ML20010F055

Forwards Info to Close Out SER Items Re Steam Voiding in Reactor Vessel Analysis,Loss of Offsite Power or Tripping of Reactor Coolant Pumps During Main Steam Line Break & Clarification of Transient Anaysis W/Potential Fuel Damage

ML20010F055
Person / Time
Site:	Waterford
Issue date:	09/04/1981
From:	Maurin L LOUISIANA POWER & LIGHT CO.
To:	Tedesco R Office of Nuclear Reactor Regulation
References
3-A1-10, 3-A1.10, Q-3-429.20, Q-3-A29.20, W3P81-2041, NUDOCS 8109090314
Download: ML20010F055 (17)

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LOUISIANA i4a OaAnONOe srsta P O W E R & L i G H T[ P O BOX G008

• NEW ORLEANS. LOUISIANA 70174

. [504) 366 2345 Ur ONdy"sNU September 4, 1981 W3P81-2041 3-A1.10 Q-3-A29.20 Q

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Mr. R. L. Tedesco Assistant Director of Licensing d ' ~d s

U. S. Nuclear Regulatory Commission ,

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Washintton, D. C. 20555 ( g g ,,,y 0 81981 w ors (* H1

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SUBJECT:

Waterford 3 SES TA Decket No. 50-382 'NA > '

SER Open Items M@ /

Dear Mr. Tedesco:

Please find taclosed information necessary to close cut the following SER open items:

a. Steam Voiding in the reactor vessel analysis.

I b. Loss of offsite power or tripping of the reactor coolant pumps during a main steam line break,

c. Clarification of transient analyses with potential for fuel damage.

This information, presented to the ACRS on August 5, 1981, will be included in the next FSAR amendment.

Yours very truly, v

i L. V. Maurin Assistant Vice President Nuclear Operations LVM/RMF/ddc cc: W. M. Stevenson, E. L. .Blake, S. Black o

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8109090314 810904 PDR ADOCT 05000382 C PDR -

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l 15A.l. SER Open Item No. 8 Statement - "The current CESEC model does not properly account for steam i f ormation in the reacter vessel . Therefore, for all events in which  !

(a) the pressurizer' is calculated to drain into the hot leg, or (b) the system pressure drops to the saturation pressure of the hottest fluid in the system during normal operation, we require the applicant to reanalyze these events with an acceptable model or otherwise justify the acceptability of Waterford 3 Chapter 15 analysis conclusions performed with CESEC. "

Response

I For all Chapter 15 events for which the pressurizer ' fluid is calculated to drain into the hot leg, or the system pressure drops below the saturation pressure of the hottest fluid in the system, the hottest fluid will be located in the relatively stagnant upper head region of the reactor vessel *. The CESEC-1 code, used in the FSAR Chapter 15 analyses did not explicitly model the neam formation in the reactor vessel upper head region. The latest version of CESEC, namely CESEC-III, explicitly models steam void formation and collapse in the upper head region of the reactor vessel. Heat transfer -

from metal structures to the reactor coolant system (RCS) fluid is modeled in addition to flashing of the reactor coolant into steam during depressur-ization of the RCS. Following the reactor coolant pump (RCP) coastdown <

due to loss of offsite power or manual shutoff following SIAS, thermal-hydraulic decoupling of the upper head region is characterized in CESEC-III by progressively decreasing flow to the upper head from the upper plenum region.

l The Steam Generator Tube Rupture (SGTR) with Less of Offsite Power (LOP) is the. most limiting event with respect to the duration of voids in the RCS.

The Steam Line Break (SI.B) inside containment with LOP and one HPSI pump i failure is the most liv.iting event with respect to maximum steam volume in the RCS. Both of these events were reanalyzed using the CESEC-III code, i The results are shown in Tables 15A.1-2 and 15A.1-3 and Figures 15A.1-1 through 15A.1-4 for the comparative analysis for the SGTR and SLB events.

