ML20125D869

Min Pressurizer Level for Various Reactor Trip Transients, Prepared for Toledo Edison

ML20125D869
Person / Time
Site:	Davis Besse, Three Mile Island
Issue date:	12/22/1978
From:	Winks R BABCOCK & WILCOX CO.
To:
Shared Package
ML20125D868	List: ML20125D865 ML20125D869
References
TASK-TF, TASK-TMR 86-272500, NUDOCS 8001160800
Download: ML20125D869 (9)

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Minimum Pressuri:er Level for Various Reactor Trip '

Transients e

by

' Robert Hinks Babcock & Wilcox Co.

Lynchburg, Virginia 24501

- D'ecember 22, 1978 Reviewed by:

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Date? l/2/22/?? /

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Approved by:

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[ f, l i g Date: / A r.M L /

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l CONTENTS Pace I 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . .

3 2.0 Objectives of the Analysis . . . . . . . . . . . . . . . . . . . . 4 ,

3.0 Conclusions ........................... 5 4.0 Summa ry of Analytical Method . . . . . . . . . . . . . . . . . . . 6 4.1 Tabulated Description of Actual Reactor Trip Transients at 03-1 ..................... 6 4.2 Relationship Between RC Pressure and Tave Following Rx Trips at 08-1 . . . . . . . . . . . . . . . . . . . . . . . . 7 4.3 Method of Analysis -- An Example Calculated ......... 8 5.0 Summary of Calculations ..................... 10 5.1 Case 1 -- Normal Reactor Trip Frem 1005 Power ........ 10 5.2 Case 2 -- Normal Reac:or Trip Frca 155 Power . . . . . . . . . Il 5.3 Case 3 - Reaccer Trip From 1005 With Loss of Main '

Feedwater . . . . . . . . . . . . . . . . . . . . . . . . . . 12 5.4 Case 4 - Reactor Trip From 15% With Loss of Main Feedwater . . . . . . . . . . . . . . . . . . . . . . . . . . 14 5.5 Case 5 -- Reactor Trip From 155 With Loss of Main Feedwater and RC Pumps (Natural Circulation Plus Auxilicry Feedwater) .................... ,

15 6.0 Limitations of Calculations ................... 16 7.0 List of References . . . . . . . . . . . . . . . . . . . . . . . . 17 List of Ficures l l

Ficure I

1. RC Pressure Versus Tave for Recent Reactor Trio Transients at Davis-Sesse 1 . . . . . . . . . . . . . . . . . . . . . . . . 18 l

2. RC System Pressure and Temperature During Specific Reacter Trips ac Davis-Eesse 1 . . . . . . . . . . . . . . . . . . . . . 19

3. Steam Generator Performance Durir.; Reacter Trio en 11/29/~7 .. 20

4. Chance in Pressuri:er Level During Reactor Trip and .

Initiation cf Auxiliary Fee: water . . ............. 21 i

5. Pressuri:er Level Af er a Eeac:Or Tri: snd L:ss of RC Pumps -'

With Initiation of Auxiliary Fee:.ta:er . . . . . . . . . . . . . 22

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1.0 'Intreducticn

Through:ut the startuo program for the avis-2 esse 1 plant a large number of a reactor trip transients have occurred. These transients have revealed neces-

sary adjustments in several of the auxiliary systems, such as higher reseating pressure of the steam relief valves and lower liquid levels in the steam gen-Cerator. In the near' future Toledo Edison Com;any plans to conduct three very significant "c:erational" transients in orde" to demonstrate proper operaticn of all related systems. They are: loss of electrical load frem full power, loss of all offsite pcwer with reactor at 15% power, and shutdown frem cutside the control room with the reacter at 15% power.

Previous reactor trip transients have revealed considerable overcooling of the reactor coolant system and a concern exists that excessive cooling of the primary system'might occur during scme of the planned reacter trip transients.

A request for analysis and calculations of minimum pressuri:er level during several types of reactor trip transients was made by Toledo Edison Company in order to understand the limits required to avoid emptying the pressuri:cr.

The scope of this report will be to describe several typical reactor tric transients, and to credict tha mininam nr=cenchor level. and to state any limitations en plant operatien to prevent emptying the pressuri:er.

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. . D VV C / M .) VV 2.0 Obieccives of thb analvsis:

Tabulate the specific reactor trip transient test data wnich was used to chn-plete the study of minimum, pressuriner level during selected reactor trip transients.

Show that F.C pressure will remain higher than 1650 psig during these selected reactor trip transients by incorporating actual Davis-Sesse 1 test data.

Show that the minimum pressurizer level for a normal reactor trip transient at any initial power level will be above a zero indicated level.

Show under what conditions minimum pressurizer level will remain above a zero indication on a reactor trip following a loss of main feedwater.

Show that minimum pressurizer level for a reactor trip coincident with less of station power will remain above a zero indicatien provided the coerator maintains a proper water level in the steam generator via the auxiliary feed-water system.

