ML19289E907

Evaluation of Transient Behavior & Small Reactor Coolant Sys Breaks in 177 Fuel Assembly Plant,790516. Vol 3, Revision 1

ML19289E907
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Site:	Davis Besse
Issue date:	05/19/1979
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O EVALUATION OF TRANSIENT BEHAVIOR AND SMALL REACTOR COOLANT SYSTEM BREAKS i IN THE 177 FUEL ASSEMBLY PLANT MAY.16, 1979 .

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VOLUME 3 RArsEn Loor PLANT (DAVIS BessE 1) 2052 130 -

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Babcock &Wilcox Revision 1 '

MaycQ,1979 29052904M i .'

EVALUATION OF TRANSIENT BEHAVIOR AND SMALL REACTOR COOLANT SYSTEM BREAKS IN THE 177 FUEL ASSEMBLY PLANT MAY 16, 1979 VOLUME 3 This volume supplements the infomation in Volumes 1 and 2

. to more completely reflect the unique characteristics of the Davis Besse plant.

Those portions of Volumes 1 and 2 which are truly generic and therefore already cover the Davis Besse plant are not repeated in Volume 3. Those portions are so noted in the Volume 3 Table of Contents. Where changes have been made specifically for Davis Besse, those sections or appendices have been either replaced or supplemented and are also so noted in the Table of Contents.

2052 W Revision 1 Revised May 19,1979

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( VOLUME 3 TABLE OF C0tRENTS

1.0 INTRODUCTION

Supplementary Infomation 2.0

SUMMARY

AND CONCLUSIONS Volume 1 is generic.

3.0 TMI-2 INCIDENT BENCHMARKS FOR CADDS AND CRAFT Volume 1 is generic.

4.0 LOSS OF FEEDWATER SAFETY EVALUATION New section added.

5.0 THE SMALL BREAK PHENOMENA - DESCRIPTION OF PLANT BEHAVIOR New section added.

Is 6.0 SMALL BREAK ANALYSIS New section added.

Accendices 1 Natural Circulation in Operating B&W Plants Volume 2 is generic.

2 Steam Generator Tube Themal Stress Evaluation Volume 2 is generic.

3 Restart of RC Pumps in a 50% Voided System Volume 2 is generic.

4 Operating Guidelines for Small Breaks - Part II New section added.

5 Michelson. Report Assessment Volume 2 is generic.

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( l.0 INTRODUCTION The information contained in Section 1 of the report, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant," dated May 7,19/9, is applicable to Davis Besse.

However, the following additional information is provided to discuss important distinguishing features of the Davis Besse plant.

1.1 Looo Arrancement The raised loop arrangement for the Davis Besse plant is shown on Figure 1-1, while the lowered 1000 arrangement is shown on Figure 1-2. While the RCS volumes are approximately the same for the two plant types, the Davis Besse arrangement has the majority of the inventory in the loops above the elevation of the reactor core.

The Davis Besse arrangement assures a longer time to core p uncovery for a given size break in the event emergency coolant is not provided.

1.2 Makeuo/Hich Pressure Iniection The Davis Besse plant employs two makeup pumps and two high pressure injection pumps as compared to the other 177 FA jlants where the makeuo pumps also serve as high pressure injection pumps.

In the event an ESFAS signal is received, the high pressure injection pumps are actuated. For Davis Besse, HPI is actuated at an RCS HLC fressure

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of IkO5[ Psi { For compar1' son, che flow cl$a5cteristic_s assumel'in [.

the ECCS analyses for Davis Besse and the B&W lowered loop plants are shown in Figure 1-3. The actual characteristics for a Davis Besse HPI pumo are shown in Figure 1-4. As can be seen, the Davis Besse high pressure injection pump provides higher flow rates for pressures lower than 1400 psig.

To increase the total delivered head of the pumps, the LPI pumps can be lined up to provide suction to tne HPI pumpr. This line-up 2052 133

(typically called piggy-back) would increase the shutoff head f rom 1630 psig to 1830 psig. -

The makeup pumps which can also be powered from on-site diesel power could be used if required to supply makeup flow at pressures higher than the shutoff head of the HPI pumps. The Davis-Besse makeup pu=p flow characteristic curve is shown on Figure 1-5. Also shown are the makeup pump /HPI pump head characteristics for the lowered loop plants.

1.3 Auxiliarv Feedwater The auxiliary feedwater system is a safety grade system consisting of two turbine driven pumps each of which is rated at 1050 gym at 1050 psig (including 250 gpm recirculation flow) . The pu=ps are lined up so that one pump supplies only one generator. This system is used exclusively for accident and transient mitigat. ion and not for normal unit startup and shutdown.

On the Davis-Besse plant, with the raised loop arrangement, the level in the steam generator is maintained at 120 inches of 550 F water (96 inches indicated)on the startup range instrumentation if there is SFAS incident level 2 initiation; otherwise, the steam generator level is maintained at 35 inches of 550 F water on tee ~startup singiiristru::ientation.

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2.6 Su=arv and Conclusions - , /

f-- The inf or=ation in Section 2 of Volume 1 of " Evaluation of Trcpient Behavior and Small Reactor Coolant Syste= Breaks; in the 177 Fuel Asse=bly Plant," dated May 7, 1979, is generic in natute nnd is therefore applicable s 1

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and ccall Reacter Coolant Syste= Breaks in the 177 Fuel Assembly Plant.," dated'hy 7, 1979, ~ ts ge:arie in nature and is therefere applicable to mDavif'Beese.

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4.0 LOSS OF FEEDWATER SAFETY EVALUATION Note: The infcrmation below in Section. 4.0__ ,_ extends the information cor.tained in Volume 1. To aid tF- reader, those areas which have been revised or added specifically for Davis Besse are noted in the right hand margin.

, 4.1 Response of the B&W System The B&W NSS is designed to accommodate certain secondary system upsets, such as turbine trip and loss of one feedwater pump, without a reactor trip. Therefore, directly acting protective functions to trip the reactor on loss of feedwatar and/or turbine trip were not provided. Instead, the reactor is protected from overoressuri:ation during loss of feedwater events by 9 reactor trip which functions on high primary system pressure. As originally designed, this rerttor trip was set at 2355 psig (relative to a nominal operating pressure of 2155 psig), On a loss of main feedwater, reduction of secordary side cooling causes a primary side pressure and temperature increase which results in a rise to the reactor trip setpoint in approximately 8 seccnds, whereupon the raactor protection system promptly shuts down the reactor terminating the initial pressure rise. .

In addition, the design makes use of a small pilot operated relief valve (PORV) at the top of the pressurizer. This valve, in the original design configuratier., was set to operate at 2255 psig, 100 psig below the reactor trip setpoint. Thus, this valve actuated en each overpressure transient which results in a high pressure

' reactor trip and specifically it actuated for each loss of feedwater

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event in the B&W plant.

2052 142 i

Following the assessment of the TMI-2 incident of March 28, 1979, changes to reactor trip and PORV relief setpoints were made to eliminate the likelihood of p0RV actuatien in less of feedwater events and other anticipated transients which have bean observed in B&W plants. These changes and their analytical bases are described in Section 4.2 below for the lowered loop plants.

An additional analysis was made of the Davis Besse raised loop configuration. The loss of feedwater transient was analyzed for 08-1 and the results confirm the conclusions reached for the realistic lower loop model case.

The FSAR case shows that for a RC high pressure' trip setooint of 2300 psig, the peak iressurizer reached is less than 2500 psig for both lowereo loop and raised plants. The analytical bases for the raised loop analysis are also described in Section 4.2.

Setpoint revisions described in Section 4.2 essentially eliminate the possibility of PORV actuation in loss of feedwater events if the auxiliary feedwater system functions normally. However, if the injection of auxiliary feedwater to the steam generators is delayed, post-reactor trip decay heat will once again cause the primary coolant system temperature and pressure to increase toward the setpoint of the pressurizer relief vcive. Section 4.3 below contains a parametric study of the effect of delay in auxiliary feedwater initiation following loss of feedwater events.

An imeciate (anticipatory) reactor trip on loss of feedwater or turbine t.ip can provide additional time before auxiliary feedwater is required following a loss of main feedwater. Section 4.4 below presents analyses of the effect of tripping the reactor immediately on loss of main feedwater.

7.052 W

Finally, Section 4.5 presents event tree analyses of reactor coolant system expected benavior following a loss of feedwater event and assuming various failures of control and safety equioment.