'The conclusions from the comparative analyses for the SGTR and SLB events bound the other depressurization events for which void formation is less

, limiting and/or non-existent. This is due to slower coolhun rates and higher minimum RCS pressures for the other depressurization event. Therefore, the qualification of CESEC-I against CESEC-III for the SGTR and SLB cvents provides the necessary justification for the acceptability of the Waterford-3 Chapter 15 analyses conclusions for depressurization events as presented in the FSAR.

• Significant differences impacting the analyses between the two CESEC versions are summarized in Table 15A.1-1.

1 Steam Generator Tube Rupture Event The reanalysis of the SGTR event indicated the following.

The modelling of the stagnant upper head region with metal structure heat transfer results in the formation of voids in this region. The void volume in the upper head region peaks at about 535 cu. ft during the transient and gradually decreases under the cocabined action of the HPSI flow and the controlled cooldown at the steam generators. The duration of the voids depends on the rate of cooldown of the primary side and the HPSI flow rate.

The amount of voids predicted !s not large enough to expand the steam bu'_ble beyond the upper head region and to the elevation of the hot legs.

Therefore, natural circulation cooldown of the RCS is not impaired.

l The prediction of the upper head voids in the reanalysis does not alter the conclusions of the previous analysis. The results cf the reanalysis not only show insignificant impact on the offsite doses, but also demonstrate that the plant can be maintai ..nl in a stable condition

. by the collapse of the upper head voids in a timely manner through 4

manual control of the cooldown rate. Table 15A.1-2 lists the results of the original FSAR analysis using'CESEC I and the results of the reanalysis using CELCC-III. -

Subsequent to collapse of the upper head voids, the plant is maintained in a stabic condition, and the operator can bring the plant tc the shutdown cooling er.try conditions, by cooling down the RCS at a pre-scribed cooldown rate using the intact steam generator and the condenser.

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Steam Line Break Event _-

The re-analys'is of the limiting

• steam line break (SLB) event indicated

. the following:

The large SLB event with concurrent loss of offsite power (LOP) resuits in the t.nximum vaid formation in the reactor vessel upper head. The event is

initiated from 102% power by a double ended guillotine break of the main steam line inside containment. Reactor trip and main steam isolation signhls 4 are generated on low steam generator pressure at 2.1 seconds into the 4 transient. Initial void formation occurs in the upper head at 7 seconds

! into the transient as a result of the rapid cooldown of the reactor coolant system (RCS), and is further enhanced by the coastdown of the

- reactor coolant pamps due to the LOP. Auxiliary feedwater ( AFW) is

! initiated on low steam generator water level in the intact steam gener-

! ator due to level shrink that occurs after the closure of the main steam j isolation valves.

i

Full flow from the turbine driven AFW pump plus one motor driven AFW pump is conservatively assumed to begin filling the intact steam generator

coincident with closure of the main feedwater isolation valves at 22.1 i seconds into the transient. (The auxiliary feedwater actuation system logic prevents feeding the ruptured steam generator.) The pressurizer ,

l empties at 30 seconds. At approximatley 110 seconds the affected '

steam generator dries out. The void in the upper head reaches a maximum of 9E ft3 at approximately 135 seconds into the transient. At about

' 150 seconds into the transient the pressurizer liquid level is re-i established. The liquid level in the intact steam generator reaches the

point at which the automatic AFW system shuts off flow at approximately 370 seocnds into the transient. Thereafter the voic' volume in the upper

head continues to reduce due to the primary coolant twell from decay and

! residual heat in the RCS components and from safety injection and charging.

j. At a maximum of 30 minutes after the initiation of the transient the

operator initiates plant cooldown using the intact steam generator.

