ML17213A908

Forwards Addl Info Re Small Feedwater Line Break Analysis. Info Needed to Close Confirmatory Issue Addressed in Section 15.10.2 of SER

ML17213A908
Person / Time
Site:	Saint Lucie
Issue date:	12/14/1982
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
L-82-542, NUDOCS 8212210101
Download: ML17213A908 (25)

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"AGGRESSION NBR;82122101'01 DOC ~ DATE! 82/12/14 NOTARI'ZEDe NO DOCKET

.FACIL:.50~389 'St,'ucie Plant<, Uni~t 2i Florida 'Power 8 Light Co> 05000389 AUTH ~ NAME AUTHOR "AFF IL'I'ATION

.UHRIGi R O'Es -Florida, Power 8,Li'ght Co,:

RBCIP ~ NAME 'RECIPIENT AFFIL'IATION EISENHUTiD,G ~,Division of Licensing t

SUBJECT. For wards- addi info Ii e dismal feedwater line break,ana]ysis, 1

Info needcded .to close. confirmatory issue addressed in Section 15,10 ' of SER.

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FLORIDA POWER & LIGHT COMPANY December 14, 1982 L-82-542 Office of Nuclear Reactor Regulations Attention: Mr. Darrell G Eisenhut, Director

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Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Eisenhut:

Re: ST. LUCIE UNIT NO. 2 DOCKET NO. 50-389 SMALL FEEDWATER LINE BREAK ANALYSIS Attached please find additional information required by J. Guttman of your staff, which supplements the feedwater line break analysis submitted in Florida Power and Light Company letter L-82-533 dated December 9, 1982.

It is our understanding that this submittal completes the information required to close the confirmatory issue addressed in Section 15.10.2 of the Safety Evaluation Report.

If you have any questions regarding this submittal, please contact us accordingly.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems and Technology REU/RJS/JES/ok Attachment @0O1 cc: J.P. O'ei1 ly, Region I I Harold F. Reis, Esquire 82i2i4 82i22iOiOi 05000389' PDR ADQCK PDR PEOPLE... SERVING PEOPLE

REANALYSIS OF SMALL BREAK LOSS OF FEEDWATER INVENTORY EVENTS WI H THE LI ITING SINGL F ILUR 'AND Sl P R VAILABLE INTRODUCTION

~Per ose The purpose of'his reanalysis is to show that the results of the small break loss of feedwater inventory event with the limiting single failure and offsite power available produce maximum pressures less than 110$ of design.

Back round The loss of feedwater inventory event presented /<equi~ty demonstrates that breaks of all sizes, when combined with the loss of offsite power, produce maximum pressures well below 1205 of design. Based on the recurrence fr equences provided in Reference 3, the NRC has concluded that the 120'.0 of design maximum pressure criterion is appropriate for large break loss of feedwater inventory events, and small break loss of feedwater inventory events combined with the loss of offsite power. However, it must be shown that small break loss of feedwater inventory events with the limiting single failure and offsite power available meet the maximum pressure criterion of 1105 of design.

.In order to demonstrate compliance with this criterion, a reanalysis of small breaks with a modified methodology was required. The methodology used pre+; oddly is applicable to the full spectrum of break sizes. However, it is extremely conservative when applied to the smaller break sizes. As a result, a new method of analysis which is still conservative was developed, and is discussed in the following section.

' " "'- "- pipes Since the recurrence frequencies presented in Reference 3 apply to greater than 6 inches in diameter, the re~nalysis need only consider breaks less than approximately 0.20 ft . This is the same break size presented p~cv>oddly as the limiting break with the original methodology. Therefore, in the following se~tions "small" breaks refer to those which are less than 0.20 ft .

METHOD OF ANALYSIS Mathematical M The methodology used in "the reanalysis of small break loss of feedwater inventory events is the same as that applied and described pr'avsom ly with the exception of the treatment of steam generator heat transfer and reactor trip on steam generator low water level. predictions of steam generator heat transfer and level behavior are based on the model documented in References 5 through 8. As discussed below, this model is conservative when applied to the small break loss of feedwater inventory events.

