ML20010E943

Evaluation of Reactivity Response for Steam Line Break Event W/Unterminated Emergency Feedwater Flow.

ML20010E943
Person / Time
Site:	Crystal River
Issue date:	05/06/1981
From:	Andrews J, Lockey M, Schomsher R BABCOCK & WILCOX CO.
To:
Shared Package
ML20010E938	List: ML20010E937 ML20010E943
References
86-1125549, 86-1125549-00, NUDOCS 8109090164
Download: ML20010E943 (15)

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i EVALUATION OF REACTIVITY RESPONSE FOR A STEAM LINE BREAK EVENT WITH UNTERMINATED EMCRGENCY FEEDWATER FLOW

PREPARED FOR FLORIDA POWER CORPORATION CR-3 PROJECT CONTRACT #5B2-7087 TASK #184 MAY 6, 1981 l

B&W REFERENCE DOCUMENT: 86-1125549-00 PREPARED BY: //64.orArr/

REVIEWED BY: Mdad REVIEWED BY: b APPROVED BY: O. b 1 e

8109090164 810903 PDR ADOCK 05000302 P PDR

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I CONTENTS g

, PAGE I. Background 1 II. Scope 2 III. Method 3 IV. Results. 4 l

V. Conclusions 5 PAGE ,

Table 1 Major Assumptions and Input Parameters 6-7 l

I Table 2 Steam Line Break Sequence of Events 8 Table 3 Steam Line Break Results 9 l

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Figure 1 Reactor Power /RC Temperature 10-11 Figure 2 Core Total Reactivity vs. Time 12 l

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ABSTRACT To reduce the patential for a loss-of-all-feedwater event, it has been

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recommended that the Steam Line Rupture Matrix Signal which isolates emergency feedwater be eliminated. A previous report, " Evaluation of Steam Line Break Consequences Associated with Removal of Rupture Matrix I

Signals from Emergency Feedwater Valves" addressed the effects of this I change on the FSAR analysis, but did not specifically reanalyze the reactivity effects associated with a Steam Line Break (SLB). This report provides analyses, using current methods, which demonstrate that a return h

to criticality will not occur following a SLB and unterminated Emergency  ;

Feedwater(EFW) flow. .

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I. Background _

Following the CR-3 transient in February,1980. Florida Power Corporation convened a Nuclea

• Safety Task Force to identify potential changes which could significantly improve plant safety and reduce the potential for major transients. One of the principal reconnendations of this task force (and NUREG-0667) was that the Steam Line Rupture Matrix System i (SLRMS) signals which close the emergency feedwater isolation valves be eliminated. Such a change would significantly reduce the potential for a loss-of-all-feedwater event and this signit'icantly improve the overall safety of the plant.

In support of this change, however, it was necessary to address the effects on the FSAR steam line break analysis which had been perfonned assuming automatic isolation of the EFW by the SLRMS. A report entitled, " Evaluation of Steam Line Break Consequences Associated with Removal of Rupture Matrix Signals from Emergency Feedwater Valves", dated May 23, 1980 and revised l

l on June 27, 1980 was prepared for that purpose. Using the same methods as i were originally used in the FSAR, the analysis described in that report

! demonstrated the acceptability of continued EFW flow from the standpoint of I containment response. However, the method originally used for reactivity response to an SLB (the SECRUP analog code) was no longer in use or available.

In lieu of the reanalysis, the report included an evaluation showing that the probability of a SLB accompanied by any stuck control rod could be conservatively estimated to be less than 2.75 x 10 -7 per reactor year.

However, the NRC staff rejected this evaluation, and, as a consequence, the proposed change has not been made.

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2 Florida Power Corporation has subsequently inst.ituted a major program to upgrade the emergency feedwater system and associated controls. The new control system, the Emergency Feedwater Initiation and Control (EFIC) system will replace the existing SLRMS and provide emergency f eedwater

! isolation to only the steam ge:v rator in the affected loop. However, pending installation of the EFIL system, removal of the rupture matrix signals to the EFW valves is still a very desirable change.

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This report is intended to address the NRC concern associated with reactivity effects following a SLB with continued EFW flow. For this l

l analysis, current methods (specifically the TRAP code described in topical report BAW-10128) are used in lieu of the methods employed in the original FSAR analysis.

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l II. Scope l

The purpose of this analysis is to document the core total reactivity response to a large double-ended steam line break with unteminated l emergency feedwater flow to both steam generators. As a result of the l proposed removal of the Steam Line Break Rupture Matrix signals from the emergency feedwater valves, it is possible to have EFW flow to both the l

affected and unaffected steam generator loops. This situation requires

! operator action to recognize and isolate EFW from the unaffected steam generator loop. This analysis predicts the system response, specifically core reactivity, for the SLB event where this operator action has not been performed. Core reactivity during the transient is detemined by the insertion of control rods, the insertion of boron by the HPI system I

3 and the reactivity feedback due to fuel temperature and moderator density changes. Of special interest is the continued cooling provided by EFW. It is desired that core suberiticality be maintained throughout the transient.