! At no time during the event does the upper head void region enter the j hot leg. Therefore, natural circulation cooldown of the RCS is not 4 inhibited.

l I Table l'JA.1-3 lists the results of the original FSAR analysis using CESEC-I i and the results of the re-analysis using CESEC-III. Figure 15A.1-3

compares the RCS pressures calculated by CESEC-I and CESEC-III for this i event. Figure 15A.1-4 presents the reactor vessel liquid level as a function I of time. Re-analysis of this event using a model which properly accounts

for void formation in the upper head region during a steam line break event does not alter any conclusions reached in the FSAR.

i i 7 The limiting SLB ever.t with potentid for post-trip degradation in fuel performance is a full power SLB inside containment, with loss of offsite 4 power coincident with the SLB (see Appendix 15A.4), failure of one HPSI pump, a stuck CEA, and an actr.'tiu. of t.uxiliary fcch.?ter.

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SUMMARY

OF SIGNIFICANT DIFFERENCES l ,

BETWEEN CESEC-III AND CESEC-1 MODEL CESEC-Ill CESEC-1 1 -THERMALHYDRAdLIC 26 N0 DES, UPPER HEAD 16 NODES i

EXPLICITLY MODELED 1

RCS FLOW _XPLICITLY MODELED INPUT TABLE RCP's FOUR, EXPLICllLY MODELED TWO 1 WALL HEAT . EXPLICITLY MODELED NONE SGTR OPTION CRITICAL FLOW DARCY EQUATION MIXING IN RV - ASYMMETRIC RESPONSE ASYMMETRIC RESPONSE l EXPLICITLY MODELED INCLUDED IN REAC-TIVITY CALCULATION FOR SLB i: -

, 3D FEEDBACK YES NO

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TABLE 15A.1-2 COMPARISON OF RESULTS l STEAM GENERATOR TUBE RUPTURE W1TH LOP CESEC-1 CESEC-Ikl PRIMRY-SECONDARY INTEGRATED LEAK (LBM)

AT.1800 SECS. 60,739 66,420 J

INTEGRATED MSSV

, STEAM RELEASE (LBM) l AT 1800 SECS. 78,300 84,600 TOTAL STEAM RELEASE (LBM) -

761,810 803,800 1.28 1.21 MINIMUM DNBR OFFSITE DOSE-THYROID (REM) 66.5* 73.0**

• LitllTING SER VALUE AT EAB

• • ESTIMATED

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TABLE 15A.1-3 COMPARISON OF RESULTS FULL POWER STEAM LINE BREAK, INSIDE CONTAINMENT, WITH' LOSS OF 0FFSITE POWER, HPSI PUMP FAILURE, AUTOMATIC ACTUATION OF AUXILIARY FEEDWATER 1

CESEC-1 CESEC-III MAXIMUM POST-TRIP

! REACTIVITY (%Ap) +0 . 0002 -0.05 4

MAXIMUM POST-TRIP CORE POWER

(% OF 3410 MW) 10,4 7,0 CORE FLOW AT .

TIME OF MAX POWER (FP. ACTION) 0.043 0.05'5-MINIMUM POST-TRIP DNBR > 1.2 > 1.2 )

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1 15A.3. SER Open Item No.10 Statement - "We require the applicant to provide evaluation of the effect of loss of AC pcwer during the stear. line break." ,

Response

The major concern for main steam line break (MSLB) events is that the CEA design provides adequate shutdown reactivity worth. Under the assumption of the most reactive CEA stuck in the fully withdrawn position, the CEA

worth must be sufficient to preclude unacceptably la ge amounts of fuel damage due to any post-trip return to power that may be calculated to occur as a result of the severe primary system cooldown. MSLB events initiated at fuii power with loss of offsite power (LOP) at time zero (coincident with the MSLB) and without loss of offsite power are presented in the FSAR.

Figures 15A.3-1 and 15A.3-2 demonstrate that the consequences of a MSLB event coincident with a LOP bound the consequences of a MSLB event with a LOP occuring at times greater than zero. Figure 15A.3-1 presents the maximum post-trip reactivity versus LOP time. Figure 15A.3-2 presents

{, the minimum post-trip DNBR versus LOP time *. The post-trip return to power

peaks occurs at about 60 seconds into the transient if there is no LOP prior to this time. Therefore, as seen in Figures 15A.3-1 and 15A.3-2, for any cases with LOP after 60 seconds the same maximum post-trip reactivity and minimum post-trip DNBR are calculated.