4.2 PORV and Hich Pressure Trio Setooint Studv A number of alternatives were considered for providing assurance that the PORY will not be actuated during anticipated transierits which have occurred or have a significant probability of occurring on B&W nuclear steam systems. The alternatives include:

1. Restricting reactor power to a value which would assure no actuated of the POEV. The reactor protection system, design Pressure, and PORV setpoints remained at their current values.

2. Lowering the high-pressure reactor trip setpoint to a value which would assure no actuation of the PORV. The design pressure

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of the reactor and the setpoint for PORV actuation remained at their current values.

3. Lowering the high-pressure reactor trip setpoint and adjusting the operating pressure (and temperature) or the reactor to assure no PORY actuation and to provide adequate margin te accomodate variations in operating pressure. The setpoint for PORY actuation remained at its current value. This alternative would reduce net electrical output.

4. Adjusting the high-pressure trip and the PORV setpoints to assure no PORV actuation for the class of anticipated events of concern.

The design pressure of *he reactor remained at its currert value.

An analysis of the impact of these various alternatives and tneir contribution to assuring :nat the PORV will not actuate for the class of anticipated transients of concern has been comoleted. The results show that the combination of decreasing the high pressure 3

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reactor trip setpoint and increasing the PORV setooint provides the required assurance.

A sensitivity study was performed to identify combinations of initial RCS operating pressure, hign-pressure trip setpoint, and PORV setpoint wnich will result in reduced probability of PORY actuation following anticipated transients. The anticipated transients of concern are:

1. Loss of external electrical load.

2. Turbine trip.

3. Loss of main feedwater.

4. Loss of condenser vacuum.

5. Inadvertent closure of r:ain steam isolation valves (MSIV).

_ The loss of external electrical load, loss of condenser vacuum, and inadvertent closure of the MSIV's all give secondary pressure increases equivalent to, or less severe than, a turbine trip.

Therefore, turoine trip and loss of main feedwater bound these ,

anticipated transients.

Anticipated transients which have not occurred at B&W plants (low probability eveMs) are: (1) low-worth rod group withdrawals, and (2) moderator dilution accidents. The moderator dilution accident can result in a high-pressure trip, but FSAR analyses show that peak pressures are well-bounded by loss of main feedwater transients. Some rod group withdrawals can be shown to conservatively result in peak pressures exceeding the pressurizer safety valve setpoint. However, these events have a very low probability of occurrence wnich assures that the recomenced setpoint changes are effec *ive for truly anticipated transients, and these events can be excluded from further consideration.

4

?052 145

The objective of reducing the frequency of PORV actuation can be reached by increasing the PORV setpoint, increasing the RCS nominal operating pressure, decreasing the high-pressure trip setpoint, or through a combination of these three adjustments.

Since these three adjustments are not all equally desirable, a parametric study was perfomed to allow for a setpoint selection which would be the most ~ desirable.

For each of the two events analyzed (turbine trip and loss of feedwater), peak RCS pressure following the event was obtained as a function of high-pressure trip setpoints for three setpoints, namely 2255 psig, 2305 psig, and 2355 psig. This parametric study was performed for three different initial operating pressure values, 2155, 2105, and 2C55 psig. For each initial pressure assumed, a corresponding average temperature was selected which would maintain

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the initial condition DNB margin at the value for which the plant was licensed. All analyses were performed assuming no PORY actuation.

The parametric study described above was performed twice, once utilizing a realistic model, then utilizing the FSAR model. The differences between the two models utilized are listed in Table 4.2-1. The carameter values used in the realistic model correspond to the TMI-2 benchmark case d'e scribed in Section 3.2. The results of the analysis are presented in Figures 4.2-1 through 4.2-4. The figures consistently show that, for a given high-pressure trip setpoint, the peak system pressure following the event increases with decreasing initial RCS pressure values. This effect is due to the longer time necessary to reach the trip setpoint from a lower initial pressure, thus resulting in a larger pressure overshoot. 'This result suggests that the operating pressure should not be changed (this is also desirable C

2052 146

from a safety standpoint because no accident analyses have been performed for RCS pressure lower the.a no=inal plus errors) . For the lowered loop plant, {

the combination of a lower high-pressure trip setpoint (2300 psig) and a higher PORV setpoint (2450 psig) will assure the desired results of no PORV actuation following turbine trip and LOFW, as evidenced by the peak pressure curves for expected (realistic) conditions. A typical pressure trace for a loss of main feedwater event is shown in Figure 4.2-5.

The second important point derived from this study for the lowered loop I

plants is that for initial RCS pressure of 2155 psig, RC high pressure trip setpoint of 2300 psig, and PORV setpoint of 2450 psig, neither the-PORV nor the code safety valves (2500 psig) are challenged under realistic assumptions. This result also holds true for the code safety valves under FSAR conservative assumptions. Thus, the revised setpoints for reactor trip on high RC pressure and the PORV are effective in reducing the frequency of possible safety valve lif t following anti-cipated transients.

An analysis has been performed to determine the sensitivity of the peak pressurizer pressure during a loss of feedwater (LOFW) transient to the high reactor coolant (RC) pressure trip setpoint for a raised loop plant (DB-1). The analysis is an exact duplicate of that described above except for differences due to raised loop design and plant specific RPS errors that were used in the FSAR case.

Because Davis-Besse Unit 1 is a raised loop plant, the primary coolant volume of the hot and cold legs will differ from TMI-2. As a direct result of volu=e differences, the flow transient times (i.e., time for coolant to ficw from reactor outlet to inlet of steam generator, 2052 147 6

etc.) will differ. The pressure drop between core outlet and pressurizer will also be different, due to the raised loop configuration. FSAR and realistic cases were considered using the same definitions as established above for the lowered loop analyses. The FSAR case considers instrumentation errors while the realistic case does not. For DB-1, the high RC pressure trip setpoint error is 16 psi, while for other 177 fuel assembly plants, the error is assu=ed to be 30 psi.

The results of the DB-1 analysis are shown in Figures 4.2-6 and 4.2-7 for the realistic and FSAR cases respectively. The figures show the relationship between the peak pressurizer pressure and the high RC pressure trip setpoint for a LOFW transient.

For comparison, the results from Figures 4.2-1 and 4.2-2 are also shown on Figures 4.2-6 and 4.2-7. In the analysis for Figure 4.2-7, the pressurizer code safety valves were set at 2500 pslg for all cases to show under what conditions the PORVs would open. Figure 4.2-6 shows that using a high RC trip setpoint of 2300 psig, the peak pressurizer pressure will be approximately 2350 psig. This provides sufficient margin to the actuation of the pilot operated relief valve (PORV) which opens at 2400 psig. Figure 4.2-7 (FSAR-conservative) shows that the pressurizer safety valves (PSV) will actuate if the PSVs are set below 2480 psig when the RC high pressure trip setpoint of 2300 psig is used. If credit is taken for the PORV opening, then the peak pressurizer pressure can be reduced

_ to 2433 psig for the high RC pressure trip.setpoint of 2300 psig.

2052 148 7

4.3 Parametric Study for AFW Delay Davis-Besse 1 is designed with a redundant safety grade Auxiliary Feedwater System; therefore, loss of all feedwater is a highly i= probable event. However, in the extremely unlikely case that all feedwater is lost for a period of time, the basic system responses provided in Section 4.3 of Volu=e 1 of " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Asse=bly Plant", dated May 7, 1979, would be applicable to Davis-Besse.

4.4 Anticipatory Reactor Trio The information contained in Section 4.4 of Volume 1 of " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant", dated May 7, 1979, which demonstrates the typical system responses to anticipatory trips, is similar for Davis-Besse.

4.5 Potential Loss of Feedwater Accident Sequences in order to examine the ef fects of potential f ailures, including a total auxiliary feedwater delay, following a loss of main feedwa?.er, event trees have been prepared for the caser in which main reactor coolant pumps are available (Figure 4.5-1) and in which reactor coolant pumps are off (Figure 4.5-2).

In a normal loss of main feedwater event, a reactor scram occurs shortly after the loss of feedwater and the safety grade auxiliary feedwater is initiated within 40 seconds. The auxiliary feedvater control system controls heat removal and the reactor system stabilizes in the hot shutdown condition -.:hin a few minutes. This nor=al path is shown in the left hand side of Figure 4.5-1.

The effect of an auxiliary feedwater safety grade level control system single failure would only affect one stea= generator and result in overcooling. This overcooling could lead to contraction of the RCS and a loss of pressurizer level indication. If this occurs, the R1S -6ald become depressurized and the high pressure injection system would be initiated 7052 149 8

automatically on detection of the low RCS pressure condition. The high pressure injection system would add water to the primary system and refill the pressurizer uhereupon the operator could take control and restore pressurizer pressure control and terminate the transient. This path is shown on the left hand side of Figure 4.5-1.

