ML18044A456

Responds to NRC 791221 Request for Info Re Applicability of Main Steam Line Break Analysis.Current Means of Manual Actuation of Auxiliary Feedwater Sys & Proposed Control Grade Sys Acceptable

ML18044A456
Person / Time
Site:	Palisades
Issue date:	01/21/1980
From:	Huston R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Ziemann D Office of Nuclear Reactor Regulation
Shared Package
ML18044A457	List: ML18044A456 ML19257C285
References
TASK-03-12, TASK-06-02.D, TASK-06-03, TASK-15-02, TASK-15-2, TASK-3-12, TASK-6-2.D, TASK-6-3, TASK-RR NUDOCS 8001250503
Download: ML18044A456 (58)

Text

• ~.

/

consumers Power company General ,Offices: 212 Wast Michigan Avenue, Jackson, Michigan 49201 * (517) 788-0650 January 21, 1980 Director, Nuclear Reactor Regulation Att Mr Dennis-L Ziemann, Chief Operating Reactors Branch No 2

.US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 -*LICENSE DPR PALISADES PLANT -

AUTOMATIC INITIATION OF AUXILIARY FEEDWATER SYSTEM AT PALISADES PLANT Consumers Power Company was requested by NRC letter dated December 21, 1979, to respond to a concern regarding the applicability of the cur:rent main steam line break analysis. Consumers Power was requested to submit an analysis within twenty days of receipt of the NRC letter assessing the effect of*

automatic initiation of auxiliary feedwater (AFWS) on the current analysis.

Du~ to the work involved in preparing such_an evaluation, an* extension of the response period until January 21, 1980 was requested.by several licensees.

Consum~rs Power Company was informed in a telephone conversation between Richard Silver (NRC) and Steve Frost (Consumers Power Company) on January 9, 1980, that the requested eitension had be~n granted and that our response should be sub~itted by January 21, 1980.

The attachments to this letter provide the information requested. Based on this information,* Consumers Power Company considers that the cu,rrent means of manual actuation of the auxiliary f eedwater system as well as the proposed control grade syst~m provides a safe means of oper.ation at tht;! Palisades Plant.

The December 21, 1979 NRC letter also requested that Technical Specifications changes be submitted incorporating the automatic actuation of AFWs.* The

• letter also identifies that procedure changes and training wili be required as f'

I

• 2 a result of this modification. Consumers Power Company will accomplish necessary procedure modification and training. Technical Specifications changes will be submitted, as necessary, at a later date.

Roger W Senior Licensing Engineer CC JGKeppler, USNRC Att (1) Response to request for information of the AFWS effect on Main Stream Line Break Accident Analysis.

(2) Best estimate MSLB analysis to assess NSSS and containment response with automatic auxiliary feedwater actuation.

.r RESPONSE TO REQUEST FOR INFORMATION Automatic Initiation of the AFWS Effect on Main Steam Line Break Accident Analysis A. RETURN TO POWER Item 1 Provide the results of analyses of main steam line breaks that are the most limiting with respect to fuel failure resulting from return to power.

Analyses should be presented covering:

a. Break inside containment.

b. Break outside containment.

c. Availability or loss of off-site power.

Justify omitting an analysis for any of the above.

Response to Item 1 The main steam line break accident inside containment was analyzed assuming off-site power available. Exxon Report No XN-NF-77-18, "Plant Transient Analyses of the Palisades Reactor for Operation at 2S30 Mwt," contains the results of this analysis. For the full power steam line break case, this analyses assumed a feedwater ramp down from 100% to 5% over the 60 seconds following reactor trip. This assumption is still conservative, when the automatic initiation of auxiliary feedwater is considered, for the following reasons:

1. The plant is presently being modified as a result of a discovered deficiency (see Licensee Event Report No 79-041) to close the main feedwater regulating and bypass valves on a low steam generator pressure signal (~ 500 psia). This will result in complete termination of main feedwater much sooner than assumed in the reference analysis.

mi0180-0377a-43

2

2. The design of the automatic system for auxiliary feedwater initiation is such that the flow to either steam generator will be automatically limited to less than a predetermined amount using an automatic flow controller.

The flow limit is not yet finalized, but is expected to be less than the maximum delivery rate from one auxiliary feedwater pump when steam generator pressure is 900 psia. This flow rate is estimated at ~ 250 gpm/steam generator which is much less than the 5% (of ~ 12,000 gpm per steam generator) assumed in the analysis. For the zero power steam line break case, a constant feedwater flow rate of 415 gpm/steam generator was assumed, which is clearly conservative considering the intended design of the system for automatic auxiliary feedwater initiation.

Main steam line breaks from full power, both inside and outside containment, and without off-site power *available were analyzed in amendments to the FSAR (Amendment 15, Section 14.3 and Amendment 17, Section 5.0). The outside break was found to be clearly less limiting than the inside break because the blowdown rate is significantly reduced for the outside break as a result of the flow venturies in the main steam lines. The inside break without off-site power available was found to result in a small amount of fuel failures (less than 1%) compared to the case with off-site power available in which zero fuel failures were predicted. In both cases the feedwater flow to the steam generators was modeled as discussed previously. Thus, the existing analyses conservatively bound the case of automatic initiation of auxiliary feedwater.

mi0180-0377a-43

3 Item 2 Provide the time sequence of all actions and eyents occurring during each of the postulated steam line break transients. These events and actions should include:

a. Reactor scram.

b. Turbine trip.

c. Steam line isolation.

d. Feedwater isolation.

e. ECCS actuation.

f. Auxiliary feedwater actuation* and control.

g. Safety/relief valve actuation (primary and secondary systems).

h. Operator actions (define credit for operator action).

i. Initiation of on-site power (if required).

