ML20011F622

Analysis of Small Steamline Break Performance W/O Low Pressurizer Pressure Safety Injection North Anna Units 1 & 2.

ML20011F622
Person / Time
Site:	North Anna
Issue date:	02/26/1990
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20011F620	List: ML20011F619 ML20011F622 ML20011F623
References
NUDOCS 9003070050
Download: ML20011F622 (37)

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l ATTACHMENT 1 lc ANALYSIS OF SMALL STEAMLINE BREAK PERFORMANCE WITHOUT LOW PRESSURIZER PRESSURE SAFETY INJECTION NORTH ANNA UNITS 1 AND 2

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i Table of Contents i

Section ~ Title Page

.i List of Tables..................... 3 Y ,

List of Figures.................... 4 1.0 Introduction.............-.......... 5 2.0 Method o f Ana lys i s. . . . . . . . . . . . . . . . . 9 3.0 . Analytical Results......... ....... 15 4.0 Conclusions........................ 36 5.0 References......................... 37 k

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List of Tables No. Title Page

.1, Presentation of Pressurizer Pressure Channel. Errors and Margins ................ 12 2 Steam Line Break Protection for' NAPS ........ 13-14 3 Sequence'of Events- 0.24 Sq. Ft. Per Loop Steam Break ...............................-34 E 4 Comparison of Statepoints- 0.24 Sq. Ft. Per . 35

. Loop 1 Break vs. UFSAR Inside Break-with Power

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List of Figures No. Title Page 1 Hi-1 Actuation Time ......... ................ 20

-2 Break Flow from Two Steam Generators ........ 2?

3. Single Loop. Steam Pressure .................. 22 4 Core Heat Flux (Fraction of HFP) . . . . . . . . . . . . 23 5, Total Reactivity - $ ......................... 24 6 .1-Loop' Cold Leg Temperature ................. 25 7 Pressurizer Pressure .'....................... 26 8 1-Loop Hot Leg Temperature ................... 27 9 1-Loop SG Inventory ......................... 28 Core Inlet Boron-Concentration .............. 29 10 -

11 Core Power (MW) .............................. 30 12 Core Heat Flux _(.067 Sq. Ft. Per Loop SLB) .. 31 13 ~1-Loop Cold Leg Temperature (.067 Sq. Ft.

Per Loop SLB) ............................. 32 14 Pressurizer Pressure. (.067 Sq. Ft. Per Loop.

-SLB) ...................................... 33 Page 4

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1.0 INTRODUCTION

In References 1 and 2, Virginia Electric and Power Company identified the potential for the low pressurizer pressure safety injection (SI) f

function to be delayed relative to the' safety analysis assumptions or to not- function for events which result in a harsh environment - inside

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containment. It has been shown that the error in the pressure transmitter

. output signal induced by a harsh environment may exceed the difference between the safety injection actuation setpoint and the bottom of the -

pressure range over which the instrument is calibrated (i.e.,' the channel span). In such a case actuation of safety injection cannot be guaranteed.

This results from the fact that the transmitter output is highly nonlinear for input pressures which are significantly outside the calibration span.

A summary of the calculated pressurizer pressure channel errors and the available margin between the actuation setpoint and the bottom ~of the calibration span is shown in Table 1. ,

As' discussed in References 1 and 2, an evaluation showed that existing safety analyses remain bounding for both large and small break Loss of

- Coolant Accidents (SBLOCA) and for the large, hypothetical main steam line break events. The evaluation further concluded that the results for a small steam line break inside containment vould continue to be acceptable, but that safety injection actuation might be delayed with respect to the assumption in the safety analysis. As a result, an unreviewed safety

. question was determined to exist in that the probability of a malfunction of equipment important to safety has increased with respect to the Page 5

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currently ' reviewed and- approved licensing basis. The purpose of this

• analysis is to present the technical information - required for NRC r

resolution of the unreviewed safety question.

Following NRC approval, the analysis results will be incorporated into the UFSAR.

