ML20076L411

Forwards LOCA (Major RCS Pipe Ruptures) Reanalysis for Cycle 2 Assuming 2% Uniform Steam Generator Tube Plugging & Containment Purging at Time of Accident

ML20076L411
Person / Time
Site:	Sequoyah
Issue date:	07/14/1983
From:	Mills L TENNESSEE VALLEY AUTHORITY
To:	Adensam E Office of Nuclear Reactor Regulation
References
NUDOCS 8307190166
Download: ML20076L411 (80)

Text

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i TENNESSEE VALLEY AUTHORITY CH ATTANOOGA. TENNESSEE 37401

) 400 Chestnut Street Tower II July 14, 1983 Director of Nuclear Reactor Regulation Attention: Ms. E. Adensam, Chier Licensing Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Ms. Adensam:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 Enclosed for your information is the LOCA reanalysis for Cycle 2, units 1 and 2 for the Sequoyah Nuclear Plant, which assumes a two percent uniform steam generator tube plugging and containment purging at the time of the accident.

If you have any questions, please call K. P. Parr at FTS 858-2688.

Very truly yours, TENNESSEE ALLEY AUTHORITY C1.R )

L. M. Mills, nager Nuclear Licensing Enclosure 8307190166 830714 PDR ADOCK 05000327 P

PDR 4 00l 1983-TVA SOTH ANNIVERSARY An Equal Opportunity Employer l

1 i

, I. 4 15.4.1 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS OF COOLANT ACCIDENT) l The analysis specified by 10CFR50.46 " Acceptance Criteria for Emergency Core i Cooling Systems for Light Water Nuclear Power Reactors", is presented in this section. The results of the loss of coolant accident analyses are shown in Tables 15.4-4c and 15.4-4d and show compliance with the Acceptance Criteria.

The description of the various aspects of the LOCA analysis is given in Ref erence 33. The approved UHI Appendix K models reported in WCAP-8479, Revision 2, were updated to include NUREG-0630 fuel rod models [2] as specified in the NRC Staff SER on 1981 Version of the Westinghouse ECCS Evaluation Model [7].

The boundary considered for loss of coolant accidents is the Reactor Coolant System (RCS) or any line connected to the system up to the first closed valve.

Should a major break occur, depressurization of the Reactor Coolant Systen results in a pressure decrease in the pressurizer. Reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached. A Safety Injection System signal is actuated when the appropriate setpoint is reached.-

These countemeasures will limit the consequences of the accident in two ways:

1. Reactor trip and borated water injection complement void fomation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

2. Injection of borated water provides heat transfer from the core and prevents excessive clad temperature.

At the beginning of the blowdown phase, the entire Reactor Coolant System contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling. Af ter the break develops, the time to departure from nucieate boiling is calculated, consistent with Appendix K of 10CFR50E13. Thereafter, the heat transfer is based on local conditions with transition boiling and forced convection to steam as the major heat transfer mechanisms.

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When the Reactor Coolant System pressure falls below approximately 1250 psia the upper head injection accumulator begins to inject barated water directly into the reactor upper head region. This water is directed from the upper head directly to all but 8 peripheral assenblies in the core via the RCC guide tubes and UH1 support columns. This flow provides additional core cooling during the blowdown phase of the transient.

1 When the Reactor Coolant System pressure falls below approximately 432 psia the cold leg accumulators begin to inject borated water. The conservative assumption is made that accumulator water injected bypasses the core and goes out th' rough the break until the termination of bypass. This conservatism is again consistent with Appendix K of 10CFR50.

15.4.1.1 Thermal Analysis 15.4,1.1.1 Westinghouse Perfomance Criteria for Emergency Core Cooling

. Systen The reactor is designed to withstand thennal effects caused by a loss of coolant accident including the double ended severance of the largest Reactor Coolant System pipe. The reactor core and internals together with the Emergency Core Cooling System are designed so that the reactor can be safely shutdown and the essential heat transfer geometry of the core preserved following the accident.