The path for a starting-system failure of one train of the safety grade auxiliary feedwater system is also shown in Figure 4.5-1. This path also leads to acceptable results.

The right hand side of Figure 4.5-1 is illustrative for the case where the two redundant safety grade auxiliary feedwater trains are arbitrarily considered failed. Normally AFW would always automatically be available to at least one steam generator even under single failure conditions. In the total auxiliary feedwater delay case, after about three minutes, the PORV will open at its setpoint of 2400 psig. If auxiliary feedwater is restored ,

promptly af terwards and the PORV reseats properly, the transient will be terminated in the normal way. If the PORV does not close properly, primary system pressure will f ail rapidly and high pressure injection system will be actuated automatically. This again will lead to a small break LOCA safely within the system design capability as bounded by analyses presented in the next stetion.

As Figure 4.5-1 indicates, the pressurizer will fill in about eight minutes of further total auxiliary feedwater delay. Primary system inventory loss up to about 30 minutes of total auxiliary feedwater delay without high f pressure injection can be acco=modated before core damage is likely to occur.

Analyses presented in the next section of this report cover this case. If auxiliary feedwater is actuated within 30 minutes, the transient can be ter- l minated without core damage.

Figure 4.5-2 shows a similar analysis for loss of feedwater event accompanied by a loss of power and, therefore, a loss of forced RCS flow. A normal system response again shown at the left hand side of the figure provides for auto-matic initiation of auxiliary feedvater within 40 seconds and control of 9 7052 150

auxiliary feedwater level to 35 inches of 550 F water on the startup range. As discussed in a later section of this report, this has been shown by test and actual plant experience to yield excellent core cooling by natural circulation. The reactor system resumes stable natural circulation shortly after the event. As before, a single train failure of AFW would lead to acceptable results.

The right hand side of Figure 4.5-2 shows scenarios for significant delay in the initiation of both safety grade auxiliary feedwater trains. As with the l previous figure, these can terminate in a small break LOCA if the pressurizer relief valve is actuated and fails to close during the course of the subse-quent event. As before, this small break is within the capability of the system to handle safely as demonstrated by the analyses presented later in this report. As before, a period of auxiliary feedwater interruption without high pressure injection of up to 30 minutes can be sustained without core damage. Initiation of auxiliary feedwater at or before this time leads to transient termination by design analyses.

4.6 Conclusions .

The analyses and discussions in Section 4 on this report lead to the following conclusions:

1. With adjustments which have been made to the setpoints of the pressurizer pilot operated relief valve and the setpoint of the high reactor coolant system pressure trip, PORV actuation is not expected for anticipated transients. This very significantly reduces the probability that improper operation of the PORV will lead to a loss of coolant event.

2. Although Davis-Besse has a redundant safety grade AFW system, the effects of the delay of both trains has been examined. Analysis shows that initiation of auxiliary feedwater at 120 seconds (three times the design value for auxiliary feedwater initiation time) does not lead to repressuri-zation to the PORV or pressurizer code relief valve setpoints.

2052 151

auxiliary feedwater level to 120 inches of 550 F water (96 inches indicated) on the startup range. As discussed in a later section of this report, this has been shown by test and actual plant experience to yield excellent core cooling by natural circulation. The reactor system resumes stable natural circulation shortly after the event. As before, a single train f ailure of AFW would lead to acceptable results.

The right hand side of Figure 4.5-2 shows scenarios for significant delay in the initiation of both safety grade auxiliary feedwater trains. As with the f previous figure, these can terminate in a small break LOCA if the pressurizer relief valve is actuated and f ails to close during the course of the subse-quent event. As before, this small break is within the capability of the system to handle safely as demonstrated by the analyses presented later in this report. As before, a period of auxiliary feedwater it.terruption without high pressure injection of up to 30 minutes can be sustained without core damage. Initiation of auxiliary feedwater at or before this time leads to transient termination by design analyses.

4.6 Conclusions The analyses and discussions in Section 4 on this report lead to the following cenclusions:

1. With adjustments which have been made to the setpoints of the pressurizer pilot operated relief valve and the serpoint of the high reactor coolant system pressure trip, PORV actuation is not expected for anticipated transients. This very significantly reduces the probability that i= proper operation of the PORV will lead to a loss of coolant event.

2. Although Davis-Sesse has a redundant safety grade AFW system, the effects of the delay of both trains has been examined. Analysis shows that initiation of auxiliary feedwater at 120 seconds (three times the design value for auxiliary feedwater initiation time) does not lead to repressuri-sation to the PORV or pressuriter code relief valve setpoints.

2052 152 g

Thus, a significant margin for manual backup to the safety grade auto-matic auxiliary feedwater initiation systems exists before actuation of the PORV is expected. If both safety grade trains of auxiliary feed-water are delayed even longer, filling of the pressurizer due to thermal swell of the reactor coolant water, requires approxi=ately 10 minutes.

3. Provision of anticipatery reactor trip on loss of main feedwater provides additional margin for auxiliary feedwater actuation before RCS repressuri-zation and pressurizer filling.

4 Finally, a systematic examination of potential f ailures in mitigating systems following a loss of main feedwater event, shows that in all cases in which auxiliary feedwater is supplied within 30 minutes of the loss of main feedwater event, the transient can be safely terninated without core damage.

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. 5.0 The Small Break Phenenena - Descriotion of Plant Behavior

-This section was-rewritten for Davis-Sesse-Unit ~1'.

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f::= :he reae::: :ocian: s.ys: e:. Due :: -he loss Of cass fre: the rea: :: :::Lan:

syste=, :he : s: significan: shcr:-:er: sy=p::= cf a les s-of-coelan: a::iden: is an us:en::elled redu::ier in -he rea:::: eccian: syste: ;;es sur e. The rea::::

e

:100 syste: is designed :: ::1; :he rea:::: c *c v ;; essure. This should occur before -he reac:c coolan: syster reaches sa:ura:10: conditions. The existence of saturated :: di:10:s withi the reac:er sys:e= is the ;;incipal 1::ger-ter: indica:ic Of a LOCA and requires special :::sidera:10: in :he d ev elopmen: ef Opera:ing p:ccedures, yelleving a reacter ::1p, it is ne:essary :: re=cve decay hea: f := the rea:::: cre to ;;even; da age. However, so iceg as the reae::: ecre is kep:

covered vi:h : cling va:er, core damage vill be avcided. The ICOS syste=s are (7

designed :c respo d au ::4tically :c 1:v reactor :ocian: pressure eccdi:1ces and take :he initial a::1ces to prote:: the reae::: core. They are sized to p:: vide sufficie:: va:e: :o keep the reae::: :::e ecvered even vi h a single failure in

he ICOS syste=s. Subsequent opera:c acticas are required ul:ima:ely :e place

ne plan: in a 10 g-:er: =coling ode. The overall obje :ive of the au:c:a:1:

e=ergency :: e ecoli:3 sys:e:s and the followup cpera::: a::icns is :: kee: the reae::: :ere coel. .

A de: ailed discussics of small break phenecesalogy is ;;esen:ed in :his se::icn. This discussion represents Par: 1 of -he opera:ing ;;ocedure guidelines f: :he develepten: Of detailed perating pre:edures. Par: 1 ;;esents the :::e de: tiled step-by-ste; guidelines and is included in Appendix .

The response Of :he ;;irary sys:e: :: a stall breah vill dif f er grea:17 depend' ; : -he breac si:e, i:s 1 :a:10 in :he sys:er. Opera:i:: :f :he rea::::

an: pu ps. :he sc:ber :f E005 ::ains fun::1:21:3, and -he evailability Of se:: darr s:de :::11:5 yigures 1 a:d 1 she-1 ;1::s of 305 ;; essure and

' 2052 164

pressurizer *evel histories f:: various :::bina:10:s Of parameters, indi:ating the vide 7t:ge Of behavi:: $hi:5 is possib~ e.

b 3.1 S:21*_ 3:eaks vi:5 Auxiliar- 7eedvater ev eu ev '

a. 23 There are four basi: : lasses of break response f:: reall breaks vi:h auxiliary f asava:er; :hese are:

~

1. LOCA large enough :: depressuri:e :he reae::: :o:.an; syste=.

2. LOCA which stabilizes at apprexi=a:ely see: dary side pressure.

3. LOCA whi:h may repressuri:e in a satura:ed conditice (no RC _ pumps)' _

4 S=all LOCA which stabilizes at a pri=ary sys:e: pressure greater than secondary syste: *:T essure.

The sys:e= ::ansients for :hese breaks are depi::ed in rigure 1.