For the SBLOCA, the low pressurizer pressure SI occurs at times less ,

than one minute into the transient. From UFSAR Table 15.3-2, the low pressure reactor trip time is 24.7 seconds for a 3 inch break and 16.7 seconds for a 4 inch break. Low-low pressure SI is generated very shortly af ter the trip (i.e., within a few seconds). The mass and energy releases for the SBLOCA are approximately an order of magnitude less than the mass and energy releases for the LBLOCA. For the ' double ended cold leg .

guillotine-large break LOCA, it takes about 20 seconds to reach the-peak containment temperature. Thus, the anticipated rise s in containment-temperature for the SBLOCA is minimal for the time interval in which the 4

pressurizer low pressure SI is expected to actuate. In additio'n, the temperature time constant for the pressure transmitters being used, Rosemount Model 1153 Series 0, is 4.8 minutes. Based on this inforniation, it is expected that the pressure transmitters would function in what is L.

-basically a mild environment. Therefore it'is not necessary to apply the harsh environment errors to generation of this SI signal for SBLOCA.

For the large main steam line breaks examined in the UFSAR, safety y injection is initiated based on secondary side indications and not on low pressurizer pressure.

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l The small steam line break event currently analyzed and presented 'in I the North : Anna UFSAR (Reference 3) corresponds to a 262.lb/sec steam release rate at 1020 psia. This break size was chosen - to bound The q

effects of a stuck open secondary relief, dump or safety valve.- It is l of interest to note that none of these events, which are normally classified. as ANS Condition II, or anticipated traasients, would result in a harsh containment environment since the release point is.outside containment. .

However, while'such a case is not specifically analyzed in the UFSAR, one could postulate a small steam line break inside containment which results in releases equivalent to or greater than the UFSAR " credible" case. In this case,. a harsh environment could be created and the low-low i pressurizer pre.ssure safety injection might not function for the reason described in References 1 and 2. In this case safety injection initiation would result from one of the following (see Table 2):

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1) High steam . flow coincident with either low RCS Tavg or low steam pressure for break sizes in excess of about 0.24 sq. ft. per loop.

-2) High header to steam line delta-P (expected to function if the main steam non-return valves close as designed).

3) High containment pressure.

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i Virginia Elect-;u cad Power Company has performed a bounding analysis of a small stetm line treak inside containment . to ,show the impact of failure of the low pressurizer pressure SI function. The analysis is .

t summarized in this report. '

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g Page 9 Yj t 2.0 METHOD OF ANALYSIS -

The following assumptions were made for the small SLB analysis:

1) The low-low pressurizer pressure SI is assumed not to be  !

available for a break inside containment, f i

2) A break size is selected which is-just underneath the size which would actuate high sc.eam flow protection.

The case being examined is a break at hot zero power (HZP) at end of life for the standard. reasons discussed in the UFSAR. The break examined 1 was a 0.24 sq. ft. per loop split break. This break size was -i determined to be the maximum size which would not result in a high steam line' flow actuation signal (see Table 2).

3) A conservative estimate of safety injection actuation time on 1

'High-1 containment pressure was made based on existing North Anna containment analyses (Ref. 4). Examination of the available data shows that the time to High-1 containment pressure actuation can be correlated well with the inverse of the initial break flow. A plot of such a correlation is shown in Figure 1.

o Also shown on the Figure is the-value of the correlating parameter (39,000 lbm/sec divided by initial break flow rate) for the 0.24 sq. f t. per loop break. The estimated Hi-1 actuation time is about 25 seconds. To conservatively account for uncertainties, the actuation time from Figure 1 was increased by 50% and an actuation time of 38 seconds was used.

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1 A detailed review of the P.eference 4 analysis was performed te GL confirm the applicability of the analysis results to the current evaluation.

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It was concluded that the methods and assumptions used maximize the calculated time to High-1 actuation, and that the use of these actuation times in the current analysis i=

appropriate and conservative. Nevertheless the additional 50%

factor-discussed above was applied. ,.

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4) Safety injection pump acceleration and valve stroke was conserv-atively modeled to take 15 seconds, l

5) The safety injection flow ccpability correspor. ding to the operation of one HHSI/ charging pump (minimum safeguards) was assumed. The time delay to purge boron-free water in the.  :

piping .from the BIT to the cold legs was modeled. A minimum l BIT boron concentration of 12250 ppm was-assumed (less than s

the minimum specified in Technical Specification Limiting-Condition for Operation 3.5.4.1).