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%F We7 3.0 Conciusions .

s 1. For a normal reactor trio from full pcwer (Case 1) with main feedwater, pumps operating and the main feedwater sys em holding a 2 foot level in each steam generator, the pressurizer. level .will' remain on scale:provided rist$amgeneratorpressureisnolowerthan.980_psig.

2. For a normal reactor trip from 15% power!(Case 2) with main feedwater pumps operating the pressurizer level'will remain:en scale if steam pres-sure 'does noti drop below 980. psig. At the same time the cold log tempera-tures must not decrease below 545F in the first 60 to 90 seconds after the reactor trip.

3. For a reactor trip from low power levels with loss of main feedwater pumps s(Case 4) thelpressurizer level' will- just barely remain "on scale" if the.

[ auxiliary feedwater r.aintainsfonly a 30" level in each steam generator, and Othe control room cperator uses the' makeup pumps to simultaneously regain approximately 12" cf pressuri:er level.

4. For a reactor trip and loss of all RC pumps and =cin feedwater pumos (Case i 15)ithe pressuri:er level ~ will not drop "off scale." If the auxiliary  ;

feedwater system was used to generate a 100" startup level in each steam generator, (and steam pressure decreased to 770 psig) then pressurizer level would decrease to a zero indication.

5. iThe' 3 foot setpoint for startup level during auxiliary feedwater system l operation appears te be satisfactory in keeping pressurizer level above.

ia zero indication except. for Case 3, the reactor trip transient from full

' power due to a loss of main feedwater flow.

6. The performance of the auxiliary feedwater system on the loss of offsit:

power transient of November 29. 1977, has been appliad to both natural circulation and forced convecti:n situations for this :nalysis. By ccm-paring the test data of SpR #430, a reactor trip on March 1,1973, with the data of f cvember 29, 1977, it is clear that a: plying the :ovember 29th l auxiliary feedwater system performance to reactor trip transients with RC l pumps operating is more conservative (1cwer pressurizer levels) than ap-plying an appropriate auxiliary feedwater system ;:erformance correspond-ing to all P,C pumps operating.

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.,4 . 0 Su- ary of Analvtical v ethod The work performed in this re: ort was iargeted toward' preparing calculated RC oressuri:er levels at Cavis-5 esse 1 fer tne following kinds of reactor trips.

(Refer to available site problem reports which describe actual plant perfermance during reactor trip transients.)

Case No.

Reactor trio transient 1 Norma'l! reactor. trip (1005) with RC pumps and r.ain feedwater system ooerating. Normal control of steam pressure and water level in both steam generators.

2 ENormal reactor trip from 155; other conditions same as 1 above.

3 .I ossL of feedwater with reactor power at 100i ERC pumps'. are op-

' erating and theioperater manually controls final steam generator i: levels.using the auxiliary feedwater system.

4 i.Lo'ss: of feedwater with reactor power at 155. Other conditiens

'same as 3 above. l 5 Loss of. offsite oowere with reactor power at 155. RC pumos and

~ main feedwater cumps shutdown.

i; Operator manually controls the auxiliary feed..ater system.

1 The procedure used to analy e these five cases was as folleus.

1. Several recent site prcblem re orts were studied to determine the rela-tionship between RC pressure and T ave foll wing reactor trip at 08-1 (section 6.1). This relationship was used to determine proper scecific volumes for the primary coolant follcwing a reactor trio. An incidental use of this data was to verify that the reactor trips listed as Cases 1-5 previously will not actuate the Safety Features Actuation System.

2. Using the RC pressure - T ave relationships just discussed, the centrac-tion of the RC coolant af ter a reacter trip was determined, hence the -

level change in the ;ressurizer was calcula:ed. The.particular techni:;ue used is described in section 6.3.

4.1 Tabulated Description of Actual Reactor Trip Transients at CB-1 Table 1 presents recent reac:cr trip transients used to determine a T ave -

RC pressure relationships assumed in this report. Each event was reported by a site problem report (Sp .) numbered as shown in Table 1.

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? q 'o Rx;= Trip 1, SRP No. 40ates -

Erief Des:rirtion of Transient .__

476 8/2/78 Fo'llowinc a c:ntrol red insertien tes: a large neutr:n power error creve feed <: ster to overfeed ne steam gen-

.. erators and overcool ths RC system. After the trio SG levels were held at 40 incnes :y main feedwater.

484 9/28/78 At 955 pcwer a failed BY iP transmitter initiated a runback. Due to erroneous 1000 i and 2 RC ficw in-dications, feecwater flou :c each steam genera:Or was re-raticed. 0:crator over fed the steam generat:rs and overc0cled F.C5 causing a low :ressure trio. After the trip SG levels were held at 100 inches by rain feedwater.

