ML19309E502

Revised Section 5.4.1,of Cycle 5,Core Performance Analysis

ML19309E502
Person / Time
Site:	Maine Yankee
Issue date:	04/17/1980
From:	Maine Yankee
To:
Shared Package
ML19309E498	List: ML19309E497 ML19309E502
References
YAEC-1201-01, YAEC-1201-1, YAEC-1202, YAEC-1202-01, YAEC-1202-1, NUDOCS 8004220530
Download: ML19309E502 (35)

Text

,.;

ATTACHMENT A REVISED SECTION 5.4.1 0F YAEC-1202 5.4.1 STEAM LINE RUPTURE 5.4.1.1 General The rupture of a main steam line is a non mechanistically postulated infrequent event. A rupture of a main steam line increases the rate of heat extraction by the steam generators, which in turn, results in a rapid cooldown of the primary system. With negative moderator and fuel reactivity coefficients, the cooldown will produce a positive reactivity addition. Of prime concern in this accident is the prevention of the core from returning to criticality and causing fuel damage. There fore, the assumptions made in the analysis are consistent with maximizing the primary system cooldown and subsequent reactivity additions. The fastest cooldown which l

results in the most rapid reactivity addition, occurs when the break is at a steam generator nozzle. This break location is assumed for all cases analyzed. Taking no credit for non-return valves in the ruptured steam line conservatively covers the possibility of a steam line rupture outside containment with respect to return to criticality. Maintaining suberiticality is a conservative limit to assure that no fuel damage will result.

The Steam Line Rupture Accident has been re analyzed for Cycle 5 using the same computer codes and methods applied in the Reference Safety Analysis. The key parameters which have changed for Cycle 5 are the EOC Moderator Temperature Defect and available shutdown CEA worth. As indicated in Table 5.4 (of YAEC-1202), the Moderator Temperature Defect is slightly larger for Cycle 5 than assumed in the Cycle 4 or Reference Analysis.

Available shutdown CEA worth, accounting for the worst stuck rod and including uncertainties is 6.64%Ap for Cycle 5 at HFP as compared to the assumed value of 6.5%Ap in the Reference Analysis and 6.4%Ap in the Cycle 4 analysis. At HZP the 2.9%Ap assumed in the Reference Analysis was modified to 3.2%Ap (in YAEC-1202) for the Cycle 5 analysis. This value remains conservative with respect to the predicted Cycle 5 available shutdown worths (refer to Table 5.6 of YAEC-1202).

Minimum boron injection times have been improved prior to Cycle 4 operation due to the installation of check valves in the safety injection lines. These check valves assure that a major portion of the safety injection line is filled with 1720 ppm borated water at all times. This modification reduces the delivery time of boron injection from 280 seconds in the Reference Analysis to 130 seconds for Cycle 4.

Further modifications have led to a delivery time for cycle 5 of 84 seconds following SIAS with 1 HPSI pump.

5.3o 8004220

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The post-trip feedwater bypass-valve setpoint is important, particulary for the HZP case, ia limiting the cooldown following a major SLR (see Maine Yankee Reportable Occurrence

1. 80-001/01L-0, dated January 30, 1980). For this reason a modification to the post-trip feedwater bypass-valve setpoint was necessary to prevent a return to critical for the low power end-of-cycle (EOC) SLR events. The analyses reported herein provide the bases for the revised setpoints.

This section does not address the radiological consequences of a SLR but is concerned with maintaining the reactor suberitical.

The radiological consequences of a SLR outside containment were evaluated in section 4.14 of Reference 3, and are not affected by the cycle 5 changes.

The peak containment pressure following a MSLB inside containment reported in reference 3, section 4.16, remains bounding for Cycle 5.

5.4.1.2 Method of Analysis The re-analysis of the steam line rupture incident was performed using a slight modification of the plant accident model nodalization described in Reference 13.