6) The non-return valves-(NRV's) in the main steam lines were assumed NOT to functicn. This is consistent with way the High-1 actuation time (see Assumption 3) was estimated. If the NRV's function, the total amount of secondary inventory available for discharge is significantly reduced, and safety injection on header-to-line Dp would be expected very quickly (in about 1

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The analysis was performed using the Virginia Power RETRAN two-loop

-model documented in Reference 5 and the RETRAN02, M00003 transient analysis code (Reference 6).

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Table 1 Presentation of Pressurizer Pressure Channel Errors and Marains '

Bottom of Nominal Safety- Margin, Total Channel = Statistical Channel Span, Injection' psi Allowance (%/ Psi)*

psia. Setpoint, psia Normal-Env. Harsh Env. #

A. SURRY UNITS 1 AND 2 1 l

1700.0 1718.0' 18.0 +,-1.97/15,8 +19.60/156.8 ,

-18.10/-144.8 i B.' NORTH ANNA UNITS 1 AND 2 1700.0 1765.0 65.0 +,-1.76/14.1 +19.7/157.7

-17.9/-143.6 -)

• Error (psi) = Error (% Span) x 800 psi span /100

1. Based on vendor's stated performance characteristics for operation during a-design bases' event (DBE). Applicability of these performance characteristics to Surry and North Anna has-been demonstrated in accordance with 10 CFR 50.49 and associated regulatory guidance.

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Table 2 Steam Line Break Protection for North Anna-A. Safety Injection Source Tech. Spec Setooint* Notes

1. Low-low pressurizer 1765 psig *May not function -i

pressure-2/3 channels for breaks inside t' containment (Ref.

1).

2. High steam line flow 40% of full load in 2/3 line coinc w/ steam flow (at no-load) *May not actuate for A) Low-low RCS Tavg, 543 F a) breaks inside '

or containment if B) Low Steam Line 600 psig NRV's close Pressure b) breaks less than about 0.24 sq. ft.

per loop.

3. High delta-P between

• May not actuate. 4 steam line and header 100 psid .for break inside containment if NRV's don't close

4. High-1 containment 17 psia

• Will not actuate pressure for breaks'outside-containment Nominal setpoint. The impact of uncertainties is included in-the analysis.

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1 Table 2 (CONT.)

Steam Line Break Protection for North Anna B. Steam Line Isolation .

Source Tech. Spec Setooint* Notes

1. High steam line flow 40% of full load in 2/3 line coinc w/ steam flow (at no-load) *May not actuate for A) Low-low RCS Tavg, 543 F a) breaks.inside or containment if B) Low Steam Line 600 psig NRV's work

_ Pressure. b) breaks less than.

about 0.24 sq. ft.  ;

per. loop.

2. High-2 containment 17.8 psia

• Will not actuate pressure for breaks outside contaiment

• Nominal setpoint. The impact of uncertainties is included in the analysis.

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3.0 ANALYTICAL RESULTS 4

Results- of the analysis are shown in Figures 2-11. Because the case '

modeled is a symmetric blowdown, results for the other steam generators and reactor coolant loops are essentially the same. A sequence of events is presented in Table 3.

Figure 2 shows the total break flow coming from the two " intact" steam generators. The flow path is from the generators to the header, then in a reverse direction back through the NRV in the faulted loop to the break. '

.If the NRV were to close as designed, the SI signal on high header to line differential _. pressure (See Table 2) would be rapidly generated.

1 This can be seen clearly from Figure 3. The header to line differential pressure setpoint is 100 psid (Table 2). Figure 3 shows that l- the pressure in a single loop drops by 100 psi very quickly ( 4 20.