3.1.1 LOCA Laree Incu:5 :: Deeressurize Reseter Co lan: Svste=

Curves 1 and 2 of yigure 1 show the respense of ROS pressure :: breaks tha: are large enough in :::bina: ice vi:h the IC05 :c depressuri:e the syste: := a

(%

stable icv pressure. ICOS inje:: ice easily exceeds :ere beil-off and ensures :::e cooling. Curves 1 and : Of yigure 2 shev the pressurize: leve. ::assien:. Rapidly f alling ;; essure causes :he he legs to satura:e quickly. Cold leg :e:pera:ure reaches sa:ura:ic somewha: late; as RC pumps coas: down c: :he RO'J depressuri:es belov :he se:::dary side saturation pressure. Since these breaks are capable of depressurizing the RCS vithout aid of :he s:ca: genera:ers, they are essentially unaffe::ed by the availabili:y of auxiliary feedva:er. Operation of :he RC pu=ps als: plays lit le role in the course of events. Other than verify 1:3 that all ISTAS a::icns have been ::=;1e:ed, the operate: seeds :: take :: a :10:s te bring

he sys:e= := a safe stable :: 41:10 . Rapid depressuri:a:1:n Of the s:ea:

genera:::s veuld ::.17 ac: :: a::elera:e ROS depressurizati .  :: is, however :::

necessary. 5:ar:i:; :he RC pu=;s :::e : hey are 1:s: is :: desirable f:: :his

lass of break.

L :g-:er: :::.ing vil; re uire :he Opera::: :: shif: he Ly pus; su::ier

:he rea:::: ::::ainmen: building sump.

2052 165 z

5.1.2 LOCA Which Stabilizes at Approximately Secondary Side Pressure Curve 3 of Figure 1 shows the pressure transient for a break which is too small in combination with the operating HPI to depressurize the RCS. The steam generators are therefore necessary to remove a portion of core decay heat. If the reactor coolant pumps are not operating, and the pressure has stabilized near the secondary side pressure, RCS pressure may eventually begin falling as the decay heat level decreases. If the RC pumps are operating, pressure may or may not decrease. System pressure could ultimately increase to some stable level as the HPI refills and repressurizes the RCS. The assumption for this case is that secondary cooling is maintained. Curve 3 of Figure 2 shows pressurizer level behavior. Curve 6 of Figure 2 shows refilling by the HPl. The hot leg temperature quickly equalizes to the saturated temperature of the secondary side and controls primary system pressure at saturation. The cold leg temperature remains slightly subcooled. If the HPI refills and repressurizes the RCS, the het legs can become subcooled. The i= mediate operator action is to verify ESFAS functions, the steam generator level has automatically gone to 120 inches of 550 F water, and, if RC pumps have not f ailed, leave only one RC pump running in each loop.

Follow-up action if RC pumps have been lost is check for natural circulation.

This is done by gradually depressurizing the steam generators. If this test is failed, intermittent bumping of an RC pump should be performed as soon as one is available. Continued depressurization of the steam generators with either forced or natural circulation leads to cooling and depressurization of the RCS. The operator's goal is to depressurize the RCS to a pressure that enables the ECCS to exceed core boil-off, possibly refill the RCS, and to ultimately establish long-eara cooling.

2052 166 3

5.1.3 LOCA Which May Repressurize in a Saturated Condition (No RC Pu=es)

Curve 4 of Figure 1 shows the behavior of a small break in which the RC pu=ps have been lost and the break is too snail in co=bination with the HPI to depressurize the primary system. Although steam generator feedvater is available , the loss of pri=ary system coolant and the resultant RCS voiding eventually lead to interruption of natural circulation. This is followed by gradual repressurization of the pri=ary sys:cm. Once enough inventory has been lost from the primary syste= to allow direct steam condensation in the regions of the steam generstors contacting secondary side coolant, the pri=ary system is forced to depressuri e to the saturation pressure of the secondary side.

Since the cooling capabilities of the secondary side are needed to con:inue to remove decay heat, RCS pressure vill not fall below that on the secondary side. EPI flow is sufficient to replace the inventory lost :o boiling in the core, and condensation in the steam generators re= oves decay g- heat energy. The RCS is in a stable ther=al condition and it will remain

('

there until the operator takes further action. The pressurizer level response is characterized by Curve 3 of Figure 2 during the depressurization, and Curve 4 of Figure 2 during the te=porary repressuri:ation phase. During this tran-sient, hot leg temperature vill rapidly approach saturation with the initial syste= depressurization, and it will re=ain saturated during the whole transient.

Cold leg temperature vill approach saturation as circulation is lost, but mSy remain slightly subcooled during the repressurizati:n phase of the transient.

Later RCS depressurization could cause the cold leg te=peratures to reach saturation. Subsequent refilling of the primary syste: by :he EPI =ight cause te=porary interrup:1on of steam condensation in the steam generator as the primary side level rises above the secondary side level. If the depressuriza-tion capability of the break and the EPI is insufficien: to offset decay heat.

the pri=ary syste: vill once more repressurize. This decreases EPI flow and inc.reases loss through the break until enough RCS coolan: is lost to once 7052 167

more allow direct steam condensation in the steam generator. This cyclic behavior will stop once the EPI and break can balance decay heat or the

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operator takes some action.

The operator's i= mediate action is to verify completion of all ESFAS functions. Following that, he should ensure thit $hesteamgenentorlevelis at 96 inches on the startup range and check for natural circulation. If it is, positive, he should depressurize the stea= generators, cool and depressurizs s

the primary syste=, and atte=pt to refill it and establish long-term cooldng.

If the system f ails to go into natural circulation, he should open the PORV long enough to bring and hold the RCS near the secondary side pressure.

Once natural circulation is established or an RC pu=p can be bu= ped, he will be able to continue depressurizing the RCS with the steam generators and es-

~

tablish long-ter cooling.

5.1.4 Small LOCA Which Stabilizes at P > P

. seg Curve 5 of Figure 1 shows the behavior of the RCS pressure to a break in which RC pu=ps are lost and cooling is accomplished by natural circulation.

Eigh-pressure injection is being supplied and exceeds ths leek flow before the pressurizer has e=ptied. The primar/ system re=ains subeccler and natural circulation to the steam generator removes core decay hea. , The pressurizer never empties and continues to control pri=ary systea pressure. The operstcr needs to ensure that ISFAS actions have occurred. 7.aere is no need to thrc!.tle HPI flow since the EPI system is incapahic of injecdicts coolant inta the primary system when the primary systa: pressure erceeds ^i708_ psig'.

5.1.5 Small Breaks in Pressurizer The system pressure transient for a s=all break in s the pressurizer will behave in a =anner si=1lar to that previously discussed. ihe in:.tial de-pressurization, however, will be more rapid as the initial invennry loss ,

is entirely staa=. The initial rise in pressurizer level shrun in Figure 3 will occur due to the pressure redue:1on in the pressurizer ard an 1.nsurge of

(

r 2052 168 ,

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r ccolant into the pressuriner from the RCS. Once the reactor trips, syste con-traction results in a decreasing level in the pressurizer. Flashing will ulti=ately occur in the het leg piping and cause an insurge into the pressurizer.

/

This ultimately fills the pressurizer. For the re=ainder of the trassient, the pressurizer will re=ain full. Toward the later stages of the transient, the pressurizer =ay contain e two-phase mixture and the indicated level will show that the prescurizer is rnly partially full. Except for closing the PORV block valve, operator actions tad system response are the same for these breaks as for si=ilar breaks in the loops.

5.2 Transients k'ith Initial Resnonses Similar :o a Small Break Several :msienta give initial alarms s1=12 - to s=all breaks. These transients 'Ill be distinguished by additional alar =s and indications or sub-i sequent systes resoens65.

Overcoolin; transients such as stea= line breaks, increased reedwater l

flow, and steam generator overrill can cause RCS pressure decreases with low-  !

, pressure reactor titp and ESTAS actuation. But stea= line breaks actuate

( -

low steam pressure alar =s for the affected steam generator, and steam generator I

overfills result in high steam generator level indications. The overcooling

ransients will represstri e the . primary system because of HPI actuation, and will return to a subcooled condition during repressurization. The i=nediate y, acti'ets for both overcoeling and small break transients are the same.