485 10/3/78 A turbine trio occurred at 705 power level. Runback l procected f:r 30 secon:s before icw RC oressura  :

tripped reac:cr. SG icvels were hcid at aoproximately 4 feet by main feecwater following the trip.

431 2/24/78 Rx tripped by high reacter outlet temperature. Tur-bine tripped and main feccwater flo..s reduced te :ere in about'1 nir.uta. Steam ;enerator sisr;up lele!! l maintaintd cb:ve 50 and 75 in:hos in ic:ps #1 ar.d i2. l 1

435 4/2/78 A planned turbine trio test fr:m 755 ocwer leval.

Reactor pcwer runback proceeded fo. accroximatoly ?O 2 seconds. Reacter trio:ec on icw RC pressure dua to main feedwater overcooling primary side. .

396 ((11/29/771 At 405 power level all sta ien : uer was lost ar.d the reactor and RC curos vers snut: wn. L:ss of main feedwater pum:s se: uo the auxiliary feet a:er system .

and 10 foot and Sfoot levels were maintained, respec-tively, in the two steam generators.

4.2 Relationship Cetween RC Pressurs a-d Tayn Folicwir- Rx Tr4 s a: :3-1 .

The relationsnip Of r.inimum RC pressure for a minimum T 37, folicwing a reac-tor trip at Davis-Eesse 1 is shown in Ficures 1 and 2. Figure 1 cisolays :.e most recent rsac:cr trip transients and indicates a siigntly higr.er RC ;res-sure due to the imoroved cperation of the steam relief valves. Figure 2 shows that there have been instances of RC cressure decreasing to 1550 psic on s:me 1 reactor trip transients. This infor ation has been included to verify the minimum T va ues usec in the fellowing analysis. Figure 2 sh:ws a 1:ss ave 1

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,0f.RC, pumps cooldown as well as a normal RC pump operating c cidown of the Rdsystem. I 1

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4.3 Meth:d of Analysis - An 5xamole Calculation i

The initial mass of reacto'r coolan: immediately prier to a reactor trip tran-sient is determined by '<nowing' three RC s.ystem volumes and evaluating the proper specific voluInes of the fluid at reactor outlet and inlet temperatures and at saturation temperature within the pressurizer. '

For example: At 100", power level we knew the following information:

Tave 582F RC pressure 2170 psia T hot leg 605.5F T cold lec 559F Hot lec volume 5471 ft3 Cold leg volume 4955 ft3 Pressurizer level 200 inches Pressurizer vclume E64 ft3 3

^ The specific volume of the hot leg reactor coolant at 505.5F is 0.023491 ft /lb '

whereas the specific volume of the cold leg c clant at 559F is 0.021541 ft3 /lb.

~l he fluid containec in tne pressurizar is at saturation tercerature ht 21/U psia pressure so its spscific volume is 0.020525 ft3 /lb. The initial (and .

final) mass of the reactor coolan: is:

V Mo = hot + Vcoldt . Vc :r

" hot " cold "pzr 5471 4955 864

• 0.023491 + 0.021541

• 0.026625 Mo = 494,425 lbs.

It is conservative to predict a minimum pressuri:er level based on a c:ntrac-tion of the fluid in :he primary system by assuming no net addition of mass to the RC system (via makeup ficw) during the initial 50 seconds cr so of the reactor trip transient..

Now, by reversing the application of the above equation, it is possible to l

find the amount of reactor coolant remaining in the pressurizer at final or

" minimum" values of RC pressure and tem;eratures.

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[Fyr eyample, what is the minimum pressuri:er level corresponding to a reactor outlet tem: era:ure of 551F, a reactor inlet temperature of 550F and a minimum RC pressure of 1735 psia?  !

The proper specific volume values are: .  !

hot = 0.021513 ft /1b v 3 3

v cold = 0.021505 ft /lb 3

vp.er = 0.02443 ft /lb Solving for the final pressurizer volume; V V I

=/Mo - hot cold V

pzr v

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" cold, xvrp .

5471 4955 1 0.0243

= 494,435 -

0.021538 0.021505]x ,

V pzr

= 245 ft3 ,

The fin:1 pr:::uri::- icvel is:

AV pzr = 864 - 245 = 519 ft3 pzr level = 519/3.2 f t / inch 3

V

= 193 inches Final pressuri:er level = 200" - 193" or 7 in. (above the zero indication).

Whenever a reactor trip transient starts at 100", peuer and the primary fluid temperature dreps belcw 550F (and 1720 psig), then the minimum cressurizer level will nearly equal a zero indication cf level which is 75 inches above ,

the bottem of the pressuri:er. No operator action involving makeur ficwr te was assumed. Also, the corresponding pressure in each steam generator to generate a 550F RC system tem erature would be approximately 950 ;sig as-suming a 5F difference between Teold and saturation temperature.

On the following pages are displayed tables of cal'culated pressuri:er levels for the different case described on page 5.

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