief valves (1095 psia). Since no safety grade high pressure make-up system is available (credit is not taken for RCIC since it is not safety grade), the vessel water level continues to drop as mass is lost through the break and through the SRVs.

Also, credit is not taken for the feedwater system, since it is not safety grade. At 120 seconds after the vessel water level reaches Level 1 of the Reactor Protection System (RPS), the ADS is automatically activated and rapidly depressurizes the vessel to well below the critical pressure required to produce a discharge mass flux at the ramshead of 40 lbm/sec-f t2 ,

4.2.1.2 Analysis Results To define the limiting break size, both a small line break area of 0.05 ft2 and an intermediate line break with a break area of 0.1 ft2 were considered.

The results of these analyses si.Jwed that the 0.05 ft2 break yields a slightly higher pool temperature (139 F) than does the 0.1 ft 2 break (13S F) when the critical discharge mass flux of 40 lbm/sec-ft 2 is reached. Therefore, the small line break was found to be the limiting case. The results of the analysis for the 0.05 ft2 break area are presented in Figure 10.

The SBA with ADS transient sequence is as follows.

At time zero the reactor is assumed to be scrammed on a high drywell pressure signal, which occurs at approximately 6 seconds after the line breaks.

Following scram, the reactor isolates automatically on low water Level 2. Following isolation, the vessel pressure peaks briefly and then decraases and oscillates about the lowest relief valve set point

^

until the time when the Automatic Depressurization 2261 245 18

System (ADS) actuates at approximately 317 seconds after the time of scram. The suppression pool temperature, which is conservatively taken to be 93 F at the time of scram, rises gradually during this period as both the break flow and relief valve flow enter the pool. Upon actuation of the five available ADS valves, the vessel experiences a rapic depressurization and the suppression pool experiences a corresponding temperature increase. At approximately 467 seconds after scram, the low pressure Emergency Core Cooling Systems (ECCS) automatically activate and slightly increase the rate of vessel depres-surization as the cool ECCS water is pumped into the vessel. The pool temperature reaches 139*F at thi time when the vessel pressure drops below 165 psia.

The vessel pressure subsequently remains below 165 psia.

5.0 CONCLUSION

S 5.1 Conclusions The results of the analysis are summarized in Table 1. These results support the following conclusions:

1. Condensation instability is avoided in Event 1 (50RV during full power operation) by opening one acditional SRV at 10 minutes after the 50RV initially opens.

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2. Condensation instability is avoided in Event ^ (50RV during isolated hot standby) by opening one addit %nal SRV when the pool temperature reaches 120 F. This is in conformance with the current CNS Technical Specifications, which require that during reactor isolation conditions the RPV shall be depres-surized if the pool temperature reaches 120 F.

3. Condensation instability will not occur in Event 3 (Controlled Depressurization from Isolated Hot Standby) when the RPV is depressurized at a cooldown rate of 300 F/hr when two RHR loops are available.

4 Condensation instability is avoic'ed in Even: 4 (Small Break Accident with ADS) without any operator action. The ADS system will automatically activate during this event and depressurize the RPV such that the condensation stability limit is not exceeded.

5.2 Justification for Eyceptions to the Technical Specifications The analysis of the events described in this document was conducted using licensing basis safety analysis values, and conformed to the current CNS Technical Specifications, except that: 1) the required cooldown rate exceeded the " normal cooldown rate" given in Technical Specification 3.7.A.1.f; and 2) the reactor pressure corresponding to the critical SRV discharge mass flux of 40 lbm/sec-ft: was assumed to be 150 psig, rather than the value of 200 osig given in Tecnnical Specification 3.7.A.1.f. The justification for the use of these exceptions to the Tecnnical Specification is as fcilows:

For the events analyzed, a cooldown rate in excess of the normal rate (100 F/hr) is required to avoid condensation instability in the pool. The most severe cooldown rate required by tne analysis to avoid condensation instability is 300*F/hr. This rate is requirec during the controlled depressurization from isolated hot stancby event.

20 2261 247

Howe" r, Technical Specification 3.7.A.1.f states that "during reacto'r isolation conditions, the reactor pressure vessel shall be depressurized to less than 200 psig at normal cooldown rates if the pool temperature reaches 120 F." Since the analysis of the 50RV from power event, the 50RV from isolated hot standby event, and the

, controlled RPV depressurization from isolated hot standby event all assume that the reactor is isolated, the Technical Specification limitation to normal cooldown rates is inconsistent with the cool-down rates required to avoid condensation instability.