Response to Item 2 The sequence of events for the full and zero power main steam line break events with off-site power available are as shown below, based on XN-NF-77-18.

MAIN STEAM LINE BREAK Sequence of Events Event Full Power Break Zero Power Break Break O Seconds 0 Seconds Reactor Scram 1.6 1.6

• Turbine Trip 1.6 Steam Line Isolation 1.6 1.6 Feedwater Isolation None None ECCS Actuation "' 18 "' 17 Auxiliary Feedwater Actuation None None Safety/Relief Valve Actuation None None Operator Actions None None Initiation of On-Site Power Item 3 For each of the above, identify the initiating signal, the protection system that initiates the action, and the extent of the action ending with the time the element (ie, MSIV ,. turbine stop, turbine control, turbine bypass, etc) reaches its new condition. The above events are to reflect the expected response of the plant and systems.

mi0180-0377a-43

4 Response to Item 3 Reactor trip, turbine trip and main steam line isolation were assumed to occur on low steam generator pressure (500 psia - 22 psi uncertainty) following a .

0.6-second instrumentation delay. Full control rod insertion in 3 seconds was assumed. Turbine stop valve closure in 0.5 seconds, and main steam line isolation valve closure in 6 seconds were assumed. ECCS actuation was assumed on low-pressurizer pressure (1600 psia - 22 psi uncertainty) following a 0.6-second instrumentation delay.

An ECCS piping purge volume of~ 10.3 ft 3 (between the last check valve and the loop) was considered. An ECCS valve opening time of 10 seconds was assumed.

Item 4 Identify and justify any equipment that does not meet Regulatory Guides and IEEE-279 requirements.

Response to Item 4 The approved main steam line break analyses in the FSAR and in ENC Report XN-NF-77-18 take credit for the main feedwater pump ramp down feature of the main feedwater control system which does not meet IEEE-0279 requirements. The Systematic Evaluation Program design bases events review will reconsider the main steam line break event and those systems utilized to mitigate its consequences.

Item 5 Provide a list of potential single failures that could af.fect each of the above actions and show how the analyses presented consider the worst single failures .from a fuel failure standpoint. Note that normal control systems should not be considered to function if their action would be beneficial with respect to fuel failures.

mi0180-0377a-43

5 Response to Item 5 The MSLB results presented in the FSAR and subsequent reload licensing submittals,assumed the following consequential failures in addition to the single failure which initiates the event (ie, the double ended pipe break ins*ide containment):

a. On reactor scram, the highest worth Control Rod was assumed to stick in the fully withdrawn position.

b. On Safety Injection Actuation, one of the HPSI and one of the charging pumps were assumed to fail to start.

c. No main feedwater isolation was assumed on MSIS. The main feed flow was assumed to coast down to 5% of full power flow in 60 seconds.

(More realistically, flow would go to zero in less than 20 seconds.)

Single failures were considered in the design basis to the extent that a failure initiates the event and safety grade equipment is designed to accommodate single failures as described above. No consequential failures other than the three mentioned above were considered. No control systems, other than the main feedwater control system, were assumed to function in the

• analysis.

Single failures concurrent with the MSLB (other than those identified above) are not, and have not been, part of the design basis as described in the FSAR and, therefore, were not considered.

mi0180-0377a-43

• l' 6

Item 6 Provide the following information as a function of time:

a. Minimum DNBR.

b. Cladding temperature if DNBR limit is exceeded.

c. Feedwater flow into faulted and nonfaulted steam generators (main and auxiliary) .

d. Steam generator liquid mass, heat transfer area covered, heat transfer rate, and pressure. *

e. Break flow rate.

f. Other steam release rates in secondary systems.

g. Primary system pressure.

h. Pressurizer level.

i. Hot channel flow rate.

j. Core inlet and outlet temperature.

k. Pressurizer safety/relief valve flow rate.

• 1. ECCS flow rate.

The analysis should be carried out until the effects of delayed neutrons and moderator feedback have turned around and the subcriticality margin is increasing.

,_ Note: The DNBR calculations must reflect the initial plant perturbations due to moderator and pressure d.ecrease and loss of off-site power (if appropri-ate). Also discuss how the effects of a*stuck rod are considered wheri calculating DNBRs after the rods have been inserted. If fuel damage occurs (ie, violation of DNBR), provide fraction of fuel that.failed and off-site dose calculations. Also provide and justify DNB correlations used in the analyses.

Response to Item 6 The transient information requested here is provided in ENC Report XN-NF-77-18 and in the FSAR as discussed in the response to Item 1 of this request.

B. CONTAINMENT PRESSURE Item 1 Review your current analysis of this event, and provide NRC with the assumptions used during this analysis. Particular emphasis should be placed on describing how AFS flow was accounted for in your original analysis.

(Reference to previously submitted info:r;1llation is acceptable if identified as to page number and date.) Any changes in your design which would impact the conclusions of your original analysis should be discussed. We are particu-larly concerned with design changes that could lead to an underestimation of the containment pressure following an MSLB inside containment.

mi0180-0377a-43

.. 7 Response to Item 1 The results of the current containment pressure analyses of main steam line break are discussed in Amendment 14 (Section 14.11) to the FSAR. These results are based on assumptions consistent with those presented in FSAR Section 14.18. The important assumptions made in this analyses are as follows:

1. Two of the three available containment spray pumps were assumed operable.

2. No containment air coolers were assumed operable (three of the four existing air coolers are equivalent in heat removal capabi--1.ity to two spray pumps) .