The Emergency Core Cooling Systen, even when operating during the injection mode with the most severe single active failure, is designed to meet the

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Acceptance Criteria [1] ,

i 15.4.1.1.2 Method of Thermal Analysis The description of the various aspects of the LOCA analysis is given in WCAP-8339[33]. This document describes the major phenomena modeled, the interfaces among the computer codes and features of the codes which maintain compliance with the Acceptance Criteria. Differences between the approved 4324Q:1/050383

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Non-UHI Westinghouse Appendix K Model and the model used for th:se analyses are reported in WCAP-8479, Revision 2. The thermal analyses reported in this scction were perfomed with an upper head fluid temperature based on T cold *

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The UHI accumulator pressure set point ensures VHI actuation prior to upper head fluid flashing.

15.4.1.1.3 Containment Analysis The containment pressure analysis is perfomed with the LOTIC-2[37] code.

The transient pressure computed by the LOTIC code can be input to the WREFLOOD code for the purpose of computing the reflood transient. The containment pressure transient input to WREFLOOD from LOTIC is presented in Figure 15.4-17 for the C =0.8 D

DECLG imperf ect mixing case. The containment data used in the containment pressure analysis to detemine the ECCS backpressure is presented in Tables 15.4-2 and 15.4-3.

The mass and energy release rates used for the containment backpressure

' calculation as a function of time during blowdow'n are given in Table 15.4-4.

15.4.1.1.4 Results of Large Break Spectrum The following 4 breaks were investigated in this analysis:

double ended cold leg break (CD = 0.8) , perf ect mixing ,

double ended cold leg break (CD = 0.8), imperf ect mixing double ended cold leg break (CD = 0.4), imperf ect mixing double ended cold leg break (CD = 0.6), imperf ect mixing Figure 15.4-1 depicts the SATAN control volume scheme and can be utilized as a reference for the transient results. The transient results are presented in Figures 15.4-2a through 15.4-15d for the 4 breaks listed above. The following nomenclature applies:

per mix - perf ect mixing in upper head during UHI water injection imp mix - imperfect mixing in upper head during UHI water injection 4324Q:1/050383

Z - flowrate in el 1, 2, - flowrate at lower half and af dplane of core, respectively 2 - flowrate in el, 3, 4 - flowrate at upper half and top of core, respectively lo. pha. void in el 1, 2 - void fraction in lower half of core, 3 ft, sections, respectively

10. pha void in el 3, 4 - void fraction in upper half of core, 3 ft.

sections, respectively Z - flowrate'in el 43, 44 - flowrate from cold leg and UHI accumulators, respectively The time sequence of events for the analyses described below are shown in Tabl es 15.4-4 a-b. Tables 15.4-4 c-d present the peak clad temperatures and hot spot metal reaction for a spectrum of large break sizes.

The SATAN VI analysis of- the loss of coolant accident is perfomed at 102

, percent of the core licensed power. The core themal transient is also perfomed at this power level. The peak If near power, the peaking factor of the license application power level, and core power used in the analyses are given in Table 15.4-4c and d. Since there is margin between the value of the peak linear power density used in this analysis and the value expected in .

operation, a lower peak clad temperature would be obtained by using the peak linear power density expected during operation.

The plant parameters used in this analysis are listed in Table 15.4-4h. The net effect of these input parameters is conservative for the LOCA analysis for the following reasons:

1. Initial systen conditions are chosen so as to maximize stored energy in the RCS and fuel.

2. Cold leg accumulator parameters are selected to cover the potential impact of 1) minimum ECC delivery rate resulting in an extended period of steam / water mixing aP during reflood and 2) maximum ECC delivery rate leading to accumulator emptying prior to bottom of core recovery.

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3. UHI volumes are selected to maximize the upper head reheat time for the perf ect mixing case and minimize the UHI system's contribution to core cooling for the imperf ect mixing case.

For the results discussed below, the hot spot is defined to be the location of maximum peak clad tenperature. This location is given in Tables 15.4-4c and d for each break size analyzed.