The specification of a normal cooldown rate is more restrictive than has been shown to be acceptable for transient events. The CNS RPV was designed and constructed to allow given numbers of transient events to occur which produce cooldown rates greater than the normal rate. The most severe cooldown calculated in the present study (100 F change in reactor vessel temperature in 170 seconds) results from the mar.ual opening of one additional SRV during the SORV from isolated hot standby event'with only one RHR loop assumed to be available. This cooldown transient is approximately as severe as that shown on the CNS RPV Thermal Cycle Design Basis Diagram.

However, the CNS RPV Certified Stress Report treated a portion of the cycle as a step change of 198 F, which is significantly more severe than the cooldown rate calculated to avoid condensation instability for this event.

The requirement of a normal cooldown rate in Technical Specifi-cation 3.7. A.1.f is therefore unnecessarily restrictive. The allow-able cooldown rates during transient events, including those analyzed in the present study, should be based on the CNS RPV Thermal Cycle Design Basis Diagram. This justifies the use of cooldown rates greater than the normal rate in the present analysis.

The justification for the second exception to the Technical Speci-fications, the use of an RPV critical pressure of 150 psig, is based on SRV test data which indicates that actual SRV capacities may approach 122.5I, of ASME rated. .

4

. 21

The pressure of 150 psig is the reactor vessel pressure correspond-ing to the critical SRV discharge mass flux of 40 lbm/sec-ft2 when l the SRV capacities are assumed to be 122.5% of ASME rated (Figure 11). l This assumption was made in the present analysis because it conserva- l tively represents the actual SRV discharge mass flow rate into the pool. The A5ME rated capacity of a valve is less than the actual valve capacity. This is conservative for pressure vessel analyses, but is non-conservative for the present analysis, in which the pool temperature rise is dependent on the actual steam flow into the pool. This justifies the use of SRV capacities at 122.5% of ASME rated in the present analysis.

5.3 Justification for Exceptions to Reference (1) Analyses Assumptions Reference (1) requested that figures be provided which depict the reactor pressure, safety / relief valve (5RV) discnarge mass flux, anc suppression poci tulk temperature versus time for the following events whicn are based on current Technical Specification limits:

(a) Stuck open SRV during power operation assumi ng reactor scra at ten minutes after the suppression pool reaches a bulk pool temperature of 110 F and all RHR systems are operable.

(b) Same events as in (a) above with only one RHR train operable.

(c) Stuck-open SRV during hot standby assuming an initial 120*F bulk poci terature and only one RHR train opera::le.

(d) Automati, De;ressuri:ation System (A05) activate: following a smali tire break assuming an initial 120 F bulk :: col tem;:erature and only one RHR train operable.

(e) Primary system is isolated and depressurized at a rate of 100 F l

per hour with an initial 120 F bulk pool temperature and only I cne RH:. train operable.

2261 249 22

The following table shows the corresponcence between the Reference (1) events (acove) and tne present analysis:

Reference (1) Event Event No. (this document)

(a) 1(a)

(b) 1(b)

(c) 2(a), 2(c)

(d) 4 (e) 3(a), 3(o)

The ana!ysis presented in this cocument conforms to tne Reference (1) event scenarios, except where they were in violation of tne plant Technical Specifications. These excections are discussec below.

Event (a)

The analysis in this document assumec that the reactor is scramrec when the suppression pool reaches a bulk pool temperature of 110 F, rather than 10 minutes after the bulk pool temperature reacnes 110*:- Reactor scram at 110 F is based on Technical Specification 3.7.A.1.e:

The reactor shall be scrammed from any operating condition if the pool temperature reaches 110 F. Power operation shall not be resumed until the pool temperature is reduced below the normal power operation limit scecified in c. above.

The present analysis assumes that this technical specification is not violated. The pool temperature alarm notifies ne ope ator that the pool temperature is approaching 90 F. The pool tempera- I ture indicator provides the operator with tne water temperature f around the pool. '

i 2261 250 23

Event (b)

Same comments as for Event (a) above.

Event (c)

The analysis in this document assumes that the SRV sticks open during hot standby when the bulk pool temperature reaches 120*F.