3. No credit was taken for moisture carry-over during.the blowdown, ie, a pure steam blowdown was assumed.

4. A worst containment heat transfer model was used consistent with that described in FSAR Section 14.18.

With regard to feedwater additions to the.steam generators, for the full power case, it was assumed that the feedwater flow was reduced linearly from full*

flow to zero during the 60 seconds following reactor trip. This is believed to be .somewhat conservative based on the considerations described .in Part A.1 of this response. No consideration was given to auxiliary feedwater in the current analysis. The analysis perfo~ed by CE in response to this information request (attachment) demonstrates that the addition of auxiliary feedwater (assumed to be runout flow from three auxiliary feed plimps which were started simultaneously with the break) will not cause containment mi0180-0377a-43

8 pressure to rise above the initial blowdown peak. The Palisades automatic auxiliary feedwater initiation system is designed to actuate auxiliary feedwater after a predetermined time delay, and to limit flow to less than a predetermined value. The flow limit has not yet been determined, but will probably be less than 250 gpm based on feed line water hammer considerations.

It is proposed that the delay be set at two minutes (+/- 30 seconds) for the reasons discussed in the answer to Question 2f below.

Item 2 Provide the following information for the reanalyses performed to determine the maximum containment pressure for a spectrum of postulated main steam line breaks for various reactor power levels for the proposed AFS design.

Item 2a Specify the AFS flow rate that was used in your original containment pressur-ization analyses. Provide the basis for this assumed flow rate.

Response to Item 2a Auxiliary feedwater flow was not considered in the original containment analysis. This assumption was based on the auxiliary feedwater system being a manual system.

Item 2b Provide the rated flow rate, the runout flow rate, and the pump head capacity curve for your AFS design.

Response to Item 2b Rated Flow Rate: 415 gpm at 2,730 feet.

Runout Flow Rate: 750 gpm from one pump.

Pump Head/Capacity Curve: Refer to Figure 9-11 of Section 9 of the FSAR.

mi0180-0377a-43

9 Item 2c Provide the time span over which it was assumed in your original analysis that AFS was added to the affected steam generator following an MSLB inside containmemnt.

Response to Item 2c Auxiliary feedwater flow was not considered in the original MSLB containment analysis.

Item 2d Discuss the design provisions in the AFS used to terminate the AFS flow to the affected steam generator. If operator action is required to perform this function, discuss the information that will be available to the operator to alert him of the need to isolate the auxiliary feedwater to the affected steam generator; the time when this information would become available, and the time it would take the operator to complete this action. Define credit for operator action. If termination of AFS flow is dependent on automatic action, describe the basic operation of the auto-isolation system. Describe the failure ~odes of the system. Describe any annunciation devices associated with the system.

Response to Item 2d With the manual auxiliary feedwater system, it was intended that the operator would not initiate auxiliary feedwater flow to the affected steam generator.

The plant emergency operating procedure for main steam line and feed line ruptures instructs the operator not to feed the affected steam generator. He would recognize the affected steam generator by observing pressure in each steam generator and preferentially feeding the steam generator with the higher pressure.

With the automatic system for auxiliary feedwater actuation it will be necessary for the operator to terminate feedwater to .the affected steam generator using diagnostic techniques identical to those described above.

With the intended design of the automatic actuation system, it is not mi0180-0377a-43

10 necessary to take credit for early operator action (before 30 minutes) to limit containment pressure to less than design. As noted previously, it is intended that auxiliary feedwater flow will be limited to less than a predetermined value (probably less than 250 gpm) and will be delayed by two minutes (+/- 30 seconds).

Item 2e Provide the single active failure analyses which specifically identifies those safety grade systems and components relied upon to limit the mass and energy release and the containment pressure response. The single failure analysis should include, but not necessarily be limited to, partial loss of containment cooling systems and failure of the AFS isolation valve to close.

Response to Item 2e The safety grade systems and components relied upon to limit the mass and energy release and the containment pressure response are:

1. The containment spray system (three spray pumps).

2. The four containment air coolers.

3. One main steam line isolation valve per steam generator.

In addition to the above, one control grade system - the main feedwater control system - is relied upon to limit the feedwater addition to the steam generator and, thereby, the mass and energy released to containment.

In the event of loss of off-site power, a diesel generator failure would minimize the amount of containment cooling equipment available to mitigate a steam line break. Two spray pumps and one emergency air cooler are supplied by one diesel, and the remaining spray pump and three air coolers are supplied by the other diesel generator. Only that equipment supplied by either diesel mi0180-0377a-43

11 is needed to limit containment pressure to less than design if the MSIVs and the main feedwater- system function as intended.

Failure of the MSIV in the unbroken steam line to close would result in the release of the inventory from both steam generators into containment. Due to the nature of the MSIVs (check valves held open by air against. the normal flow of steam), it is highly unlikely that either would fail to close on demand.

The signal which closes the valves (two of four low steam pressure signals from either steam generator) is fully redundant.

Failure of the main feedwater system to terminate flow to the steam generators following reactor trip could cause an excessive steam release to containment.