III. Method The Double-Ended Steam Line Break event was analyzed using the TRAP 2 (version 6) digital computer code. Major assumptions and input parameters are provided in Table 1. The transient examined is a double-ended rupture of a single main steam line. The analysis assumptions were chosen to provide a conservative response with respect to core overcooling thus maximizing the core reactivity response and thus increasing the potential for recriticality. For this reason, a conservatively large secondary inventory, end of cycle kinetics parameters, and appropriate response

, delay times were used. No loss of offsite power was assumed. Main steam isolation is assumed to occur by closure of the main steam isolation valves rather than the turbine stop valves, thus allowing additional overcooling.

Control rod reactivity insertion is assumed to provide sufficient reactivity to account for power deficit and only a 1% shutdown margin at HZP conditions.

Shutdown reactivity provided by boron in the HPI system is assumed to be supplied by only 1 HPI pump. Positive reactivity feedback provided by the reactivity coefficients has been conservatively estimated as end of life values to maximize reactivity feedback response.

The double-ended rupture of a 22 inch (ID) main steam line results in a rapid increase in steam flow, secondary depressurization and an increase

4 in heat transfer across the steam generator. Overcooling of the primary pressure results in a low RC pressure trip at 1.1 seconds into the transient. Reactor trip initiates turbine trip and TSV closure, but this function was ignored in the analyses to provide conservative overcooling.

The secondary steam pressure continues to decrease actuating a Steam Line Rupture; Matrix signal (600 psia SG pressure) at 4.7 seconds into the transient. This signal initiates closure of the Main Steam Isolation valves and main feedwater isolation. EFW is subsequently initiated as a result of the loss of all main feedwater. The steam generator in the affected loop continues to depressurize and boil dry at a rate faster than the steam generator in the unaffected loop. Primary system pressure decreases to the HPI actuation signal at 5.3 seconds, thus starting HPI pumps and the addition of borated water to the primary system.

l Core reactivity becomes negative innediately following reactor trip

providing at least a 1% suberitical margin. This margin is diminished by the positive reactivity feedback caused by the decrease in core average temperature due to the overcooling. Increased negative reactivity is l provided by the introduction of borated water by the HPI system. This

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negative reactivity competes with the positive reactivity caused by prolonged cooling by EFW injection. The borated water surolied by one HPI pump provides sufficient negative reactivity to overcome continued EFW flow, thus eventually resulting in increasing subcritical margin. A minimum suberitical margin of .10% ak/k occurs 16 seconds into the transient followed by an increasing j suberitical margin.

III. Results The resulting reactivity response shodn in Figure 2 indicates that the

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• q j 5 core will remain subcritical ti.roughoyt the transient with a minimum suberitical margin of <10'; ak/k occurring at 16 seconds. Reactor power and reactor coolant _ pressure are shown in Figure 1. A sequence of events i is provided in Table 2 and sumary of pertinent results in Table 3.

l The analysis indicates that removal of the Steam Line Rupture Matrix signal to the EFW valves will not result in sufficient overcooling to cause l core recriticality with credit being assumed for only one HPI pump. Operator I action would normally be expectett to occur to terminate EFW flow to the affected steam generator loop, thus allowing it to boil completely dry and tenninate its contribution to the avercooling.

IV. Conclusions l Since subcriticality can be maintained even for the conservative assumptions presented in this analysis, it is concluded that removal of the SLBRM signals I, to the EFW valves does not result in an unacceptable core reactivity response.

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6 TABLE 1 MAJOR ASSUMPTIONS AND INPUT PARAMETERS p

A. Thermal Hydraulics Parameter Value >

1. Power level (102%), MWt 2619

2. RC pump heat, MWt 18 8

3. Primary Flow rate,1bm/hr 1.3 x 10 7

4. Secondary flow rate, lbr.1/hr 1.06 x 10

5. SG outlet pressure psia 925

6. Steam generator inventory,1bm 46200 lbm 3

7. Initial pressurizer inventory ft 800

! 8. EFW flow rate 1

( affected loop gpm 880 l, unsffected-lo'op gpm 520 EFW temperatura, F 40 i

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l l B. Kinetics Parameters Parameter Value Doppler coefficient ak/k/F -1.3 x 10-5 Moderator coefficient ak/k/F -3.0 x 10 -5 Boron Worth ppm /% ak/k 108 Shutdown margin, % ak/k 1.0 f

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l C. Trips a6d Setpoint Times

( Parameter Value

! 1. Low pressure RPS trip, psig 1800 l 2. HPI actuation setpoint, RC pressure psig 1500

3. SLB rupture matrix signal psig 600 t

a. MSIV closure, delay, s. 2.5

[ stroki , s. 5.0

b. Main feedwater isolation, s. 17

4. EFW actuation delay (sec.) 50 E-

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STEAM LINE BREAK SEQUENCE OF EVENTS

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Event Time - (seconds) l l Rupture occurs 0.0 Reactor trip on low primary pressure 1.1 Control rods begin to fall 1.5 Low steam generator pressure occurs 4.7 HPI actuation signal reached 5.3 MSIV's closed 12.2 HPI flow established 15.3 Minimum subcritical margin reached 16.

l Main feedwater isolated 24.2 EFW flow establisheo 50.

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STEAM LINE BREAK RESULTS l

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( ' Minimum suberitical margin, %Ak/k .10 1

I Time of minimum suberitical margin, s. 16 Y

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