The model includes neutron kinetics with fuel and moderator temperature feedback, the shutdown CEAs, the safety injection system, the reactor coolant system, the steam generators and the main steam and feedwater systems. The modified nodal configuration is shown in Figure 5.4-1.

The modification consists of removing one node i

(19) from the feedtrain model to represent the auxiliary feedwater header, in addition to adding a fill curve representation of the auxiliary feedwater pumps.

These modifications allow the model to appropriately divide the auxiliary FW flow between the intact and affected steam generators in those cases were automatic initiation of auxiliary feedwater is assumed.

For convenience the modified model nodalization shown in i

Figure 5.4-1 was used for the HFP case with auxiliary feedwater and for both HZP cases. The HFP case without auxiliary feedwater used the HFP model described in Reference 13.

Four cases were run in order to assess the effects of the modified post-trip feedwater bypass valve override opening setpoint and automatic initiation of auxiliary feedwater flow on both the HFP and HZP breaks. The allowable post-trip feedwater bypass valve openning varies with whether the auxiliary feedwater system is assumed to be automatically initiated.

Since the allowable post-trip bypass flow rate is higher if automatic initiation of auxiliary feedwater is not assumed (i.e.

AFW system remains manually initiated), and it is desirable from

an operational standpoint to have a higher post-trip feedwater flowrate (in order to recover steam generator level faster after the initial decrease in level due to the collapsing of the steam generator voids), two setpoints have been provided for the revised feedwater bypass valve post-trip override openning setpoint. The higher setpoint will limit normal post-trip feedwater flow to 500 gpm/SG and is assumed for those case's where the auxiliary feedwater system is not automatically initiated. For those cases with automatic initiation of auxiliary feedwater the feedwater bypass valve post-trip setpoint is reduced so as to limit normal post-trip feedwater flow to 400 GPM/SG. Hence the four cases analyzed here are:

1) HFP with 500 GPM/SG post-trip FW bypass valve setpoint (37%), no auxiliary feedwater assumed.

2) HZP with 500 GPM/SG post-trip FW bypass valve setpoint (37%), no auxiliary feedwater assumed.

3) HFP with 400 GPM/SG post-trip FW bypass valve setpoint (32%)

and auxiliary feedwater assumed to automatically initiated on low SG 1evel after a 5 minute delay.

4) HZP with 400 GPM/SG post-trip FW bypass valve setpoint (32%)

and auxiliary feedwater assumed to be automatically initiated on low SG 1evel after a 5 minute delay.

Other assumptions applicable to all cases except as noted were:

A. The EOC ARI moderator temperature defect with worst stuck r

CEA given in Table 5.4 (of YAEC-1202) is used with a +25%

uncertainty. Moderator reactivity is based on lower Plenum (core inlet) Node 3 temperature.

B. The slope of the reactivity vs. fuel temperature function at EOC is increased by 25%.

C. No credit is taken for closure of the non-return check valve on the ruptured steam line. This results in the blowdown of all three steam generators until the excess flow check valves close and covers the possiblity of a steam line rupture outside containment.

D. Reactor trip occurs 0.9 seconds after low steam generator

[

pressure setpoint of 478 psia (uncertainty included) is reached.

i E. Excess flow check valves on the intact steam lines close 5 seconds after a steam generator pressure signal at 392.7 psia (uncertainty included).

F. Safety injection actuation signal generated at 1622 psia (1600+22 psia) pressurizer pressure after a 0.9 second delay.

G. The reactor coolant pumps are assumed to be running throughout the transient maximizing the heat transfer to the steam generators and resulting in the fastest cooldown.

H. One HPSI pump is available for the injection of cold borated i

(1720 ppm) ECCS. Even though two HPSI pumps delivering cold water influences the RCS cooldown, the beneficial effects received from the added boron overwhelms the cooldown resulting in the one pump case being limiting.

I. Worth of borated SI water is assumed to be a function of I

moderator density (see Figure 5.4-2).