This is in agreement with the Reference (1) request. However, the assumption that only one RHR loop is available does not conform to the plant licensing basis.

The current analysis of this event assumes that 1) an SRV fails in the stuck open position, and 2) the main condenser is unavailable as a heat sink. The assumption of an additional failure, such as one RHR loop, is highly improbable, and violates the plant licensing basis.

Also, if only one RHR loop was available, the operator would not go into hot shutdown. An analysis of this event, assuming that only one RHR loop is available, was performed. However, the results are presented for information only, and should not be used for licensing purposes.

Event (d)

The present analysis assumes that the bulk suppression pool temperature is 90 F, rather than 120 F, when the small line break accident occurs. This is in accordance with Technical Specifica-tion 3.7.A.1.c:

Maximum suppression pool temperature during normal power operation - 90 F.

2261 251 24

The pool temperature alarm will alert the operator that the pool temperature is approaching 90*F. The operator is required to initiate pool cooling before the pool temperature reaches 90 F. In the present analyses, it is assumed that the RHR loops are on for pool cooling at 3 minutes after the pool temperature reaches 90 F.

The present analysis was performed with both RHR loops available.

The assumption that only one RHR loop is available does not conform to the plant licensing basis, or to 10CFR50 Appendix K requirements.

Event (e)

The initial bulk pool temperature is assumed to be 90 F rather than 120*F for the reasons discussed in Event (d) above.

The assumption that only one RHR loop is available does not conform to the plant licensing basis. In addition, if only one RHR loop was available, the operator would not go into hot shutdown. An analysis was performed for only one RHR loop operable. However, the results are presented for information only, and should not be used for licensing purposes.

The present analysis did not assume that the vessel is depressurized at a rate of 100 F/hr, but rather calculated the controlled depressuri-zation rate which is required to avoid the condensation instability threshold. The justification for cooldown rates greater than 100 F/hr is given in Section 5.2

6.0 REFERENCES

1. Letter, G. Lear (NRC) to Nebraska Public Power District,

" Suppression Pool Temperature Transients," (Part A), dated December 1977.

2. " Steam Vent Clearing Phenomena and Structural Response of the BWR Torus (Mark I Containment)," NEDO-10859, April 1973.

25 2261 252

I Alllt 1

SUMMARY

OF REStlLIS CNS POOL IIMPERATURE RESPONSE Discharge Mass

• Haximum Pool Event Event No. of SRV's No. of flux (C) at Temp (*f)* at Description No. Manually Opened RilR Loops Pool lemp = 150 F G >40 lbn/sec-ft2 Stuck Open 1 (a) 1 2 34 140 Relief Valve at Power 1 (b) 1 1 35 142 Stuck Open 2 (a) 1 2 29 143 Relief Valve From Isolated flot Standby 1**

2 (b) 1 33 144 Controlled 3 (a) VOCO200*F/ifr 2 44 151 Depressurization VOC0300*f/ilr 2 33 148 From Isolated 3 (b) VOC0300*f/llr 1** 47 151 llot Standby VOC0400 F/llr 1** 35 149 d MA (UA 4 None N/A 139 N VOC = Variable Operator Controlled. The operator would intermittently open and close the number of valves required to cool 60wn at the indicated rate.

~

^ Values Rounded to Hearest Whole Number N ** Ihis event does not conform to the plant licensing basis, and is presented for information only.

m Also, the operator would not go into llot Shutdown with only one RilR loop available.

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APPENDIX EVENT DESCRIPTIONS .

2261 265 9

e 1

38

EVENT 1* - STUCX-OPEN RELIEF VALVE FROM POWER OPERATION Event Sequence Time Temo Event Description

1. t = 0.0 T Pool temperature alarm at normal power a op o eration, Tech Spec limit (90 F).

Initiate actions to turn RHR loop (s)*

on for pool cooling. SRV fails open.

t a+ 3 minutes ** RHR loop (s)* on for pool cooling and torus spray.

t s

T Reactor Scram ***

s (Ts = 110 F. for CNS) t s+ 10.5 seconds Isolation (assuming mechanistic isolation on low water Level 2)"#

t a+ 10 minutes ## One additional SRV(s) manually actuated to depressurize the reactor.

Time tsand the number of SRV's to be manually actuated by the operator to be determined by analysis.