If off-site power were lost, all feedwater to the steam generators would be terminated because the condensate pumps which are electric driven would trip and, thereby, cause the main* feed pumps to trip. In addition, MSIV closure would eliminate the source of steam (other than the residual steam in the steam lines) to the main feed pump turbine drivers. With off-site power available, MSIV closure would stop steam to the main feed pump turbines, but the condensate pumps would continue to supply feedwater to the broken steam

• generator. To prevent this from occurring, the_ plant is presently being modified so that the main feed regulating and bypass valves*will be automatically closed on the same signal which closes the MSIVs. The main steam line isolation set point c~ 500 psia) is approximnately the shutoff head of the condensate pumps, hence valve closure at that time would stop delivery of water to the broken steam generator by the condensate pumps.

mi0180-0377a-43

12 FSAR Worst Case The main steam line break pressurization event analyzed for the FSAR assumed the double-ended guillotine rupture of a main steam line in containment. A reduction in feedwater flow from full flow to zero over the 60 seconds immediately following reactor scram was assumed. Both MSIVs were assumed to close on low steam generator pressure (500 psia). For the full power case that was analyzed, off-site power was assumed to be available to maximize the rate of energy transfer from the primary to the secondary and, thereby, maximize the mass and energy release to containment. Further, it was conservatively assumed that only two containment spray pumps were available to remove energy from containment (ie, no credit was taken for the three air coolers and one spray pump which are powered off the other emergency safeguards bus and would be available if off-site power were available). (In other words, whereas only a diesel failure coincident with loss of off-site.

power could take out the air coolers, the blowdown was conservatively based on the availability of off-site power.)

The above failure assumptions are believed to result in an overestimate of the peak MSLB containment pressure. The Systematic Evaluation Program design

• bases events review will reconsider the main steam line break event and those systems utilized to mitigate its affect on the containment.

mi0180-0377a-43

13 Item 2f For the single active failure case which results in the maximum containment atmosphere pressure, provide a chronology of events. Graphically, show the containmnent atmosphere pressure as a function of time for at least 30 minutes following the accident. For this case, assume the AFS flow to the broken loop steam generator to be at the pump runout flow (if a runout control system is not part of the current design) for the entire transient if no automatic isolation of auxiliary feedwater is part of the current design.

Response to Item 2f The MSLB analysis performed by CE in response to this request (attachment) shows that the containment peak pressure is determined primarily by the initial inventory in the ruptured steam generator. Even though the analysis assumed three auxiliary feedwater pumps (Palisades has only two) to be providing essentially runout flow c~ 2,500 gpm) to the broken steam generator, the containment heat removal systems were found to be more than adequate for removing the added energy being released to containment.

As part of the Palisades automatic auxiliary feedwater. system design, flow to either steam generator will be automatically limited to less than one-tenth (250 gpm) that assumed in the CE analysis. Further, the system design is such that automatic initiation may be delayed by a predetermined amount of time.

In order to assure that any auxiliary feedwater added to the broken steam generator will not adversely affect the containment, it is proposed that the delay be set to two minutes (+/- 30 seconds). This is longer than the time to peak containment presure for the main steam line break containment pressuriza-tion event analyzed for the FSAR (the peak pressure of 51.8 psig was calculated to occur at about 68 seconds, see Figure 1) but is soon enough to assure that more than adequate steam generator water inventory will be maintained in a loss of main feedwater event.

mi0180-0377a-43

14 Item 2g For the case identified in (f) above, provide the mass and energy release data in tabular form. Discuss and justify the assumptions made regarding the time at which active containment heat removal systems become effective.

Response to Item 2g The mass and energy release data for the full power MSLB case discussed in Amendment 14 to the FSAR is provided on Table 1. As noted previously, auxiliary feedwater was not considered in the FSAR. We have asked our Architect Engineer to recalculate the containment pressure response assuming that at two minutes, 250 gpm of cold feedwater would begin entering the broken steam generator, immediately flash to steam, and be transmitted to containment. With regard to containment heat removal systems, the activation times assumed in the FSAR will be likewise assumed in this reanalysis. This analysis will be completed by March 31, 1980 and provided to the NRC at that time. Based on the results of the CE analysis (attachment), we anticipate that containment pressure will not exceed design.

mi0180-0377a-43

: **r.:.-;*. . . . - ,.rT.  :

" ... !....... ,. ............. .!..... *~* .. *'-

1 , - l . *1* *~

• r-1, . J. ~-T-:-1---:-

B' [__

• *I' i

• .. . ...... ' . . 1.. . . i:

"' . . I* .J.. . ....  : I . I . . I rn.

• * * * *. *. I* *

• I * * * ** 1.' * * * * ** * * * ' * * '

I.... ' '.

__ JI ____ ;___

--~ j___I_:, ____ --~J 1

l ___ _:_~---~-:-~_ ,~-- __ J___ :______ _i__ ~ . _______ 3/79____ _

_____ L__ :_

--~ ~-*" *:*, __:_

I  ;  ;

.~. . .!::~ ..... !.... _ ;_ / 2..*\.*\-7-11~07 I I I . . I I

-- .. ;:__ ._ .l. 1 -~ ... ! . :

. ..... I I ..; . II . . .: , . . I - *i 1*.:

.I . . . . ~ . . .  ;

..::?..t:-:..1-1..:: i".cr..~J>.-:-, *r1:.~,.,J:1,,.P.l\T___ 1 I '

• ' 1

• I '

• I f--..! .... *--~----- _____!; ____* .. . ..!I............ li:-.:.i>~c:..1 1 .-*~---1----.:.. .....1.. :..).__ ---,~-- ....:..:--1------L:-

. . . . ,. . .: *-... . . . '1 . '; . ... . .. I. . . .. . . I

. '\'.= I

' j. . .1...

I. . . . ;I . I . ..

--* I** . '. * * . l I I ...

1* **

. !I ....

1- _:_

I I

I I

0 f* *

~m* \(*\f"\,(;..P.J ... .:By1l.-D1N (;, ..J?.fl..~~'.'.>*.lf'.C.:.J:R~1~s1i::11t:_~ __ ,:__

._.c

.. . . ; I ;I  ; Cl'>_J. A i I I I i I ' '

I--,

~-!'. F:o~,.