I' A lost-of-feedwater transient will result in a high reactor syste=

pressure alarm but does not give an ESTAS actuation alar =.

l A loss of integrated control system power transient starts with a high RC pressure trip. After the reactor trip, this becomes an overcooling transient

- and will give low reactor system pressure and poasible ESTAS actuation.' Stea generator levels, rd=a: n high and the syste: becomes subcooled during re;Tessuri-

=ation.

N \

'Y c . i 7052 169 s

Design features of the B&W NSS provid; automatic protection during the aarly part of small break transients, thereby providing adequate time for small breaks to be identified and appropriate action taken to protect the system.

5.3 Transients That Might Initiate a LOCA There are no anticipated transients that =ight initiate a LOCA since the PORV has been resce to a higher pressure and will not actuate during anticipated transients such as loss of =ain feedvater, turbine trip, or loss cf offsite power.

However, if the PORV should lift and fail to reseat, there are a nu=ber of indications which differentiate this transient fro = the anticipated tran-sients identified above. These include:

ESFAS actuation Quench tank pressure /temperatura alar =s Saturated primary system Rising cressurizer level

(

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These additional signals will identify to the operator that in addition to the anticipated transient, a LOCA has occurred. In the unlikely event that small breaks other than a =alfunctioning PORV occur after a transient, they ci: be identified by initially decreasing RCS pressure and convergence to saturation conditions in the reactor coolant. S=all break repressurization, if it occurs, will follow saturation conditions. By remaining aware of whether the reactor coolant remains subcooled or becomes saturated af ter transients, the operator is able to recognize when a small break has occurred.

?C)52 170 7

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~~

6.0 SMALL BREAK ANALYSIS Section 6 of Volume 1 of " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant,"

dated May 7,1979, presents computer analysis results of small breaks in the RCS which are intended to augment those in the SAR submittals.

Three types of analyses are provided:

1. Quantification of the maximum delay permissible before auxiliary feedwater must be established.

2. Evaluation of small breaks in the pressurizer which support tne philosophy by which B&W establishes the small break worst-case location.

3. Evaluation of breaks which support the philosophy by which the small break ' pactrum break sizes are chosen (specifically breaks which

~

(

will undergo a repressurization during the transient).

The work performed for Section 6 was done primarily on lowered-loop plant arrangements. However, these analyses also confirm the validity

~

of References 1 through 3 in vold:dTfEr'the raised-loop, Davis-Besse configuration.

Thus, the conclusions drawn from Section 6 apply equally well to the raised-loop arrangement. The sections below discuss the applicability of Section 6 of Volume 1 to Davis Besse.

6.1 Introduction Section 6.1 of Volume 1 of " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assemoly Plant," dated May 7,1979, is applicable to Davis Besse.

6.2 Events Evaluation (a) Quantification of the Maximum Delay Permissible Before Auxiliary 4

Feedwater Mast be Established: Davis Besse-1 has a redundant safety grace auxiliary feecwater system. With tnis system, loss

, 7052 174

of all feedwater is considered to be an extremely unlikely event.

The three analyses described ih Section 6.2.1 of Volume 1 which were performed on the lo ered loop plant model conservatively character.ize the expected performance of the Davis Besse-1 system in the extremely unlikely event that all feedwater is Icst for a period of time. Tha essential modeling assumptions and analysis method d? scribed in Section 6.2.1.2 of Volume 1 are appropriate for Davis Besse with the exception of the auxiliary feedwater system interruption which is as discussed above.

Two cases are given for breaks which are large enough to continue depressurization and to actuate the HPI system after all secondary cooling is teminated. The essential system behavior predicted by the two analyses given for this category of breaks in paragraph 6.2.1.3.3 and paragraph 6.2.1.3.4 of Volume 1, is typical of Davis Besse as well as for the lowered loop plants. In these two analyses, system pressure promptly falls below the high pressure injection system actuation setpoint and remains below the shutoff head of the high pressure injection pumps at Davis Besse. Thus, the high pressure injection system at Davis Besse would be effective throughout these transients and would deliver more water per unit time that is considered in the detailed analysis given in the referenced sections. Hence, for these two breaks and tne class of breaks they represent, safe system performance is expected for the Davis Besse plant without core damage and we can conclude that the criteria of 10 CFP. 50.46 are satisfied 2052 @c z

for the Davis-Besse piant without the use of auxiliary feedwater for all breaks larger than 0.02 ft , assu=ing 2 ECCS trains are available.

The analysis of a 0.01 ft2 break given in paragraph 6.2.1.3.5 of Volume 1 also characterizes the expected system behavior for the Davis-Besse plant for a break of that size. In this analysis, the prompt reactor coolant system depressurization is not sufficient to actuate the high pressure injection system. Auxiliary feedwater is assumed not to be available during the first 20 minutes. Thus, during this period, reactor core decay heat is being removed by boil-off of pri=ary coolant inventory through the pressurizer relief system. Since the high pressure injection system is not functioning, its head-flow characteristics are unimportant to the predicted behavior. The raised loop configuration of the Davis-Besse plant results in a more favorable condition with respect to available rearcor coolant system inventory to cover the core at the end of 20 minutes than for the specific case analyzed in paragraph 6.2.1.3.5 of Volume 1.

This is due to the fact that with the raised loop configuration, almost all the inventory outside the reactor vessel is held at a level higher than the reactor vessel nozzles. Thus, given an equivalent boil-off of reactor coolant system volume, the raised loop Davis-Besse configuration will have more available water remaining in the reactor coolant system to cover the core than the lowered loop configuration for which the detailed analysis was performed. This is discussed more completely in Section 6.2.1.3.6 of Volume 1.

Initiation of auxiliary feedwater to the steam generators within 20 minutes results in primary system depressurization 2052 176 3

to the high pressure injection system initiation point.

Primary system pressure continues to fall so that the Davis-Besse high pressure injection system pumps will be fully effective.

Based on these analyses, we conclude that delay of both redundant trains of safety grade auxiliary feedwater flow of up to 20 minutes in the Davis-Besse plant will not result in uncovering of the core and the core cladding temperature will remain within a few degrees of saturated fluid temperature. These analyses support the conclusion that compliance with 10CFR50.46 is assumed for the Davis-Besse plant.

Section 6.2.2 of Volume 1 is not applicable to Davis-Besse.

(b) Small Breaks in the Pressurizer: It has been shown (see References 1, 2, and 3 of Section 6, Volume 1) that breaks located in the pressurizer and the hot legs are less severe than breaks located in the cold leg.

To reverify that fact, two cases involving leaks in the pressurizer were analyzed. These cases were provided in Section 6.2.3 of Volume 1. The basic system behavior for either raised loop or lowered loop plant is the same and can be characterized as follows:

(1) A rapid system depressurization due to steam release out the relief valve . This will result in reactor scram and ESFAS actuation.

(2) The indicated pressurizer level will initially increase due to the break. After reactor scram, the pressurizer level decreases due to system contraction. Following saturation of the hot legs, there will be an insurge into the pressurizer resulting in the pressuri-zer going solid.

?052 9

(3) After the pressurizer is filled by a two-chase mixture, a low quality mixture will be discharged through the valve. This will result in a large increase in the leak flow rate.

(4) Primary side pressure will ultimately be maintained at approximately 1200 psia.

(5) Long-term cooling will be established by one HPI train.

The Davis Besse pilot operated pressurizer relief valve is 80% larger in relief capacity than the pilot operated relief valve modeled in the specific cases given in Section 6.3.2.1 of Volum 1. However, as discussed more completely in paragraph 6.2.3.2.3 of Volume 1, the general phenomena observed will be the same as those observed in the breaks specificalb snalyzed and the consequences of the resulting break would be acceptable. In fact, as discussed in paragraph 6.2.3.2.3, breaks of this size and larger are bounded by the existing LOCA analyses. Thus, these analyses show that for the Davis Besse plant, small breaks in the too of the pressurizer equivalent to and larger than the relief capacity of the pilot operated relief valve do not lead to core uncovering or core damage and meet the acceptance criteria of 10 CFR 50.46.

(c) Evaluation of Small Breaks which will Underco a Recressurization Durina the Course of the Transient: Since the course of these breaks may be expected to be affected by the head and flow characteristics of the plant's high pressure injection system, plant-specific analyses have been performed for Davis Besse.

This analysis is described in Section 6.2.5 of Volume 1. 2052 178 The results indicate that the core will remain covered throughout S'

the transient and that the criteria of 10CFR50.46 are satisfied, thus establish' ag the capability of the Davis-Besse plant to safely accommo-date reactor coolant system breaks in this range.