Steam Generator Heat Transfer RCS pressurization is largely a function of the rate at which the ruptured steam generator's heat transfer decreases as its inventory is depleted. (The "ruptured," generator refers to +e @earn generator nearest the pipe break). Section 158.3AdNufiiedts the sensitivity of RCS pressurization to steam generator heat transfer.

behavior. The study verified that RCS pressurization is maximized by under-estimating the affected steam generator liquid mass corresponding to the initiation of heat transfer degradation (i.e.,

over-estimating the rate of heat transfer decrease). The original methodology took a simplistic and clearly conservative approach by assuming heat transfer degradation was instantaneous upon steam generator dryout. However, this approach is modified in order to more realistically predict the behavior.

A gradual heat transfer reduction is expected as the steam generator tubes are exposed to increasing void fractions which force the tubes from the normal nucleate boiling heat transfer regime into transition boiling and eventually into liquid deficient heat transfer. Transition boiling is anticipated when the local void fraction exceeds 0.9 (Reference 9). Liquid deficient heat transfer develops when local qualities approach 0.9. Under full power conditions and utilizing the steam generator model documented in References 5 through 8, the onset of these heat transfer regimes corresponds to steam generator liquid inventories of approximately 70,000 ibm and 35,000 ibm, respectively.

However, the referenced model conservatively ignores the transition boiling regime, thereby delaying heat transfer degradation until fluid conditions correspond to liquid deficient heat transfer.

Thererfore, the modified treatment of steam generator heat transfer b'ehavior is conservative, since it under-estimates the liquid mass associated with the initiation of heat transfer degradation.

Steam Generator Low Water Level Trip As discussed. in Section 158.3+the orig(nal loss of feedwater inventory event method credited low water level trip in the ruptured steam generator only after its liquid inventory had been depleted.

This assured conservative treatment of low level trip even if the loss of feedwater inventory event caused rapid steam generator depressurization (i.e., large breaks) and consequent swelling of the

downcomer level due to flashing of the downcomer liqu1d. However, for sufficientl ll breaks the steam generat essure remains constant or inc es prior to reactor trip and wncomer level swell will occur due to flash1ng. Therefore, in the reanalysis of small break loss of feedwater inventory events steam generator,low water level trip 1s credited with a larger liqu1d inventory remain1ng.

For the steam generators, the low level trip setpoint corresponds to a downcomer liquid level of approximately 24 feet above the tube sheet and a liquid inventory of over 70,000 ibm under full power conditions (based on the reference steam generator model). However, the reanalysis of small break loss inventory events conservat1vely delays low level trip of'eedwater unt11 heat transfer degradation begins with approximately 35,000 ibm of liquid rema1ning in the ruptured steam generator.

The NSSS response to the small break loss of feedwater inventory event with the limiting single failure and offsite power available, was modeled using the CESEC computer program described in Section 15.0. In addition, the input to the CESEC code was modified to account for the steam generator low level trip and heat transfer degradation methodology described in the previous paragraphs.

Input arameters and initial conditions The input parameters and initial conditions used to analyze the NSSS response are discussed in Section 15.0. The initial conditions for the principal process variables were varied to determine the set of initial condit1ons shown in Table /

In addition to conservatively delaying steam generator low level trip coincident with the assumed heat transfer degradation, the initial primary system pressure was adjusted to achieve, where possible, a coincident reactor trip signal on high pressurizer pressure. This maximizes the primary pressurization potential of the small break loss of feedwater inventory event, by maximizing the primary system pressure at the time of the reactor trip.

'As a result'f the evaluation method applied to the loss of feedwater inventory analysis, the only mechanisms for mitigation of the reactor coolant system (RCS) pressurization are the pressurizer safety valves, the reactor coolant flow and the main steam safety valves. The last two influence the RCS-to-steam generator heat transfer rate.

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There are no valve or main ere~ failures which can degrade st~afety valve capacity. Nor surizer safety ere any credible fyj]ures which can reduce steam flow to the ruptured steam generator.< >

A decrease in RCS to steam generator heat transfer due to reactor coolant flow coastdown can only be caused by a failure to fast transfer to offsite power or a loss of offsite power following turbine trip (i.e., two or four pump coastdown, respectively). Because offsite power 1s assumed to be ava1lable for th1s analysis, the failure to fast transfer 1s assumed following the turbine trip. This results 1n the coastdown of two reactor coolant pumps in diagonally opposite loops.