.; I  : *1 . . II r ; *I

. I . .. . I. . .. . I . . <.:'

---

------, .... ___ ,. _. _____________ ,_. _ ::> u 1.:-

1

' ' ' ~ Lo . ..  : f\ .I l ')b 9 ... \'\ e>..n.' _ ~ .1 - [!..,"' t\.

,

• L1 tJ-. ~ ' \.......

- '~ 1<.. 1. *. - *-* ---

.::.  ;  ! *:

• 1*  : . I  : I *  :

• I  : i  :

1*

. v ":-.R.\C.l},!.<,.:_:i:.n...'ll':lhl.-c.* pt:~1:*'.!.!.~\!."\0:::.1.'7._.i

. ~ ---,,<>[.R'- - -

.~Ll:-:t'(ll6.:-1_*:,- :-~- .......!'.'..!...~~<.:--$

I I:.. ::* I . : .. *j. : .. ' I* ; 1 -I* . .,... .;: . .I .. : ..... .

1

---T-*--1--*---* - *:1"*::j*.

--,--..-- *----r*--i---,-- --.. I..

-*-T*-**--* --....,--- *:'.

__ ,.___ l _______ -i---1.. . .. i  : .  ; .. ..

f  :-  :

l .. . . ... .  ; ..

I*. .

f ,  : "

1-. . * * ** * * * * * ,. *

**:*.'-...  ! i i l zi *  :-.. I }-

--1*--- --r-- --* --~ *---+-1-+/-j--*__j__. _ ._ _ ---;*::-:-T:::-*-

t ".

---

r----j--- _,_ ___ ---*--- *-**. i--

('.J q-

. :: : *:I -. k.9-_. ::.: . : . . :.:: ::." .. :i

• I*..___: *-.. . "":___:_. _. _..:.__ l*(:_:_ .. ........:.. I ... . . I I:.:....,:...:.. I'

. : :.... :: ! : : : : : : : ::: ~ : !:"HY:::! :. :

I ..... 1- .... *1* ...:4:.:..:.r.~--:...:.. , ... J:::, ... I L.--i---i::;:

I :-;.. , ..

N

..... ~--*-  :...:~ -~...:..-~~-: __ :.......:.:..:...:_+:.:...:...' .-- . h.:.B . . .. :.:....: 1-~*.

\.!)

I _.:_* *___ ----'* - -*  ; I* I *: -T" - *oc.

...Lt.Ol~T_..:.._l_~~-.L.....'-

I *' 1

• I ,' * * * *

j.. .:.-:.:-:.. _:... '.:. :. -.:.

-* --'- ..:..L..: . . . ' * '

=rl..__:'_______. ;~_*:1  :~ . "1- _:_~. i __ . .I~l . ." :. '

"i°

---

~-_.,......:..: I- . .. :~.! J.. .:.~l I *  : ..

I .-**-'-*-..--*'1*, ______

I  :.

• *

• I * * ' * * * ' *

. .. I I I. ' I i I .

_:_.i. __:_ ___

. I 1* .

. I

• t.. ' .. -,I .

bf'" . . . .

0 .. ,.. . " 1'

• I ** ,.  ; ~ ~ ... I

' :Cii "._)"'" Ila- :

. . "i'"*H .. :

, ... I . . v--~

-:~- ... ' '. ,- -~ :. ' *. * ..