The HSLB event initiated at zero power which is presented in the FSAR  ;

maximizes the potential offsite doses by maximizing the initial steam

, generator liquid inventory instead of maximizing the patential for post-trip ,

degradattan in fuel perforrance. The effect of LOP upon dose is to make the' condenser unavailable for cooldowns using the intact steam generator.

1 The technical specification on the power dependent CEA insertion limit ensures that there is no approach to DNB far MSLB events initiated at zero '

power. ,

j Therefore, the full power MSLB with LOP coir.cident with the MSLB case

presented in the FSAR is also bounding for the zero p'ower MSLB event in terms of fuel performance. The zero power MSLB with LOP coinci/ ant with the MSLB is bounding in terms of offsite dose.

The results presented in Figures 15A.3-1 and 2 were obtained using the CESEC III code (see Section 15A.1). The effect of automatic actuation of auxiliary feedwater was included in the analyses. ,

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15A.4 SER Open item N9 11 Statement -

"We requ.ce the applicant to provide information which

-explains why the stuck-open atmospheric dump valve event for Waterford 3 results in fuel damage whereas the steamline break event does not restit in exceeding DNBR limit. " .

Response

The stuck open atmospheric dump valve (ADV) event in comoination with

'; loss of offsite power (LOP) upon turbine-generator trip potentially results in a limited number of fuel pins experiencing DNB. The main steam line break (MSLB) events do not result in a minimum DNBR of less than 1.19. The focus of concern for analysis of degradation in fuel per-

'ormance is upon two different transient regimes for the stuck-open

AbV and fGLB events. The analyses for increased main steam flow events, such as the stuck open ADV, maximize the potential for pre-reactor-trip degradation in fuel performance. The analyses for MSLB events maximize the potential for post-trip degradation in fuel performance.

The core protection calculators (CPCs) will assure that the minimum DNBR will not be less than 1.19 prior to reactor trip for any overcooling transient, unless a single fallare of loss of offsite power (LOP) is assumed. However, when LOP is assumed, the increased main steam flow event can result in a calculated minimuin DNBR of less than 1.19, as shown

, in Figure 15A.4-1. The increased main steam flow causes a reduction in DNBR and the CPC's will initiate a reactor trip signal on the projected minimum DNBR of 1.19: The reactor trip signal generates a turbine trip which is assumed to result in a LOP. The LOP then causes the reactor coolant pumps to begin to ccast down. The resultant reactor coolant flow reduction causes the DNBR to decrease below 1.19 before the insertion of the control element assemblies reduces the reactor power.

The above sequence of events and consequences are the same for any increased main steam flow event., whether the initiating event is a stuck open ADV, inadvertent opening of a turbine bypass valve, or a small steam line break.

A large MSLB will have less potential for degradation in fuel performance l prior to reactor trip than the above ever.ts since reactor trip will occur i earlier in the transient due to low steam generator pressure. The major '

concern for the MSLB events is that the CEA design provides adequate shutdown reactivity worth. Under the assumption of the most reactive CEA stuck in the fully withdrawn position, the CEA worth must be sufficient to preclude. unacceptably large amounts of fuel damage. This would be due ;

to any post-trip return to power that might be calculated to occur as a result of the severe primary system cooldown causec' by the MSLB., The analyses for the FSAR demonstrate that the CEA worth for the Waterford Unit 3

HSSS is sufficient to preclude any fuel damage during the post-trip portion of a ftSLB event.

Therefore, the increased main steam flow event with LOP is calculated to-result in a minimum DiiBR less than 1.19 during the pre-trip time period, whereas the MSt.B events result in post-trip values of minimum GNBR greater than 1.19.

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