J. No credit is taken for Low Pressure Safety Injection or Accumulators.

K. Single phase (steam only) critical flow through ruptured steam line is assumed in order to maximize energy removed by the ruptured steam generator.

l L. Constant overall primary-secondary heat transfer coefficient (determined at full power conditions) to maximize energy removal.

i M. The feedwater system is modeled explicitly (see Reference 13 l

of YAEC-1202 and Figure 5.4-1) to ensure proper calculation of post-steam line break feed flow and enthalpy. The feedwater regulation valve (FWRV) is ramped closed from the 100% feed flow position in 10 seconds after receipt of the i

reactor trip signal. Feedwater flow is then allowed through the FWRV bypass valves which are assumed to open to their i

post-trip override setpoints.

For the HZP case the FW bypass valves are assumed to be initially open enough to match the assumed reactor power.

Upon reactor trip they are assumed to open to their post-trip override setpoints.

i N. For cases 3 and 4 automatic initiation of 2 auxiliary feedwater pumps is assumed at time zero due to the uncertainty of SG 1evel response in the affected steam generator. Auxiliary feedwater is thus assumed to reach the steam generators 5 minutes after initiation of the event.

4 5

5.4.1.3 SEQUENCE OF EVENTS CASE NO. 1 HFP W/.:00 GPM FW BYPASS SETPOINT, NO. AUXILIARY FW.

TIME (SECONDS)

EVENT 0

Normal operation at HFP (2630 MWt).

0+

Double-ended rupture of main steam line.

3.3 Pressure in affected steam generator drops to 478 psia 4.2 Reactor trip signal generated FWRV begins to close.

4.2 FWRV bypass valve start to open.

5.6 Pressure in affected steam generator reaches 392.7 psia and excess flow check valves begin to close.

10.6 Excess flow check valves fully closed terminating blowdown from intact steam generators.

14.2 FWRV fully closed and FWRV bypass valves fully open to their post-trip settings.

18.2 Pressurizer pressure drops to 1622 psia initiating a safety injection actuation signal (SIAS).

112 Minimum reactivity margin

-5.03%Ap

=

112+

Shutdown Margin Increasingly Negative CASE NO. 2 HZP W/500 GPM FW BYPASS SETPOINT, NO. AUXILIARY FW.

TIME (SECONDS)

EVENT 0

Normal operation at HZP (1 MWt).

0+

Double-ended rupture of main steam line.

3.3 Pressure in affected steam generator drops to 478 psia 4.2 Reactor trip signal generated FWRV begins to close.

4.2 FWRV bypass valve start to open.

5.0 Pressure in affected steam generator reaches 392.7 psia and excess flow check valves begin to close, i

~

10.0 Excess flow check valves fully closed terminating blowdown from ir. tact steam generators.

14.2 FWRV fully closed and FWRV bypass valves fully open to their post-trip settings.

15.5 Pressurizer pressure drops to 1622 psia initiating a safety injection actuation signal (SIAS).

283 Minimum reactivity margin

.134%Ao.

=

283+

Shutdown Margin Increasingly Negative CASE NO. 3 HFP W/400 GPM FW BYPASS SETPOINT, AUTO. AUXILIARY FW.

TIME (SECONDS)

EVENT O

Normal operation at HFP (2630 MWt).

0+

Double-ended rupture of main steam line.

3.4 Pressure in affected steam generator drops to 478 psia 4.3 Reactor trip signal generated FWRV begins to close.

4.3 FWRV bpass valve start to open.

5.6 Pressure in affected steam gtinerator reaches 392.7 psia and excess flow check valves begin to close.

11.5 Excess flow check valves fully closed terminating blowdown from intact steam generators.

14.3 FWRV fully closed and FWRV bypass valves fully open to their post-trip settings.

18.2 Pressurizer pressure drops to 1622 psia initiating a safety injection actuation signal (SIAS).

l 615 Minimum reactivity margin

-4.81%Ap

=

615+

Shutdown Margin Increasingly Negative 1

I

O 6

CASE No. 4 HZP W/400 GPM FW BYPASS SETPOINT, W/ AUTO. AUXILIARY FW.