A )pectrum of small breaks, of size less than or equal to 0.20 ft, were analyzed using the methodology described in preceeding paragraphs to determine the limiting break the size. The results of this analysis are provided in Figure / which plots maximum primary pressure vs. break size. As can been seen, the limiting break size is the 0.20 ft break.

The reason that the largest break produces the most adverse pressurizat1on is due to the more rap1d degradat1on of heat transfer in the ruptured steam generator. The rate of heat transfer degradation is a major factor that determines the primary coolant pressur1zation of the event (i.e., the more rapid the reduct1on in steam generator heat transfer, the greater the primary pressurization). As was previously stated, heat transfer degradation is conservatively assumed to begin when the ruptured steam generator inventory decreased to 35,000 ibm. The larger break sizes require a shorter time interval to deplet'e this remaining inventory, resulting in a more rapid heat transfer degradation, and greater primary coolant pressurization.

Detailed results of this limiting break size are presented 1n the following section.

RESULTS dynamic behavior of the important NSSS parameters following the

'he small break loss of feedwater inventory event with the failure to fast transfer to offsite power following turbine trip is presented in Figures R 9'. - =

The sequence of events provided in Table Q, suomarizes the important results of this event

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(1) It should be noted that the coincident occurrences (failures)

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considered in Chapter 15 do not include spurious 1ndependent failure's, only consequential failures and pre-existing failures.

Accordingly, spurious closure of a main steam isolation valve is not considered credible during the loss of feedwater inventory event.

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A 0.20 ft2 ruptur in the main feedwater line is assumed to instantaneously inate feedwater flow to bot sm generators, and establish cr tical flow from the generator n a st the break at an initial rate of 1979 ibm/sec. This causes a deer ease in steam generator liquid mass as shown by Figure The br eak discharge enthalpy is assumed to remain that of saturated liquid until the ruptured steam generator empties, at which time saturated vapor enthalpy is assumed.

e The absence of subcooled feedwater flow causes a constant heatup and pressurization of the .steam generators during the first 26.6 seconds which reduces the primary-to-secondary heat transfer rate. Rising primary coolant temperatures and pressures result. Due to the temperature reactivity feedback during this period core power is reduced from an initial value of 1025 to 99.85 at 26.6 seconds.

At 26.6 seconds the ruptured steam generator produces a low water level reactor trip signal. This reactor trip signal is coincident with a high pressurizer pressure trip signal. ,Also at this time, heat transfer in the ruptured steam generator begins to degrade due to insufficient inventory. This degradation initiates a rapid heat up and pressurization of the reactor coolant system. At 27.5 seconds the reactor trip breakers open followed by an assumed instantaneous turbine trip. Immediately following turbine trip, the failure to fast transfer to offsite power occurs, resulting in'the coastdown of two reactor coolant pumps. These occurrences'urther aggravate the primary pressurization.

Closure of the turbine leaves the pipe break as the only steam relief path, thereby reducing the energy flow from the .intact steam generator below that of the primary-to-secondary heat transfer rate. The resulting steam generator pressurization reduces the primary-to-secondary temperature difference. In addition, the loss of reactor coolant flow following the loss of electrical power to two pumps decreases the heat transfer coefficient of the coolant in the steam generator tubes. A significant heat transfer reduction occurs.

Compr'ession of the pressurizer steam volume due to the high insurge flow raises the pr essure to the safety valve setpoint at 28.3 seconds. Thereafter, every increase in the surge flow causes a slight pressurization which opens the safety valves such that their volumetric discharge rate matches that of the insurge. At 30.2

..., ... seconds, the surge line flow reaches its maximum value of 1458 ibm/sec.

At this point in time, the reactor coolant system pressure is at a maximum of 2712 psia. Also, the increased pressure establishes a surge line pressure gradient which provides sufficient flow to allow the reactor coolant to expand under the existing heatup with no further pressurization. The rate of heatup decreases subsequent to core heat flux decay, causing primary pressures to drop.