~: I . . I. :..

~~~;~;*:~.:~: ::1::;:: :::~* -~T~-1--*~-~-~--~ .~. *: .. :. r~f-~-- *r--~~Tc.?~~. ;~iD'.\*

.. . !!J .... -; . . . . . ' . *-'- ***-* . I .. . ... ' .

• I . . ... j .... .. \ . .. .. . I ..

I ..

;r . 0 ' . .  : * . I ........

~]~ t :u~I-~~+

.. ' H***'***  !***** 1-Bf2 r i stI j~ -!21r~1i11t_ --u~t-I********' 1** I' .... , **1* I

t u~

+

~::;

1.,1: ... J1* H*!**:***I* **1 **1'*1**1**

i**--!---*j=-=-i6___*, :,.:. ~. :-: ~---\>T-::f*--l-::*t---- _, ~=-':J~:;~e~ 1-~.--1<

r-35

..o ...* i:

_J

~

.. I . :* .. ". " .'1*"** .1. . I . I ' .. I " . 'I I. " . I '

. I

---+..:.**++fr I .... '  : II

.~.;

! I * ...

x::

I -~:<_::r1 -:-L__ * .:*-:H> --:-<I:-:*: \"c"~ ..1$ 1N*:'~*'-'car:1'~,:"~:.". T~r'\y~;,i\7"o~cr.

.... .... ,. *L**.... 1..... ;. . .1 .. j I . *::.:I.. . 1** *..._J1____ .. ;

~-!-*.. * ~-*_.---~ ~~~_: ___

I . , ._.

1

• -- **** _ .- t * ! :_:_:._ . 1 ** - *- * - *_. _ _ * - '_ .

~;

I1--- .. j~:1-:--:~-- . _

I ...

• 1 . .. ; , ,

I.

• . . j. .. .. . I * .

• I. .

.I *

• 1* **

1 .' .,. ' .' . '

T.t . *:1~~~r:::**---:1:-;_***-:*:1:***** -~:~*1--

• i .. I* .) . . , . . .  ! . . . I* . . . . : .... I *.* ... 1

-:i-~-* :::~*::.:**--**-~\::** ~:~r0: ----r~--- ,~

b( 1,  !' '-}- --+- >!-- f-'  !-- c+' cl -+-1  :+ ' r h:-

-r-:-)~- *~~-+-

.. 1:rr-- -! . ~==- -f=~~1-~r+--10 JI ' I ' h . . I J

___ -cr-T 1---.---

--T~1-*

.. .,,,1.... ______

J.

I...

• I I..... ..,I!

l.I'

---

r~- _,I_____ I;_fil_~---.TH'\l:.. ! SC.<:.~r*l,~)

I I I '

t (")

I*

!' I:

,O I' .

~--- _..:.:.! .----~*-* _ ~

I

~

iI' I

1

--"-------1---~----i"---:--

I . L ** * * ' I . I * * '

• 1

__ j_: _____: _ .. l I  : ..

!- ... :* l -: . : . *I 1:: ".!

. . I..... I :" : :.. l  ; . . *:  ;

~

!. *-1:* .. 1 ... *:*:!*::* :::*1 I

--*1

' II I

i....

tt* :

I

• 1 I '

.. : *I ...

. i __ ...t ~J __ --~!- * '. _~ __ _l ___ j_ -----~ ____i __ _.l__,_* ______ L .. ___ _

mi0180-0377b-45 1 of 2

--*------ -- - - ---------* - -- - --~ --- -- - - -- - -- -- - - - ---- -- -- - ---- -- -- ------- - - ----- - *-- - -------- -- - -- -- ---~--

TABLE 1 (Contd)

Isolated Steam Generator Time mA PsG h (Sec) mN *(Lbm/h) (psia) -(Btu/Lbm) o.o 1.43 l.656E07 760 1200.4 0.1 1.43 l.656E07 760 1200.4 0.2 1.43 l.656E07 755 1200.6 0.6 1.37 l.587E07 725 1201.3 1.0 1.32 - l .530E07 700 1201. 8 1.3 1.29 l.496E07 685 1202.2 1.5 1.27 I. 473E07 670 1202.5 2.0 1.23 1.416E07 - 645 1202.9 3.0 1.11 1.279E07 60S 1203.6 3.S 1.07 1.233E07 S90 1203.8 4.0 1.00 l.1S3E07 S7S .1204.0 s.o 0.86 9.930E06 S60 1204.2 S.4 0.79 9.13SE06 SSS 1204.2 6.0 0.66 7.6SOE06 sso 1204.3 6.4 O.S7 6.622E06 SSS 1204.2 6.8 0.49 S.709E06 S60 1204.2 7.2 0.38 4.339E06 S6S 1204.1 7.8 0.13 - 'l

-* l.484E06 S80 1203.9 8.1 0.0 0.0 S90 1203.8

>8.1 o.o 0.0 mi0180-0377b-4S 2 of 2

i, I ATIACHMEtlT BEST ESTIMATE r1SLB Mll\L YS IS TO J\SSESS t~SSS ANO CONTAINMENT RESPOMSE \</ITH AUTm1ATIC AUXILIARY FEEDWJ\TER ACTUATION Combustion Engineering January, 1980 soo12so SQ.r

1.0 IrlTROOUCTIOil -

- - - --- - - --- ---- - -- ____,_ -~ -~- --- --- - --- -~- - - --- - -- - --

-- - -- - - ----- A--set--o-r-ca-icu_TatTons-tias-bee-n -rerfonned on a genP.ric basis \*1ith plant characteristics representative of CE oreratinq plants to model containment buildinq f)ressure and ter.irerature response and overall rtSSS behavior, including core reactivity, following a Main Steam Line Break (i1SLB) inside containment. The intent of these calculations is to deterr.:ine if the containMcnt building response (pressure) and the core reactivity response

{return to power) are acceptable followinq a MSLB when auxiliAry feed-water is added riithout regard to the identification of the affected stea~

genera tor. The aux i 1i ary feedwater fl ci\*1 is assumed to be activated at the initiation of the transient to maximize its effects. Main feedwater flow including post trip rampdown ii simulated. No isolation of main or auxiliary feedwater is considered unless a high v1ater level condition is reached.

2. 0 ASSUt !PT roris Arm CASES Assunptions for the analyses are given -in Table l. The four cases analyzed are listed in Table 2.

3.0

• DISCUSSIOil OF RESULTS flaximum containnent pressure and least negative. core reactivity for the four cases ar~ listed in Table 3. Both the containment pressure and the reactivity (return to power) values are within acceptable limits.

Main feedv1ater flmv, auxiliary feed*,*tater flo\'1, core reactivity change, core pm1er, containment pressure, primary 1oop tempera tu res, and steam generator secondary ter.iperatures for the four cases are detailerl in Figures A-1 through A-7, B-1 t~rough B-7, C-1 through C-7, and D-1 through 0-7, respectively.