TIME (SECONDS)

EVENT O

Normal operation at HZP (1 MWt).

0+

Double-ended rupture of main steam line.

3.25 Pressure in affected steam generator drops to 478 psia 4.15 Reactor trip signal generated FWRV begins to close.

i 4.15 FWRV bypass valve start to open.

5.0 Pressure in affected steam generator reaches 392.7

-psia and excess flow check valves begin to close.

l 10.0 Excess flow check valves fully closed terminating blowdown from intact steam generators.

i 14.15 FWRV fully closed and FWRV bypass valves fully open to their post-trip settings.

i 15.6 Pressurizer pressure drops to 1622 psia initiating a safety injection actuation signal (SIAS).

i 580 Minimum reactivity margin

.042%Ap l

=

580+

Shutdown Margin Increasingly Negative 5.4.1.4 RESULTS i

Significant parameters for each case are plotted in Figures 5.4-3 through 5.4-24.

In all cases the reactor is prevented from returning to criticality by the combined actions of the low steam generator pressure reactor trip signal, high pressure safety injection system, and feedwater control system. No operator action is assumed prior to 10 minutes in any case. Operator action, throttling of SI flow, is assumed after 15 minutes to prevent the RCS from being charged solid and potentially overpressurizing or violating NDT limits.

The coincident loss of offsite power and/or manual tripping of the RCPs upon SIAS are considered by noting that a decrease in the RCS cooldown rate due to the reduced primary to' secondary heat transfer j

coefficient during natural circulation and the reduction in i

feedwater flow due to the loss of power to feedwater pumps for the

~

loss of offsite power. Thus the cooldown rates (and hence positive l

reactivity addition rates) presented in cases 1 thru 4 are bounding (with. respect to) either los's of offsite power or operator trip of RCPs on SIAS. Figure 5.4-25 and 5.4-26 show the RCS boron l

concentration vs. time (assuming 1 HPSI) for cases 1 thru 4 and for l

a case where the RCS flow is conservatively assumed to be natural circulation with a power to flow ratio of 0.5 (natural circulation measurements at Maine Yankee indicate a power to flow ratio of 0.2).

It is evident that at the times of peak reactivity the boron concentration assuming natural circulation are not significantly different. Hence considering the expected slower RCS cooldown no return to power is predicted for either coincident loss of offsite power or operator tripping of the RCPs upon SIAS.

5.4.

1.5 CONCLUSION

S The low steam generator pressure reactor trip signal, high pressure injection system and feedwater control systems provide adequate protection from core damage in the event of a steam line rupture accident.

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FIGURE 5.4-12 Stnm Lins Rupture Accid nt Thre2 Loop Full P:wsr w/ Auto. Aux. F;2d 400 GPM/S.C. Feed Flow Reactor Coolant System Pressure Vs. Time fi! Us !!EWi 5Wii Wii!! i!!!Mi R lijf =Riis !@is in !!i! l@M si !E is!El ini HE !E M 51!i!E nMi WI!!! iin !!i! !@ !W s# liii tRini ei i18 IE!Mi WE Es in!nE !!!! sii sfi M W Mi i!E E si !E im Wi iisMi n!i im MMi $! Ei iinliHWW En lili ili! !!! !$ # 11dii isi iin.ifft d Ed. is 1111,utm,.!.,J,!!u..ile.te tm d-- -mmt 4 n -. m. _m m.. u.. _. mis J.ii_s itE rm ir.ii. m 1 m ..Enta tm t Ei i!H sn is u!!ss Riin En Mi M fin iM in SiiB HEHs WEsEu! iin E sifin liEE 2500 ~

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FIGURE 5.4-14 St:am Linn Ruptura Accident Three Loop Full Power w/ Auto. Aux. Feed ~ 400 GPM/S.G. Feed Flow Flow Rate to Steam Generators Vs. Time uN,' HF. g13 g :;=: w.. un

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