Figures 15.4-2a through 15.4-16d present the transients for the principal parameters for the break sizes analyzed. The following items are noted:

Figures 15.4-2a The following quantities are presented at the clad through 15.4-4d burst location and at the hot spot (location of maximum clad temperature), both on the hottest fuel rod (hot rod):

1. fluid quality

- 2. mass velocity

3. heat transfer coefficient.

The heat transfer coefficient shown is calculated by the LOCTA IV code.

Figures 15.4-Sa The system pressure shown is the calculated pressure through 15.4-9d in the core. Core flowrates and core void fraction are also presented.

Figures 15.4-10a These figures show the hot spot clad temperature through 15.4-11d transient and clad temperature transient at the burst i location. The fluid tenperature shown is also for the hot spot and burst location. The nodal notation of the figures is defined in Table 15.4-4f.

Figures 15.4-12a These figures show the core reflood transient.

through 15.4-13d 4324Q:1/050383

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Figures 15.4-14a These figures show the Emergency Core Cooling Systen through 15.4-15d flows for all cases analyzed. Acth UHI and cold leg accumulators flow rates are included in the figures. As described earlier the cold leg accumulator delivery during blowdown is discarded until the end of bypass is calculated. The cold leg accumulator flow assumed is the sum of that injected in the intact cold legs.

Figures 15.4-16a These figures show the total cold leg accumulator through 15.4-16d mass injection prior to end of bypass, accumulator mass spilled out' break, calculated bypass deficit, and vessel inventory. ,

I The containment pressure transient for the limiting case break (CD = 0.8 imperfect mixing) is presented in Figure 15.4-17. Figures 15.4-16 and 15.4-18 shows the heat removal rates of the upper and lower compartment heat sinks

. respecti ve l y; the heat transfer model used is described in Reference 37.

Figure 15.4-19 gives the flowrate exiting the ice condenser drains. Figure 15.4-20 presents the tenperature transients in both the upper and lower j compartments. Figures 15.4-21 and 15.4-22 illustrate the heat removal rates by the sump and the ice condenser drain. Total lower compartment heat renoval is the summation of the rates given in Figures 15.4-18, 15.4-21 and 15.4-22.

The results of the break spectrum show that the double ended cold leg guillotine break, with a discharge coefficient of 0.8 and imperfect mixing in the upper head, is the worst break in tenns of calculated peak clad temp erature.

Due to the variances in UHI accumulator volume delivery as outlined in Table 15.4-21, the worst break (CD = 0.8) was subsequently run with a difference in volume delivery between the perfect (1060 ft3) and imperfect mixing case (900 ft3 ). The band applied in the Sequoyah ECCS analysis is developed as a bound to the possible variation in the total UHI water delivered. Variation in total delivery is postulated to result from two causes: variation which l results from measurement error together with the uncertainty associated with 4324Q:1/050383 l au [ ...' ~$'[ ~* $ ***f"?[**" 'V"'~ " ~ * * * * .

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.. a system performance and variation which results frcm a single failure in the UHI systen.

Each source of variation in total UHI delivery and its associated volume uncertainty is given in Table 15.4-4e. These are applied to_.the nominal setpoint. The sources of volume uncertainty not related to single failure are the result of system uncertainties. It is therefore appropriate to consider a statistical treatment for their application. These uncertainties are combined statistically by applying a Monte Carlo evaluation of the five renaining volume variations listed in Table 15.4-4e. Cases presented herein provide the results of a conservative application of this range of values.

The imperf ect mixing case was run utilizing the infonnation of Table 15.4-4e to develop a low delivery volume since the upper head drains early in the transient and subsequently voids the lower plenum and core; thus the minimum volume does indeed represent a conservative case. The imperfect mixing case was run at a high gas (1300 psia) pressure to allow for a 150 psi uncertainty in accumulator setpoint pressure. The high pressure for the imperfect mixing case represents the most conservative case since the smaller accumulator volume would be delivered in a shorter amount of time and earlier in the blowdown transient, thereby providing for a longer core heatup time. For the perfect mixing case, the information of Table 15.4-4e was used to develop a high delivery volume. This high volume and the low setpoint pressure (1200 psia) provides the most conservative combination of these UHI parameters because this results in the slowest injection and the longest period of upper head reheat (inherent in the perfect mixing case) prior to draining of this coolant into the core.