At 30.0 seconds the main steam safety valves opened stabilizing the secondary side temperature and allowing the rising primary coolant

temperature to develop greater heat transrer ~o tne in~ac~ sredlll generator. The~et generator is forced to a m ximum of 1342 psia at 33.8 secondsgre the heat transfer begins crease. The core-to-steam generator heat rate mismatch fs re u d sufficiently by 37.4 seconds to allow closure of the pressurizer safety valves, and the reactor coolant system enters a cooldown. Under the influence of steam blowdown through the ruptured steam generator to the break, the cooldown proceeds even after the steam generator safety valves close.

After this point, a main steam isolation signal is generated on low steam generator pressure which closes the main steam isolation valves, decoupling the intact steam generator from the ruptured steam generator and the break. The intact steam generator repressurizes, thereby reducing its heat transfer and eventually causing a primary system heatup. With the main steam safety valves re-opening, the primary-to<<secondary heat inbalance is eliminated shortly thereafter. The HSSS enters into a quasi -steady state with a very gradual cooldown and depressurization due to decreasing core decay heat and with emergency feedwater flow maintaining an adequate liquid inventory within the intact steam generator for heat

. removal. By 1800 seconds the operator initiates a controlled cooldown to shutdown cooling utilizing the atmospheric dump valves.

C0NCLUS10~

This evaluation shows that the plant response to the limiting small feedwater line break event with the most adverse single failure with offsite power available produces a maximum RCS pressure which is within 110$ of design (2750 psia).

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References "USNRC Standard Review Plan, Section 15.2.8, Feedwater System Pipe Breaks Inside and Outside Containment (PWR)", NUREG-75/087, November 24, 1975.

R.E. Henry, H.K. Fauske, "The Two Phase Critical Flow of One-Component Mixtures in Nozzles, Orifices, and Short Tubes", Journal of Heat Transfer, Transactions of the AStlE, May, 1971.

"Response to NRC Round One guestion 440.42 on the CESSAR-FSAR".

R F ~l Sk~V ~>~yi:~ ~ep~~5 Q,~,.~ St~~~-~7+r CENPD-107 Su lement 1 "ATWS flodel modification to CESEC," September 1 . Section 3.0 .

CENPD-107 Su lement 1 Amendment 1-P "ATWS model modifications to CESEC,'ovember 197 . Section 3.3 .

CENPD-107 Su lement 3, "ATMS model modification .to CESEC," August 1975. Sections 240.8, 240. 11 and 240.9).

CENPD-107 Su lement 4 "ATWS model modification to CESEC," December 1975. Section 1.6, 1.8 and 4.2).

Forced Convection Hoilino Studies, Final Re ort on Forced Convection Va orization Pro'ect V.E. Schrock and L.N. Grossman, TID-14632 (1959).

TABLE ASSUMPTIONS FOR THE REANALYSIS OF THE LIMITING SMALL BREAK LOSS OF FEEDWATER INVENTORY EVENT Assumed Parameter Val ue Core Inlet Temperature, F 560 Core Mass Flowrate, 10 ibm/hr 164. 9 Reactor Coolant System Pressure, psia 2115 Steam Generator Pressure', psia 1026 CEA Worth for Trip, 10 =

ap -10.0 Pressurizer Safety Valves Rated Flow, ibm/hr 460,000 Initial Pressurizer Liquid Volume, ft 1120 Initial Steam Generator Inventory, ibm 173,000 Feedwater Pipe Break Area, ft 0.20

.Steam Bypass Control System .Manual Pressurizer Pressure Control System Manual Pressurizer Level Control System Manual 1 ~ ~

~TABLE SE UENCE OF EVENTS FOR THE REANALYSIS Of THE LIMITING SMALL BREAK LOSS Of FEEDPIATER INVENTORY EVENT Time Setpoint

~sec Event or Value 0.0 Rupture in the Main Feedwater Line, ft 0.20 0.0 Complete Loss of Feedwater to Both Steam Generators 0.0 Initial Steam Generator Break Flow, ibm/sec 1979 26.0 High Pressurizer Pressure Trip Condition Reached, psia 2475 26.6 High Pressurizer Pressure Trip Signal Generated 26.6 Low Level Trip Signal in Ruptured SG 26.6 Heat Transfer Degradation in Ruptured SG Begins 27.5 Reactor Trip Breakers Open 27.5 Turbine Trip on Reactor Trip 27.5 Failure to Fast Transfer - Two Reactor Coolant