An additional case is reported in paragraph 6.2.4.3.2 of Volume 1 showing the effect of asy==etric feeding to the steam generators. This case demonstrates that one generator is sufficient to guarantee core cooling in the lowered loop design. The difference in loop arrangement and HPI pump characteristics in the Davis-Besse design do not affect the characteristics of the results of this case, and this case demonstrates the conclusion that one loop guarantees sufficient ' cooling for the Davis-Besse design as well.

6.3 Conclusions The analyses which have been performed and are documented in Section 6 of Volume 1 demonstrate that the Davis-Besse ECCS systems will control small breaks and satisfy the criteria of 10CFR50.46. Specific conclusions applicable to the Davis-Besse plant include the following:

1. In the highly unlikely event that both redundant trains of the safety grade auxiliary feedwater system are delayed, restoration of auxiliary feedwater at 20 minutes (and possibly later) is suf ficient to assure that core damage does not result from small breaks (assuming two EPI trains operational).

2. Analyses of relief valve failures at the top of the pressurizer show that a single ECCS train is sufficient to assure that the core remains covered during these transients.

7052 179 6

3. The consequences of breaks in the hot legs or in the pressurizer has been demonstrated to be bounded by small break analyses performed for breaks in the cold leg pump discharge piping.

4. For very small breaks in the Davis-Besse plant which require the

-team generator as a heat removal system, it has been shown that system repressurization may occur. However, the establishment of steam condensation by the steam generator as a heat removal mechanism controls the repressurization and assures effective action by the high pressure injection system to maintain the core covered with water and prevent core damage.

5. It was demonstrated that asymmetric auxiliary feedwater injection (a single steam generator) provides adequate heat re= oval to assure that core uncovery does not occur.

7

Appendix 1 Natural Circulation In BG' Operating Plants The infor=ation in Appendix 1 - Rev. 1 of Volume II of " Evaluation of Transient Behavior and Small Reactor Coolant Syste= Breaks In the 177 Fuel Assembly Plant" dated 5/7/79 is applicable to this section.

?052 181 O

=

Appendix 2 - Steam Generator Tube '.her=al Stress Evaluation The infor=ation in tppendix 2 Volume II of " Evaluation of Transient Behavior and S=all Reactor Coolant System Breaks in the 177 Tuel Asse=bly Plant" dated 5/7/79 and Supplement 1 submitted by letter fro: J. E. Taylor to Dr. R. J. Pattson dated 5/10/79 is applicable to this section.

7052 182 e

Appendix 3 - Sensitivity Study on the Effect of Starting a Reactor Coolant Pu=p in a Highly Voided System The infor=ation in Appendix 3 Volume II of " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" dated 5/7/79 and a letter from J. E. Taylor to T. M. Novak dated 5/10/79 is applicable to Davis Besse.

9052 183

APPENDIX 4 OPERATING GUIDELINES FOR

^

SMALL BREAKS The attached information has been prepared to serve as a basis for developing detailed emergency operating procedures for situations involving small breaks in the reactor coolant system. It defines the symptoms, immediate actions, precautions and followup actions. Fo'r various combinations of equipment availability, the required actions are outlined to take the plant to the long-term cooling mode.

The procedural guidelines form Part II of a two-part document which has been provided to the operating utilities. Part I is included as Section 5.0 of this report and describes small breaks in phenomenalogical terms.

. ?052 184

/

PART II - OPERATING GUIDELINES FOR SMALL BREAKS 1.0 SYMPTOMS AND INDICATIONS (IMMEDIATE INDICATIONS [

1.1 EXCESSIVE REACTOR COOLANT SYSTEM (RCS) MAKEUP

• 1.2 DECREASING RCS PRESSURE 1.3 REACTOR TRIP 1.4 DECREASING PRESSURIZER LEVEL

• 1.5 ESFAS ACTUATION

• 1.6 LOW MAKEUP TANK LEVEL *

• May not occur on all small breaks.

2.0 IMMEDIATE ACTIONS 2.1 VERIFY CONTROL ROOM INDICATIONS SUPPORT THE ALARMS RECEIVED, VERIFY AUTOMATIC ACTIONS, AND CARRY OUT STANDARD POST-TRIP ACTIONS.

2.2 VERIFY THAT INJECTION FLOW FROM EACH HPI-PUitP IS BALANCED WHEN HPI IS INITIATED.

2.3 VERIFY THAT APPROPRIATE ONCE-THROUGH STEAM GENERATOR (OTSG)

LEVEL IS MAINTAINED BY FEEDWATER CONTROL (BY ICS CONTROL OF MAIN FEEDWATER OR SFRCS CONTROL OF AUXILIARY FEEDWATER).

2.4 MONITOR SYSTEM PRESSURE AND TEMPERATURE. IF SATURATED CONDITIONS OCCUR, INITIATE HPI.

3.0 PRECAUTIONS 3.1 IF THE HPI SYSTEM HAS ACTUATED BECAUSE OF LOW PRESSURE CONDITION, IT MUST REMAIN IN OPERATION UNTIL ONE OF THE FOLLOWING CRITERIA IS SATISFIED:

?052 185 z

1. THE LPI SYSTEM IS IN OPERATION AND FLOWING AT A RATE IN EXCESS OF 1000 GPM IN EACH LINE AND THE SITUATION HAS BEEN STABLE FOR 20 MINUTES.

OR

2. ALL HOT AND COLD LEG TEMPERATURES ARE AT LEAST 50 BELOWTHh SATURATION TEMPERATURE FOR THE EXISTING RCS PRESSURE, THE HOT LEG TEMPERATURES ARE NOT MORE THAN 500 HOTTER THAN THE SECONDARY SIDE SATURATION TEMPERATURE, AND THE ACTION IS NECESSARY TO PREVENT THE INDICATED PRESSURIZER LEVEL FROM GOING OFF-SCALE 0

HIGH. IF 50 SUBC00 LING CANNOT BE MAINTAINED, THE HPI SHALL 0

BE REACTIVATED. THE DEGREE OF SUBC00 LING BEYOND 50 AND THE LENGTH OF TIME HPI IS IN OPERATION SHALL BE LIMITED BY THE PRESSURE / TEMPERATURE CONSIDERATIONS FOR THE VESSEL INTEGRITY.

3.2 PRESSURIZER LEVEL MAY BE INCREASING DUE TO RCS REACHING SATURATED CONDITIONS OR A BREAK ON TOP OF THE PRESSURIZER.

3.3 IF HIGH ACTIVITY IS DETECTED IN A STEAM GENERATOR, ISOLATE THE LEAKING GENERATOR. BOTH STEAM GENERATORS MAY NOT BE ISOLATED SINCE OTSG COOLING IS REQUIRED FOR DECAY HEAT REMOVAL.

3.4 OTHER INDICATIONS WHICH CAN CONFIRM THE EXISTENCE OF A LOCA:

3.4.1 RC DRAIN TANK (QUENCH TANK) PRESSURE (MAY RUPTURC DISK).

3.4.2 INCREASING REACTOR BUILDING SUMP LEVEL.

3.4.3 INCREASING REACTOR BUI.LDING TEMPERATURE.

3.4.4 INCREASING REACTOR BUILDING PRESSURE.

3.4.5 INCREASING RADIATION MONITOR READINGS INSIDE CONTAINMENT.

3.4.6 REACTOR COOLANT SYSTEM TEMPERATURE BECOMING SATURATED RELATIVE TO THE RCS PRESSURE.

3.4.7 HOT LEG TEMPERATURE EQUALS OR EXCEEDS PRESStJRIZER TEMPERATURE.

_ . - - 2052 186 3

3.5 HPI COOLING REQUIREMENTS COULD DEPLETE TE BORATED WATER STORAGE

. '- TANK, AND INITIATION OF LPI FLOW FROM THE REACTOR BUILDING SUfG

- TO THE HPI PUiPS WOULD BE REQUIRED.

3.6 ALTERNATE INSTRUENT CHANNELS SHOULD BE CHECKED AS AVAILABLE ,

TO CONFIRM KEY PARAMETER READINGS (1.E., STSTEM TEMPERATURES, PRESSURES AND PRESSURIZER LEVEL).

3.7 MAINTAI't A TEMPERATURE VERSUS TIME PLOT AND A CORRESPONDING TEMPERATURE - PRESSURE PLOT ON A SATURATIO!! DIAGRAM. THESE PLOTS WILL MAKE IT ROSSIBLE TO TRACK THE PuiiT'S t CONDITION

. , THROUGH PLANT C00LDOWN. PRIMARY TEMPERATURE AND PRESSUPI WILL DECREASE ALONG THE SATURATION CURVE UNTIL SUSC00 LED CONDITIONS

~

ARE ESTABLISHED. THIS WILL ,BE INDICATED BT PRIMARY SYSTEM PRESSURE NO LONGER FOLLOWING THE SATURATION CURVE, AS PRIMARY SYSTEM TEMPERATURE DECREASES. WHEN THIS OCCURS, PRIMARY SYSTEM PRESSURE Si:0ULD BE CONTROLLED BY ADJUSTING HPI FLOW, TO MAINTAIN

• 50 F0 SUBC00 LING.

a . .