15.4.1.1.5 Effect of Containment Purging To assess the impact of purging on the calculated post-LOCA Sequoyah containment pressure, first a calculation was performed to obtain the amount of mass which exits through three available sets off purge lines during the initial portion of a postulated LOCA transient. Purge line isolation closure time is assumed at 4.0 seconds after receipt of signal; during this interval, 4324Q:1/050383

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the full flow area is presumed available. In addition, the tine to reach the S.I. signal setpoint and the delay necessary to generate the S.I. signal are conservatively assessed as 1.5 seconds total. Thus, flow through 3 pairs of  ;

fully open available purge lines was evaluated from 0.0 to 5.5 seconds for the postulated Double-Ended Cold Leg break. _

The calculation employed the 50-node TMD computer code model which is described in section 6.2.1.3.4. Referring to Figures 6.2.9 and 10, 24-inch purge supply lines are connected to volumes 34, 37 and 25; purge exhaust lines are connected to 36 and 25. Possible combinations of supply lines and exhaust lines open to the atmosphere were considered. Each of these purge lines is 2

represented by a flowpath of cross-section area equal to 2.948 ft and a

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total flow resistance factor equal to 3.98 (entrance and exit loss, three fully open butterfly valves and a debris screen). The most conservative two pairs of 24-inch purge and supply lines were assumed open in this calculation. In addition two 12-inch lines connected to TMD node 29 were modeled as open.

In a computation for ECCS performance the greatest impact on containment pressure occurs for the purge case of maximum air mass loss, which is based upon the two 12-inch lines being open and involves three purge lines open in the lower compartment [TMD elenents 34, 36 and 37] and one purge If ne open in the upper compahtment together with a cold leg break in TMD volume 1; 2620 lbs. of air are calculated to be lost in this case. The maximum air loss case is the limiting case because any steam lost via purging in an ECCS backpressure evaluation would otherwise be calculated to condense in the ice bed anyway. Therefore, any steam lost via purging is ultimately of no consequence in the containment pressure detennination, while any air loss directly reduces calculated pressure. To incorporate the TMD-calculated results, the initial compression peak of the LOTIC code was adjusted to consider the mass lost due to purging. The calculated LOTIC containment pressure thus reflects the loss of mass via purging during the first few seconds of the LOCA transient.

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The impact of the reduced containment pressure on ECCS performance is inclub.4 in the calculated peak clad temperature of 2137*F. Basing the plant Technical Specification peaking factor on this result pennits purging of the Sequoyah containment during normal operation to be conducted through three sets of purge lines. _,

15.4.1.17 Conclusions - Thermal Analysis For cases considered, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CFR50.46. That is:

1. The calculated peak fuel element clad temperature provides margin to the requirenent of 2200*F, based on an qF value of 2.237.

2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zin:aloy in the reactor. ,

3. The clad tenperature transient is terminated at a time when the core geometry is still amenable to cooling. The clad oxidation limits of 17 percent are not exceeded during or after quenching.

4. The core tenperature is reduced and decay heat is renoved for an extended period of time, as required by the long-lived radioactivity remaining in the core.

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References

2. D. A. Powers and R. O. Meyer, " Cladding' Swelling and Rupture Models for LOCA Analysis", NRC Report NUREG-0630, April,1980. l

7. " Westinghouse ECCS Evaluation Model,1981 Version", WCAP-9220 (Proprietary Version), WCAP-9221 (Non-Proprietary Version), February,1982.