The results of the analyses using b~st estimate models for steam generator moisture carryover and containment passive ~eat sin~ heat transfer demonstrate that the additional auxiliary fed1*1ater has a negligible impact on containment peak pressure. The containment [)Cak rressure is detennined primarily by the initial inventory in the ruptured unit. This

inventory is released within the first few minutes, depending upon the break

~ -- -*

size, so that the contribution of a_u_xi_lj~TY__feed1§_t~_r fJ0_1Lto the _______ _

ruptured unit over this ti~e frame is small. Over the lonqer ti~e frane, n

the secondary inventory is boiled off at essentially the decay heat rate which the containment active heat refiloval syster.is can accor.modatc 1*1hilc

- reducing containment pressure. The excess feedv1ater \*1hich is not *boiled off remains in the steam generator, causing the secondary level to rise.

The containment peak pressure is essentially an initial inventory limited phenomenon.

The results of the analyses also show that the additional auxiliary feed-water has a negligible impact on core reactivity. Cases A and C assume no stuck rods and a best estimate moderator cooldown curve. For conrarisoo, Cases 8 and 0 assume that the most reactive rod is stuck and tha: the moderator cooldown curve is a licensing curve. All cases took credit for boron injection via three charging pumps; however, safety injection boron credit was not taken. These cases do not have a return to power for the following_ reason. The initial primary loor temperature decreases are linited bY the two-phn.se blol'idol'm process associated viith 1arqe break P-2 ft 2),

. since much of the break flow is saturated liquid wl1ich has not absorbed significant amounts of energy from the primary loop. For smaller break areas (~2 ft 2 ), the blm*1dovm is pure steam \'lhich rices.require lar<Je ar;;ounts of energy per unit mass to boil via primary to secondary heat transfer; hm'lever, the rate of primary-to-secondary heat transfer is controlled b.v the blowdown flowrate which in turn is limited by the small break area. The net result is that over approximately the first 100 secnnds of the event, the amount of core and loop coolrlown is about the sa~e regardless of break size. This time frame is most important since .the presence of delayed neutrons minimizes the amount of cooldown needed to produce a core criticality problem.

Without a return to po1-1er (via 11rimary loop cooldO\*m and delayed neutrons),

the remainder of the transient is a gradual increase in reactivity due to loor cooldown l'lhich is coupled to the containment 11ressure~ f)lus a decrease

• in reactivity due to boron injection. In time (approxi~ately JOO sccon~~),

the reactivity decrease due to borafion overtakes the reactivit! ir1crcJ~cs due to loop cooldown; thereafter, the total reactivity steadily decreases.

The ruptured steam generator is at the containfilcnt backpressure and rlith

. ~~* '

RCPs operating the sensible heat from the non-ruptured unit is quickly removed resulting in RCS and SG secondar:1 temperatures essentiall.v in

~ ~- --- *- --

equ i1 i br i um with the con-tainmcnt conditions in--ab-out 10 m1nu-te$.-

With licensing assumptions, the peak in the reactivity transient is calculated to be 1*1.ithin the first t1*10 r:iinutes of the event. A~ 3 minute time delay, if added tQ the autonatic actuation circuit, would justify a statement that automatic auxiliary feed1*1ater actuation r'lill not inpact 1

existing SAR core cooldown MSLB analyses.

4.0 COMPARISON \HTH LICENsrnr, CALCUL1\TIONS Th~ following items are important in comparing the results contained herein with those obtained with traditional licensing models and assumptions:

1. The moisture carryover model used is a best estimate model l'1hich gives a two-phase b1Ol'tdo\'m for 1arge break a re as. The ti.-10-phase bl m*1dovm results in a lower containment pressure and less initial pri~ary loop cooldown than a pure steam bl01*1dm*m. Chapter 15 analyses assume_ a pure steam blowdown regardless of break size.

2. Chapter 15 analyses assume that the most reactive rod is stock.

Moreover, the remaining rod wo~th i~ assigned a conservative value in conjunction with a conservative moderator cooldown curve.

3. A best estimate containment heat transfer model *provides containment pressurization results significantly 101*1er than those provided in Chapter 6 analyses.

,. 'I TAGLE l

\ '

ASSUM PT I Otl S NSSS Initial Conditions Power '*2700 Ml-It Core Inlet Tenperature 548°F Prinary Pressure 2250 PSIA Secondary Pressure 875 PS !J\

Secondary Temperature 529°F Containment Data*

Free Volume 2.5 x 10 6 ft 3 Design Pressure 44 psig Heat Sinks SAR values Heat Transfer Model Best esti~ate model Number of Fan Coolers 4 (no single failure) 6 Fan Cooler Capacity, each 68 x 10 B/hr at 280°F containment te~oerature 100°F CCW Tempe~ature Fan Coale~ Actuation Setpoint Fans are operational