4.0 FOLLOWUP ACTIONS 4.1 IDENTIFICATION AND EARLY CONTROL' .

4.1.1 IF HPI HAS INITIATED BECAUSE OF LOW PRESSURE, CONTROL HPI

.IN ACCORDANCE WITH STEP 3.1. ,

. 4.1.2 IF BOTH HPI TRAINS HAVE NOT ACTUATED ON ESFAS SIGNAL, START SECOND HPI TRAIN IF POSSIBLE. _ _. __ ._. .

4.1.3 IF RC PRESSURE DECREASES CONTINUOUSLY, VERIFY THAT CORE FLOOD TANKS (CFT'S) AND LOW PRESSURE INJECTION (LPI)

HAVE ACTUATED AS NEEDED, AND BALNICE LPI'. ,

D #

3 2052 l87 n.

y

4.1.4 ATTEMPT TO LOCATE AND ISOLATE LEAK IF POSSIBLE. LETDOWN WAS ISOLATED IN STEP 2.1. OTHER ISOLATABLE LEAKS ARE

, PORY (CLOSE BLOCK VALVE) AND BETWEEN VALVES IN SPRAY LINE (CLOSE SPRAY AND BLOCK VALVE).

4.1.5 DETERMINE AVAILABILITY OF REACTOR COOLANT PUMPS (RCP'S)

AND MAIN AND. AUXILIARY FEEDWATER SYSTEMS.

4.2 ACTIONS WITH FEEDWATER AVAILABLE TO ONE OR BOTH GENERATORS 4.2.1 MAINTAIN ONE RCP RUNNING PER LOOP (STOP OTHER RCP'S).

IF NO RCP'S AVAILABLE, GO TO STEP 4.2.4 BELOW.

4.2.2 ALLOW RCS PRESSURE TO STABILIZE 4.2.3 ESTABLISH AND MAINTAIN OTSG COOLING BY ADJUSTING STEAM PRESSURE VIA TURBINE BYPASS AND/OR ATMOSPHERIC DUMPS.

C00LDOWN AT 1000F PER HOUR TO ACHIEVE AN RC PRESSURE OF 250 PSIG. REFER TO PRECAUTION 3.7 FOR DEVELOPMENT OF TEMPERATURE AND PRESSURE PLOTS. ISOLATE CORE FLOOD TANKS WHEN 500F SUSC00 LING IS ATTAINED AND RC PRESSURE IS LESS THAN 700 PSIG. GO INTO LPI COOLING PER APPENDIX A.

4.2.4 IF RCP'S ARE NOT OPERATING:

4.2.4.1 ESTABLISHANDCONTROLOTSGLkVELTO96 INCHES INDICATED ON THE STARTUP RANGE INSTRUMENTATION.

4.2.4.2 IF RC PRESSURE IS DECREkSING, WAIT UNTIL IT STABILIZES OR BEGINS INCREASING. IF IT BEGINS INCREASING, GO TO STEP 4.2.4.4.

4.2.4.3 PROCEED WITH A CONTROLLED C00LDOWN AT 1000F/HR.

BY CONTROLLING STEAM GENERATOR SECONDARY SIDE PRESSURES. MONITOR RC PRESSURES AND TEMPERATURES DURING C00LDOWN AND PROCEED AS INDICATED BELOW.

7 7052 188

4.2.4.3.1 IF RC PRESSURE CONTINUES TO DECREASE, FOLLOWING SECONDARY SYSTEM PRESSURE DECREASES AND WITH PRIMARY SYSTEM TEMPERATURES INDICATING SATURATED CONDITIONS, CONTINUE C00LDOWN UNTIL AN RC PRESSURE OF 150 PSI IS REACHED, AND PROCEED TO STEP A.4 0F APPENDIX A.

4.2.4.3.2 IF RC PRESSURE STOPS DECREASING IN RESPONSE TO SECONDARY SIDE PRESSURE

. DECREASE AND REACTOR SYSTEM BECOMES SUBC00 LED, CHECT. TO SEE THAT THE FOLLOWING CONDITI0fiS ARE BOTH SATISFIED: -

~

A) ALL HOT. AND COLD LEG TEMPERATUES ARE BELOW THE SATURATION TEM? ERA-

. TUPI FOR TE EXISTING RCS PRESSURE.

.AND B) RCS HOT LEG TEMPERATURES ARE NOT

- 0

- MORE THAN S0 F H01stR THAN THE STEAM GENERATOR SECONDARY SIDE SATURATION TEMPERATURE.

IF THESE CONDITIONS ARE SATISFIED AND

' REMAIN SATISFIED, CONTINUE C00LDOWN TO ACHIEVE AN RCS TEMPERATURE (COLD LES) 0F

} 0 280 F, AND PROCEED TO STEP A.1 0F 4

APPENDIX A.

NOTE: IF THE CONDITIONS ABOVE ARE MET AT BELOW 700 PSIG, THE CORE FLOOD TANKS 7

SHOULD SE ISOLATED.

7' 7052 189

NOTE: IF THE PRIMARY SYSTEM IS 500F SUSC00 LED IN BOTH HOT AND COLD LEGS AND PRIMARY SYSTEM PRESSUP,E IS ABOVE 250 PSIG, STARTING A REACTOR COOLANT PUMP IS PERMISSIBLE. IF SYSTEM DCES NOT RETURN TO AT LEAST 500F SUB-COOLING IN TWO MINUTES, TRIP PUMPS. IF FORCED CIRCULATION IS ACHIEVED, PROCEED TO STEP 4.2.

4.2.4.3.3 IF RC PRESSURE STOPS DECREASING AND THE CONDITIONS OF 4.2.4.3.2 ARE NOT MET OR CEASE TO BE MET OR IF RC PRESSURE BEGINS

. TO INCREASE, THEN PROCEED TO STEP 4.2.4.4 BELOW.

_ 4.2.4.4 RESTORE RCP FLOW (ONE PER LOOP) WHEN POSSIBLE PER THE INSTRUCTIONS BELOW. IF RC PUMPS CANNOT BE OPERATED AND PRESSURE IS INCREASING, GO TO STEP 4.2.4.6.

4.2.4.4.1 IF PRESSURE IS INCREASING, STARTING A PUMP IS PERMISSIBLE AT RC PRESSURE GREATER THAN 1600 PSIG.

4.2.4.4.2 IF REACTOR COOLANT SYSTEM PRESSUPI EXCEEDS STEAM GENERATOR SECONDARY PRESSURE BY 600 PSIG OR MORE " BUMP" ONE REACTOR COOLANT PUMP' FOR A PERIOD OF APPROXIMATELY 10 SECONDS (PREFERABLY IN OPERABLE STEAM GENERATOR LOOP.) ALLOW REACTOR COOLANT SYSTEMPRESSURETOSfASILIZE. CONTINUE C00LDOWN.

IF REACTOR COOLANT SYSTEM PRESSUPI AGAIN EXCEEDS SECONDARY PRESSURE SY 500 7 0052 190

PSI, WAIT AT LEAST IS MINUTES AND REPEAT THE PUMP " BUMP". BGMP ALTERNATE PUMPS S0

{ THAT NO PUM? IS BUM?ED MOE THAN ONCE IN 1

0&lCD

@ U U\\

'Oh1 uu A]'fdd AN HOUR. - THIS MY BE REPEATED, WITH AN INTERVAL OF IS PENUTES, UP TO 5 TIMES.

AFTER THE FIFTH *BCMP", ALLOW THE REACTOR COOLANT PUMP TO CD?.iINUE IN OPERATION. [

4.2.4.4.3 IF PRESSURE HAS STABILIZED FOR GEATER THAN ONE HOUR, SEC050ARY PRESSUE IS LESS THAN 100 PSIG AND PRIMARY PRESSURE IS GREATER THAN 250 PSIG, BUMP A PUMP, WAIT 30 MINUTES, JND START AN ALTERNATE

~

PUMP. .

e 4.2.4.5 IF FORCED FLOW IS ESTABLISH 3, GO TO STEP 4.2.3.