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., ; 112.6 121.9 Acc. Einpty (CL) 108.8 28.0 28.0 28.1 f Pianped injection 63.3 P

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START 0.0 Rx Trip Signal .45 S. I. Signal 3.0 Acc. Inj ection (CL) 18.8 End of Blowdown 112.1 Bottom of Core Recovery 112.1 Acc. Empty (CL) 114.9 Pumped Inj ection 28.0 End of Bypass 55.9 UHI Acc. Ini, 2.7 i

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CD = 0.8 DECLG CD = 0.6 DECLG CD = 0.4 DECLG

.e J Results Peak Clad Temp. *F 2137 2077 1860 Peak Clad Location Ft. 7.25 7.5 7.5 Local Zr/H 2O Reaction (max) 6.3 5.1 2.3 Local Zr/H2O Location Ft. 7.25 7.5 7.5 Total Zr/H p0 Reaction <0.3 <0.3 <0.3

. Hot Rod Burst Time sec. 62.7 58.8 89.6 Hot Rod Burst Location Ft. 6.25 7.25 5.75 Calculation Assumptions Core Power, Mwt,102 percent of 3411 Peak Linear Power, kw/ft,102 percent of 12.18 Peaking Factor (At License Rating) 2.237 o Accumulator Water Volume (Cold Leg Delivered) 1050 Ft3 per accumulator 3

j Accumulator Water Volume (UNI Delivered Value) 900 Ft

I. Steam Generator Tube Plugging Level 2 percent, uniform k

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TABLE 15.4-4d LARGE BREAK - PERFECT MIXING CD = 0.8 DECLG Results i

Peak Clad Temp. *F 1982 Peak Clad Location Ft. 6.25 Local Zr/H 2O Reaction (max) 3.1 Local Zr/H2O Location Ft. 7.5 Total Zr/H 2O Reaction <0.3 Hot Rod Burst Time sec. 6.9 Hot Rod Burst Location Ft. 6.0 4

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Calculation Assumptions Core Power, Mwt,102 percent of 3411 Peak Linear Power, kw/ft,102 percent of 12.18 Peaking Factor [At License Rating) 2.237

! Accumulator Water Volume (Cold Leg Delivered) 1050 Ft3 per accumulator i Accumulator Water Volume (UHI Delivered Value) 1050 Ft3 Steam Generator Tube Plugging Level 2 percent, unifom l

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TIME (SECONDS)

.

• w a.

I 4 i Figure 15.4-2d. Fluid Quality - DECLG (CD = 0.4), Imp Mix 4ie e se e..

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f'*

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NODES 9 AND 13 g -100 -

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. TIME , SECONDS

\ . . .

+ .

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r I ._..

Figure 15.4-3a. Mass Velocity - DECLG (Cg = 0.8), Imp Mix l .- > -

1 g-., . . _ _ . .

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f .

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1 300 i - - . ~

i 0 100 200 300 .

. TIME (SECONDS) o 9

Figure 15.4-3b. Mass Velocity -DECLG (Co = 0.8), Per Mix

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NODES 13 AND 14

!- - h . ' -. >=

(3 . +<r- p 1 i .

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se

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100 150 200 250 j

, I TIME (SECONDS)

I - - - -

f,

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, a: _; i y re. " ' E t: Figure 15.4-3c. Mass Velocity - DECLG (CD = 0.6), Imp Mix &

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r e .

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TIME (SECONDS) _

i e.

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i Figure 15.4-Sa. .RCS Pressure - DECLG (Co = 0.8), imp Mix s h? .ff

, .%. es#E g tbYt g(_ J .

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l

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w-. . Ic

, n <: Figure 15.4-Sb.'.. f1CS Pressure - DECLG (CD = 0.8), Per Mix t

,rAZ. .

t'n a i * , l.,

• s' -

+::

s e t

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g j f i j ,f f , f ', j i l l l {

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Figure'15.4-Sc. RCS Pressure -DECLG (CD = 0.6), imp Mix t

I r

g . .

l..