.Pumps Coast Down v 27.8 CEAs Begin to Drop into Core

28. 3 Pressurizer- Safety Valves, psia 2525

30. 0 Main Steam Safety Valves Open 1282 30.2 Maximum Surge Line Flow, ibm/sec 1458 30.2

Maximum RCS Pressure, psia 2712 33.8 Maximum Steam Generator Pressure, psia 1342 36.8 Ruptured SG Dries Out 37.4 Primary Safety Valves Close, psia 2523

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0 2700 0-- -O DATA 2650 BREAK MAXIMUM AREA PRESSURE (SQ. FT.) (PSIA)

0. 01 2684 2600 0.05 2686

0. 10, 2697 0.20 2712 2550

'INCLUDES ELEVATION AND REACTOR COOLANT PUMP HEADS 2500

0. 05 0. 10 0. 15 0.20 BREAK SIZE, SQ. FT.

REANALYSIS OF SMALL LOSS OF Figure FEEDVJATER INVENTOR Y EVENTS MAXIMUMRCS PRESSURE's BREAK AREA

120 100 6

~ 80 60 F% 40 C)

C) o 20 0 10 20 30 40 TIlNE, SECONDS REANALYSIS OF SMALL LOSS OF FEEDWATER INVENTORY EVENTS

- LIMITINGCASE CORE POWER vs TIME

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'Cl 120 100 C) 80 60 OC 40 20 0

0 10 20 30 40 50 TIME, SECONDS REANALYSIS OF SMALL LOSS OF Figure FEEDWATER INVENTORY EVENTS - LIMITING CASE CORE HEAT FLUX vs TIME

DOPPLER MODERATOR CI CD

-5

~7

.CEAs ..

0 10 20 30 40 50 TDhE, SECONDS REAM/LYSIS OF SMALL LOSS OF FEEDN!ATER INVENTORY EVENTS - LIMITINGCASE REACTIVITIES vs TIME

670 650 OUTLET 630 I

610 AVERAGE 590 C) 570 550 0 10 20 30 40 50 TIME, SECONDS REANALYSIS OF SMALL LOSS OF Figure FEEDWATER INVENTORY EVt.NTS - LIMITING CASE CORE COOLANT TEMPERATURES vs TIME

C 24000 20000 LOOP WITH INTACT STEAM GENERATOR

> 16000 LOOP WITH C) RUPTURED STEAM

~ 12000 GENERATOR CD 8000 4000 0

0 10 20 30 50 TIME, SECONDS REANALYSIS OF SMALL LOSS OF Figure FEEDWATER INVENTORY EVENTS

- LIMITING CASE REACTOR COOLANT FLOW vs TIME

2900 2700 REACTOR COOLANT SYSTEM'RES

- 2500 SURIZER a 2300

>- 2100 1900

'DOES NOT INCLUDE ELEVATION OR REACTOR COOLANT PUMP HEADS i~ ~

1700 10 20 30 40 50 TIME, SECONDS REANALYSIS OF SMALL LOSS OF Figure FFEDWATF.R INVENTORY FVtNTS - LIMITING CASE PRIMARY SYSTEM PRESSURES vs TIME

P 1400 INTACT STEAM GENERATOR 1300 RUPTURED STEAM GENERATOR

~ 1200

~~ 1100 5 1000 C3 ~

~

900

'800 0 10 20 30 40 50 TIINE, SECONDS REANALYSIS OF SMALL LOSS OF Ft gut'e FEEDWATER INVENTORY EVENTS - LIMITINGCASE STEAM GENERATOR PRESSURES vs TIME

180000 150000 INTACT STEAM

~120000

~ GENERATOR 90000 RUPTURED STEAM GENERATOR 60000 30000 0

0 10 20 30 40 50 TIME, SECONDS REANALYSIS OF SMALL LOSS OF Figure FEEDWATER INVENTORY EVENTS - LIMITING CASE STEAM GENERATOR LIQUID MASS vs TIME