4.2.4.6 IF A REACTOR COOLANT PUMP CNNOT 5E OPERATED AND REACTOR C00LANT SYSTEM PPNRE REACHES 2300 PSIG, OPEN PRESSURIZER PORY TO PSUCE EACTOR COOLANT SYSTEM PRESSURE. RECLOSE PCRV WHEN RCS PESSUE

~

FALLS TO 100 P,SI ABOVE THE SECONDARY PRESSURE.

REPEAT IF NECESSARY. IF PDEV IS NOT OPERABLE, PRESSURIZER SAFETY VALVES 1D.LL RELIEVE OVERPRESSURE.

4.2.4.7 MAINTAIN RC PESSURE AS INDICATED IN 4.2.4.6 IF PRESSURE INCREASES. MAINTAIN THIS COOLING MODE UNTIL AN RC PUMP IS STARTED OR STEAM GENERATOR COOLING IS ESTABLISHED AS IXDICATED BY ESTABLISHING CONDITIONS DESCRIBED IN 4.2.4.3.1 OR 4.2.4.3.2.

WHEN THIS OCCURS, PROCEED AS DIRECTED IN THOSE STEPS.

Y GO TO STEP 4.2.2 IF FORCED FLOW IS ESTABLISHED.

8 2052 191

. APPENDIX A

. LPI COOLING A.1 0 DETERMINE IF PRIMARY COOLANT IS AT LEAST S0 F SUSC00 LED. IF NOT GO TO STEP A.3.

A.1.1 START LPI PUM?S. IF BOTH PU'4PS ARE OPEPABLE GO TO STEP A.2.

FOR ONE LPI PUMP OPERABLE PAINTAIN OTSG COOLING AS FOLLOWS.

TE OPERABLE LPI PUMP WILL BE USED TO PAINTAIN SYSTEM INVENTORY.

A.1.2 0 OBTAIN PRIMARY SYS' TEM CONDITIONS OF 280 F AND 250 PSIG.

A.I.3 ALIGN THE DISCHARGE OF THE OPERABLE LPI PUMP TO TE SUCTIONS

, OF THE HPI PUMPS AND TAKE SUCTION FROM THE BWST. IF THE BWST IS AT THE LOW LEVEL ALARM, ALIGN LPI SUCTION FROM THE RB SUMP AND. SHUT SUCTION FROM BWST. - -

A.1.4 STraT THE OPEPABLE LPI PUMP SPECIFIED ABOVE. THE HPI-LPI SYSTEMS WILL NOW BE IN " PIGGY BACK" AND HPI FLOW IS MAINTAINING SYSTEM PPISSURE.

A.I.S GO TO SINGLE RC PUMP OPERATION. -

A.1.6 WHEN THE SECOND LPI PUMP IS AVAILABLE AO GN IT IN THE DECAY HEAT MODE AND COMMENCE DECAY HEAT REMOVAL. (DECAY HEAT SYSTEM FLOW GREATER THAN 1000 GPM). SECURE PIPAINING RC PUMP WHEN DECAY HEAT FIMOVAL IS ESTABLISHED. ,

CAUTION: VERIFY THAT ADEQUATE NPSH EXISTS FOR THE DECAY HEAT PUMP IN THE DH REMOVAL MODE. IF INADEQUATE, TRANSFER TO LPI MODE.

A.1.7 REDUCE REACTOR COOLANT PRESSURE TO 150 PSIG SY THROTTLING HPI - .

FLOW. CONTROL RC TEMPERATURE USING THE DECAY HEAT SYSTEM COOLER BYPASS TO MAINTAIN SYSTEM PRESSURE AT LEAST 50 PSI ABOVE SATURATION PRESSURE, TO ASSUPI THAT NPSH REQUIREMENTS FOR THE DECAY HEAT PUMP ARE PAINTAINED. .

, ?052 192 ,

~

A.1.8 SECURE THE HPI PUM? AND SHIFT THE LPI PUMD SUPPLYING IT TO THE LPI INJECTION MODE.

0 A.I.9 REDUCE REACTOR COOLANT TEMPERATURE TO 100 F BY CONTROLLING THE DECAYHEATSYSJEMCOOLERBYPASS.

NOTE: IF ONE OF THE LPI/ DECAY HEAT PUMPS IS LOST, ETUPJi TO OTSG

~

COOLING USING NATURAL CIRCULATION OR ONE REACTOR COOLANT PUMP (A1). ..

A.2 COOLDOWN ON TWO LDI PUMDS .

A.2.1 MAINTAIN RCS PESSURE AT 250 PSIG AND REDUCE RCS TEMPERATURE TO 2800 F. ,

A.2.2 ALIGN ONE LPI PUMP IN THE DECAY HEAT REMOVAL MODE.

A.2.3 SECURE ONE RC PUMP. A SINGLE RC PUMP IS NOW OPERATING.

A.2.4 START THE DECAY HEAT PUMP IN THE DECAY HEAT EMOVAL MODE, AND WHEN DECAY HEAT SYSTEM FLOW IS GREATER THAN 1000 GPM, SECURE

= .

THE RUNNING RC PUMP.

~~

A.2.5 REDUCE RC PESSURE TO 150 PSIG BY THROTTLING HPI FLOW. CONTROL

- %RC TEMPERATURE TO MAINTAIN AT LEAST 50 PSI MARGIN TO SATURATION

• !.*. g

~ '

PRESSURE.

~

N.2.6 START THE SECOND LPI PUMP IN THE LPI INJECTION MODE. SECURE HPI PUMP.

A.2.7 SHIFT LPI SUCTION FROM THE BWST TO THE REACTOR BUILDING SUMP WHEN SUFFICIENT NPSH IS AVAILABLE. ,

NOTE: THIS IS DESIRABLE TO AVOID UNNECESSARY QUANTITIES OF WATER IN CONTAINMENT. .

0 A.2.8 REDUCE REACTOR COOLANT TEMPERATURE TO 100 F BY CONTROLLING THE

. DECAY HEAT SYSTEM COOLER BYPASS.

NOTE: IF ONE OF THE LPI/ DECAY HEAT PUMPS IS LOST, RETURN TO OTSG COOLING USING NATURAL CIRCULATION OR ONE RC PUMP.

'052 193 .

/0

.3 .. .-_e--

, t.

A.3 COOL DOWN RC SYSTEM AT SATURATION A.3.1 MAINTAIN RC PRESSURE AT 250 PSIG.

A.3.2 ALIGN ONE LPI PUMP TO SUCTION OF THE HPI PUMPS AND THE SUCTION TO THE REACTOR BUILDING SUMP. (SHUT BWST SUCTION VALVE FOR THIS PUMP.)

A.3.3 WHEN THE BWST LEVEL REACHES THE LO-LO LEVEL LIMITS, START THE LPI PUMP AND SHUT THE HPI PUMP SUCTION FROM THE BWST.

A.3.4 WHEN PRIMARY SYSTEM TEMPERATURE BECOMES SUBC00 LED BY AT LEAST 50 F, GO TO A.1.1.

A.4 C00LDOWN WITHOUT REACTOR COOLANT PUMPS A.4.1 RCS INITIAL CONDITIONS ARE: PRESSURE 150 PSI, TEMPERATURES AT SATURATION.

A.4.2 ALIGN LOW PRESSURE INJECTION SYSTEM FOR SUCTION FROM REACTOR BUILDIN3 SUMP AND PLACE INTO SERVICE.

A.4.3 CONTROL RC TEMPERATURE WITH DECAY HEAT COOLERS. IF ONLY ONE LPI PUMP IS AVAILABLE, CROSS CONNECT DISCHARGES AND BALANCE FLOWS.

A.4.4 ISOLATE CORE FLOOD TANKS.

A.4.5 GO TO STEP A.1.1 AND FOLLOW THE PROCEDURE GIVEN THERE, IGNORING THE INSTRUCTIONS RELATING TO RC PUMP OPERATION.

2052 194

-me h = *

//

Appendix 5 - B&W Assessment of " Decay Heat Removal During A Very S=all Break LOCA for B&W 205 Fuel Assembly PWR", January,1978, C. Michelson.

The information in Appendix 5 of Volume II of " Evaluation of Transient Behavior and S=all Reactor Coolant System Breaks in the 177 Fuel Assembly Plant", dated 5/7/79 is applicable to Davis Besse. The only difference between the 177 FA plant described in ite: 1 on page A 5-3 and Figure A-5-1 and Davis Besse is that, Davis Besse has the dual advantages of the raised steam generators (similar to the 205 FA plants) and the injection location for auxiliary feedwater ncar the top of the steam generator s1=11ar to the other 177 FA plants.

2052 195

.