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3, f! , 1 e e

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Figure 15.4-6a. Flow Rate - DECLG (Cg = 0.8), Imp Mix Lower Half of Core i u l

,! r V ii. 1:

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13 -

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I' l' DJ o 1 Figure 15.4-6b. Core Flow Rate - DECLG (CD = 0.8), Per Mix Lower Half of Core f-

.y ? '. !, . .

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(."' , TIME (SECONDS)  ;

!i - - - -

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1 f ;( e. , ...

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TIME (SECONDS) p..'a-gi4 . ,. o O! l

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Figure 15.4-6d. Core Flow Rate - DECLG (CD = 0.4), imp Mix Lower Half of Core l i

a 14 s .

4. l;. ; '4,

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Figure 15.4-12c. Reflood Transient - DECLG (CD = 0.6), Imp Mix I

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Figure 15.4-12d. Reflood Transient - DECLG (CD = 0.4), Imp Mix m

b&

v&* .

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e Figure 15.4-3d. Mass Velocity - DECLG (CD = 0.4), Imp Mix s 4*. p

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( Figure 15.4-4a. Heat Transfer Coefficient- DECLG (CD = 0.8), Imp Mix ,

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TIME (SECONDS)

L . - _ . _ .

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Figure 15.4-4c. Heat Transfer Coefficient - DECLG (CD = 0.6), Imp Mix h

+. ,

n I .' ,

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.. TIME (SECONDS) -

i -. . .

..u ,

} ,

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g . _. _ _

e j Figure 15.4-4b. Heat Transfer Coefficient - DECLG (Cf = 0.8), Per Mix .

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c. .

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< ~

w ..;.

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- TIME (SECONDS) -

1 Figure 15.4-4d. Heat Transfer Coefficient - DECLG (Co = 0.4), imp Mix Y

g -A.

d I . 4 ,

4

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=

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0 0 100 200 ~ 300 TIME (SECONDS) i i

l Figure 15.4-13a. Reflood Rate -DECLG (CD = 0.8) Imp Mix l

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x J

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TIME (SECONDS) -

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em

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e Figure 15.4-13b. Reflood Rate - DECLG (Co = 0.8', Per Mix

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g l 100 200 ~300  ;

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Figure 15.4-13d. Reflood Rate - DECLG (Co = 0.4), imp Mix i

I g' t e lk y +u , .. w

1. s. * -

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2000 -

1000 - -

0 2 l '1 0 20 40 6,0 80 TIME (SECONDS) 1 l

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1 Figure 15.4-14a. Accumulator Flows - DECLG (CD = 0.8), imp Mix

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TIME (SECONDS) -

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,-- $ .e

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m.= 2 ::.. _~4

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+

t._...=_,..- . _ . _ . .

m l l -

. . , _ ...- n .

- 0 100 200 300 TIME (SECONDS) 1 i

i

.O Figure 15.4-15b. SI + Accumulator Flow - DECLG (CD = 0.8), Per Mix l

- y

..w  %

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m f

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e. _ , ^ .

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w 0 100 200 300

', TIME (SECONDS) 9 Figure 15.4-15c. SI + Accumulator Flow - DECLG (Cg = 0.6), imp Mix F

i r =

g s'* -

l -9 e , u e g.

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Figure 15.4-15d. Si + Accumulator Flow - DECLG (CD = 0.4), imp Mix aW. t on. * "

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Figure 15.4-16. Upper Compartment Structurai Heat Removal Rate - DECLG (CD = 0.8), imp Mix ,

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94 e,.eopn .w , e.wyg .p.

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Figure 15.4-17. Containment Pressure - DECLG (CD = 0.8), Imp Mix h

'$r

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TIME (SECONDS) .

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Figure 15.4-18. Lower Compartment Structuraf Heat Removal Rate -DECLG (CD = 0.8), imp Mix

+

r -

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E N* -

p

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l Figure 15.4-20. Compartment Temperature - DECLG (CD = 0.8), Imp Mix 6

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TIME (SECONDS) l t

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~

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TIME (SECONDS) .

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Figure 15.4-22. Heat Removal by LC Drain - DECLG (CD = 0.8), imp Mir

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