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seconds in -this case). Thus from the standpoint of delaying safety injection and increasing the severity of the cooldown, failure of the NRV's to close_is the limiting assumption.

l Core heat flux is shown in Figure 4. The peak heat flux is reached at about 340 seconds. The turnaround in heat flux results from a l-combination of the continued decay in steam flow (Figure 2), a leveling off of the secondary pressure decay (Figure 3) and the negative reactivity j

l effects of soluble boron addition to the core via safety injection (Figure l

10). The peak heat flux attained is only about one-fourth of that 1;

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resulting 'from the current limiting hypothetical case (break inside containment with offsite power available) presented in the UFSAR (see Table 4 for a detailed comparison).

Figure 5 shows the total core kinetics reactivity in dollars. The design end of cycle (EOC) shutdown reactivity of 1.77% was assumed to be inserted at the beginning of the transient. Recriticality occurs at about i

60 seconds. A reactivity balance was performed which showed that with i a more realistic shutdown margin l assumption (representative of recent 1

d. : : reload cores), recriticality would most likely not occur at all.

The cold leg temperature on the faulted loop is shown in Figure 6. ]

The intact loop temperature' response is essentially identical since the

> case being modeled is a symmetrical blowdown.

l Figure 7 shows the pressurizer pressure response. The initial rapid i

depressurization is retarded somewhat as the pressurizer drains and the ~

upper head begins to flash. The effect of safety injection terminates the depressurization at about 360 seconds and a gradual repressurization ,

begins.

I Hot leg temperature is shown in Figure 8 and is very simlar to the cold leg . response (Figure 6). Figure 9 shows the steam generator inventory in. the loop fed by the turbine-driven AFW pump. Because of the large AFW flow to the generator, inventory never drops below about 80% of the initial value. In the two SG's fed by the motor- driven AFW pumps, the Page 16

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Page 17 inventory drops to about half the initial value before starting to increase again. The effects of the heat transfer degradation in the steam-generators-fed by the lower capacity motor-driven pumps is conservatively neglected to maximize the effects of the cooldown.

1 Figure l10 shows the boron concentration at the inlet plenum, as previously discussed. Figure 11 shows core power, which correlates closely with core heat flux (Figure 4).

Table 4 provides a comparison of the limiting statepoint (maximum core ,

heat flux) .for the current case with the UFSAR hypothetical break (inside containment with of fsite power). From a DNB perspective, the hypothetical case will bound the current case for the following reasons:

- the heat flux for the UFSAR case is almost four times that for the current (0.24 sq. f t. per loop) case.

- The inlet temperature gradient across the core is negligible for the current case, where the UFSAR case has about a 150 F gradient. A large temperature gradient results in higher radial power peaking. 1

- RCS pressures are in the same approximate ranges for the two cases. The lower pressures for the current case will be more than offset by the substantial reduction in core heat flux.

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As a result of these factors, and because of the inherent conservatism of the current analysis, particularly in the area of the overall-core ,

reactivity calculation, as discussed previously, explicit power peaking and DNB calculations are not required for this caso. Since the UFSAR case meets the ANS Condition II DNBR criterion with margin, so will the current l case. 4 4

The analysis case ' presented above is expected to boun'd those which L.

would result for other break sizes. As discussed previously, larger break'- -

sizes will generato a high steam flow / low steam pressure, or. / low Tavg signal much earlier than the Hi-1 actuation ' assumed in the above case.

For smaller break sizes, the Hi-1 containment actuation signal would be delayed due to the lower _ mass and energy releases to the containment. >

However, the cooldown and return to power would be expected to be less ,

severe for the smaller breaks. In the limit, there will exist some break ~

size small enough such that H M1 actuation will not occur. 1

Review of available North Anna containment analysis data shows that the smallest break size examined yielded an initial break flow of 430 lb/sec and resulted in Hi-1 actuation at 97 seconds. This break flow J corresponds to a- RETRAW break area of about .067 sq. ft. per loop.

l l An additional case was examined to show the effects of no SI on very small steam breaks. . A break size of .067 sq. f t. per loop, as discussed l

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above, was. simulated. No SI was assumed, although the North Anna containment results showed SI on Hi-1 actuation would occur. The most I E significant results are shown in Figures 12-14. As expected, the small

'I break size limits the amount of power generation.that can be achieved in i the core. (Figure 12) to well within those achieved for the limiting ,