Corresponds to Event 1(a) if two RHR loops available and to 1(b) if one RHR loop available.

The operator can complete the actions necessary to turn the RHR loop (s) on within three minutes.

• • • Mode switch in Shutdown.

1. The bulk suppression pool temperature is assumed to be 90*F when an SRV inadvertently fails open. This is the maximum pool temperature allowed during normal power operation (Technical Specifications Section 3.7.A.1). Also Section 3.7.A.1 specifies that the reactor shall be scrammed from any operating condition when the suppression pool temperature reaches 110 F.

1. 1. The operator can determine which valve is stuck open within ten minutes.

• 1. Isolation is conservatively assumed to occur. Otherwise the main condenser could be used to depressurize the reactor and remove the decay heat *. hat is rejected to the suppression pool if isolation occurs.

2261 266 39

EVENT 2* STUCX-OPEN RELIEF VALVE FROM ISOLATED HOT STANDBY Event Secuence Time (Min.) Event Descriotion

1. t =t = 0.0 A transient has occurred, which causes the reactor to a s scram and isolate. The suppression pool temperature is at the normal limit (90 F). The operator initiates actions to turn RHR loop (s)* on for pool cooling.

t a+ 3 minutes ** RHR loop (~s)* on for pool cooling and torus spray.

t,<t<t, Reactor pressure maintained using SRV's.

t Operator begins reactor pressure vessel depressurization o

by opening additional SRV(s).*** Single SRV sticks cpen at 120 F.

The number of SRV's to be manually actuated by the operator to be determined by the analysis.

• Corresponds to Event 2(a) if two RHR loops available and to 2(b) if one RHR loop available.

1. The bulk suppression pool temperature is assumed to be 90*F rather than 120*F when the reactor is scrammed. This is the maximum pool temperature allowed before pool cooling would begin.

• • The perator can complete the actions necessary to turn the RHR loopss) on within three minutes.

• • • Section 3.7. A.l in the tech specs specifies that during reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200 psig if the pool temperature reaches 120 F.

2261 2u7 40

EVENT 3* ISOLATION AND REACTOR DEPRESSURIZATION Event Seouence Time (Min.) Event Descriotion

1. t =t = 0.0 Reactor isolation and scram. Pool temperature at a s normal limit (90*F). Initiate actions to turn RHR loop (s)* on for pool cooling.

t a+ 3 minutes ** RHR loop (s)* on for pool cooling and torus spray.

0<t<t C Reactor pressure maintained using SRV's (intermittent operation).

t Initiate cooldown*# using SRV's at 120*F.*"*

c Time t and the number of SRV's to be manually actuat6dbytheoperatortobe determined by analysis.

Corresponds to Event 3(a) if two RHR loops are availacle and 3(b) if one RHR loop is available.

• • The operator can complete the actions necessary to turn the RHR loop (s) on within three minutes.

• 1. Cooldown rate required to avoid pool instability to be determined by the analysis.

• "" Section 3.7. A.1. in the tech specs specifies that during reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200 psig if the pool temperature reacnes 120 F-

• The bulk suppression pool temperature is assumed to be 90 F rather than 120 F. This is the maximum pool temperature allowed before pool cooling would begin.

2261 268 41

EVENT 4* SMALL 3REAK ACCIDENT WITH ADS Event Seouence Time (Min.) Event Description 0.0 SBA occurs during normal plant operation **

t s

Reactor scram on high drywell pressure.

t g **# Main steam line isolation initiated automa-tically on low water Level 2.

tg + 3.5 seconds *** MSIVs fully closed.

t g Feedwater flow to reactor stops.

t d Water level drops to Level 1.

td + 120 seconds ADS automatically activates, depressurizing the reactor vessel.

No operator actions assumed, event runs to completion.

The suppression pool temperature versus discharge mass flux is determined by the analysis.

This event requires that the HPCI System fails; otherwise ADS would not be activated. Additional assumptions for this accident event are:

1. Loss of normal auxiliary power.

2. Limiting small break which results in the highest pool temperature when the SRV discharge mass flux > 40 lbm/ft sec.

The bulk suppression pool temperature is assumed to be 90 F rather than 120 F when the SBA occurs.

The MSIV closure time is 3.5 seconds.

• • 1. Approximately 7 seconds is required for the water level to drop to Level 2 following reactor scram.

2261 2c;9

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