1 hypothetical breaks-(Table 4). Note' that while the heat flux is slightly higher than achieved for the larger (0.24 Sq. F.t. per loop) break, this

- stems from the conservative assumption that no SI at all occurs. The actual North Anna analysis for this break size resulted in.SI on'High-1 containment pressure over 100 seconds prior the predicted return to '

criticality (97 seconds vs >200 seconds-see Figure 12). As a result, the conclusions drawn for the .24 sq. ft. per loop case are valid for the entire. spectrum,of break sizes. ,

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FIGURE 1-HI-1 ACTUATION TIME-

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TABLE 3  ;

SEQUENCE OF EVENTS- 0,24 SQ. FT. PER LOOP STEAM LINE BREAK j l

I 5' Time, seconds Event 0.0 Steam line break occurs 19.6 Low-low RCS Tavg setpoint reached 38.0 Safety Injection on High-1 containment pressure 54.1 Low steamline pressure setpoint reached 340 Peak heat flux  !

,' 900.0 End of simulation 4

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i -- TABLE 4 STEAMBREAK ACCIDENT STATEPOINTS Hypothetical Break 0.24 Sq. Ft. per Loop Outside Cont. Inside Cont. With Power UFSAR Case A UFSAR Case B Current Case

.n Core Heat 13.08 23.72/21.80 6.21 Flux, % of 2775 MWT(Note 1)

RCS' Pressure, 1423 1163/910 761 Psia Loop A Inlet Temp,:F 411 363/355 391-Loop B Inlet Temp, F 495 513/500 392 Core Boron 45.7 26.9/70.6 262.1 Concentration, PPM -

RCS Flow, 100 100 100 Reactivity, ,

% deltaK/K .0409 .0521/.0159 5.4E-5

_ Time, sec. 58 45.25/70.0 340 g1 DNBR >W-3 DNBR Limit >W-3 DNBR Limit- >W-3 DNBR Limit Note 1- Fraction of Initial Core Rating of 2775 Mwt.

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4.0 CONCLUSION

S An analysis of small steam line break (SLB) events inside containment has demonstrated that there is adequate protection to ensure that the applicable accident analyses acceptance criteria are met for the entire spectrum of break' sizes without relying on the low-low pressurizer pressure safety injection initiating function. Recent evaluations have shown that this function might not occur in a harsh containment environment.

The analyses have shown that other sources of safety injection,-i.e.

high-1 containment pressure, high steam flow with low steam pressure or low Tavg, or high steam header to line differential pressure, provide adequate protection when required and the low-low pressurizer pressure may therefore be considered a diverse source of protection which need not be relied on to demonstrate acceptable results.

Assuming that the low-low pressurizer pressure safety injection

- function does not actuate in a harsh environment does not result in calculated conditions for any steam line break which are more limiting h than those calculated for-the large (hypothetical) steam line break case examined in the UFSAR. As a result, NRC review and approval of these analyses will resolve the outstanding unreviewed safety question involving operation of low-low pressurizer pressure safety injection.

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5.0 REFERENCES

1. Licensee Event Report (LER) 89-043-00, " Low Pressure SI May Not Actuate During a Harsh Environment in Containment Due to Instrument Loop Inaccuracies", Surry Power Station Units 1 and 2, January 12,1990.

2. Licensee Event Report (LER) 89-018-00, " Uncertainty Associated Vith Harsh Environment Below ESF Transmitter Range", North Anna Power Station Units 1 and 2, January 11,1990.

3. UFSAR Section 15.2.13, " Accidental Depressurization of the Main Steam System," Rev. 4, 6/88. >

4. WCAP-11431, " Mass and Energy Releases Following a Steam Line Rupture For North Anna Units 1 and 2", February 1987 (Proprietary). i WCAP-11432. " Mass and Energy Releases Following a Steam Line Rupture For North Anna Units 1 and 2", February 1987 (Non-Proprietary).

5. VEP-FRD-41A, " Reactor System Transient Analyses Using the RETRAN Computer Code," May 1985.

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6. EPRI NP-1550-CCM-A, "RETRAN A Program for Transient Thermal- l Hydraulic Analysis of Complex Fluid Flow Systems," November, 1988.

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