@t = a Nu~ber of Sprays 2 (no single failure)

Spray rate, each 2700 GPM Spray Actuation Setpoint 10 PSIG + 60 seconds

• -*Other Data Steam Generator Isolation Signal (MSIS) setpoint 500 psia Decay Heat Curve AtlS-5 Main Feedwater Flow Ruptured Unit: Ramped to 10~ over 60 seccn~s follov1ing Rec.ctcr Trio: (HL represents twice the bypass nominal 'laluc of s*:, this accounts for ru~p run-out wi:

reduced bacK8ressure),

temp6ratu~~ ~s r~duc~d to lC~

to account for turbine off-line. Flo\-/ t::r:.. in::itod if ti:::

e 1c va t i on o f u :; : :; r ~ ,_:- '.' e 1 t ao

s reached. See Fi c

;ures A-1 ,

B-1, C-1 , and 0-1 .

~E 1 ---- continued Main Feedwater Flow -- continued Unaffected Unit: Same as ruptured unit except that flow is ramped to 5~. See Figures A-1, B-1, C-1 and 0-1.'.

Auxiliary Feedwater Flow Ruptured Unit: Initiated at t = 0. Flow rate is a function of unit pressure. All control valves assumed to* be fully opened.

Unaffected Unit: No flov1; .all fl01*1 is totally diverted to the ruptured unit.

Reacto~ Coolant Pumps *Operating during the transient.

CEA Insertion Worth All rods in (ARI) -8.9% (no stuck rod)

?1ost reactive rod sttick "'.'7 .12% (best *estimate)

Moderator \forth SAR Value See Fi ~tire Best Estimate Value See Figure 2

. Doppler Horth See Figure 3 Moisture Carryover On Steam -~*

Generator Secondary Side Best Estimate r.1cidel Boron Inj~ction Parameters Safety Injection Credit Hot Taken Charging Pur.ips

!lumber of Pur.ips 3 Fl O\'/ P-a te 44 GP~1 rier rrnr.ip J\ctuation Tirne SIAS Boric Acid Concentration 8~~ by \*1ei0ht Boron Horth no PPM/~~

Boric Acid Conversion Factor 1749 PPr\ boronr~ b~' wei0ht boric acid f1i xi ng t1ode 1 Used Slug Flm! 11odel Loop Transit Ti~e 10.5 seconds

• .; t TABLE 2 CASES

~ CEA Scram 1.-/orth (~)

• 11odera tor Curve Creak Area (Ft 2)

A -8.9 Figure 2 6.63(l)

B -7 .12 Figure l 6.63{l) c -8.9 Figure 2 1.99( 2 )

D -7 .12 Figure l 1.99{ 2 )

(1) Double-ended severance of main stealil line (t\*10-phase blo\'tdm-m).

(2) Largest break area corresponding to pure stear.i blowdm-1n.

T!IBLE 3

'RESULT$

least Nega.tive ::.-:

Case Containment Peak Pressure (PSIG) Core React i vi t'I <~

A 29.7/83.0 (sec.) -4.31 29-.7/83.0 (sec.) -~.34 B

c 35.0/231 .9 (sec.) -3.54 D 35. 0/231. 9 (sec.) -1.55

FIGUl:E 1

• ' REACTIVITY VS t10DERATOR TEliPEPJ1TURE 6 (S/\R VALUE) 5 ~

Q

<J 4 ....

~

z 0

~

.~

w

. (/) 3: .

z

>-I

}-

>-i

~

I-0 2 lW 0::

l 0

-1 ' .

300 350 400 450 500 550 600

~ ,, 0 Dc J~\

J *. ..,.. ,-: :.> T :- . '. c. :- '.) .* -;- ,

l-1\1*\l*..... , ,

I :: ":

,~ , .* - i \ , .... ~ ,.....,,,._,

0 ,-~

7 .....

,. .:: .. FI c;ur,E 2

.. REACTIVITY VS rtODEMTOP- TE11PERi\TUf":E 6 (BEST ESTP~/\TE l/J\LUE) 5 Q

<l 4

• .~

. z0

~

.~

L.1..f V> 3.

z1-i

}-

)---(

t-(

r- 2 u

w 0:::

1 0

-1 300 350 400 450 P..*\Q J'i

• oc) ', -:-,; rJ L ' \ '. ** l 1

..... ~ \

j":"'

• l "-. I 0

I *

~:")\~

• 'i",

• • ** ~

I~':

v ~ ' i... t o,c

.*- ' I.

• - J*

~.v

. 1.!3 ~t FIGURE -

t ... DOPPLER REACTIVITY VS FUEL TC:PER1UURE l - --

.. 1.6 1

• r

.. **-* .. ~

Ii 1.4 i ... . .*-

. . , .2 ~

I - . ******- ...... *-* *--** *-* .......

1.0 l

~

... i I

• o.n I

~

-. Q__.

<l .

~ *-- *-- . *- ****-** -* -**-* ........ .;.;. .*.... ***-

z C'

I-c::::

w 0.2 . ..... .. . -- .... - ...

t

~..

(/)

-: 0.0 500 1000 1500 2500 l

i I-

. -0.2 . *. *- -... -*- ... . . .. .. .. -

.* - .. t."' ~ - ** ***-** .. : ....

I-u  !  :*. :..

< -0.4 ...... -* .. - .... **- . *- ..... - - - .......... -**- -. - *-*-*** - *-***-**---*----- -

w et.: I .. .

-0.G i

. l

... *-* -0.8 ~ ......... -*-***-* __ ___ . **- ------* -- --- -

f

-1.0 ~

.I

.. - - i

- .I I

I iI ..

I

.1 I ..

FUEL TE11PERJ\TURE °F

1800 j

16CC u FIGURE A-1 w

Cf)

!:: t\AIH FEEO\IATER co 140C FLO\-/ VS TirlE

_J 3:

0

_J Lt... 1 ti(')("'

'- u \..)

cc: .

~

1--

.C:

3:

Cl

.u w . 1OGC Lt...

z a:

800 6CO

• 40C 20G AFFECTED u;1 IT UtlAFFECTED Uri IT 0 L---------1--___.l....--__J_--__l_----=-::---J 4GO BGG I ( ' ,.-. ; ' ........ ""I""'\..

0 200 6CG * ._, v .. ;

• t .. ~ v

1. I l......~ '" S =-.. ,..

I

.., uu ,-----------------------------~

360 -

FIGURE A-2 320 -

Et1El1GEtlCY FEEO!-IATEP. FLOH VS TillE 1l0 FLO\/ TO Ut!AFFECTED Ur!IT

- 280 r-240 I...

3:

0

__J I

LL er:

LJ...j 200 ._

I- .

a:

3:

0

~

w LL

• u 1so L z

LLJ C..'.J er:

w

~

w 120 I II I

I 80 l..

i I

I 40 IT I

I I

• -. *;1 1u r--------------------------------.

9* FIGURE.A-3 REACTIVITY CHMIGES 1 vs.

8 TH1E 6

110DER/\TOr.

4 I- 2 z

LLJ u

CY-IJ..j DOPPLER 0...

0 I-u a: -2 w

Ct::

-4 TOTAL

-6 'I

• ~

-8 CEA

.___ _ _ _J__ ~~--'~~:___J_~~~--J

-10 -

0 200 400 sco sec ~ cco T I ~1 E ( S :. C l

$1 \ , I 1 11,,,J

.. e v

O:a:J FIGURE J\-4 0.3 0 CORE PO\.IER VS TI!1E

o. 7 o 0.6 0

(!'.:

0.5 0 w

3:

0 CL w

(!'.:

0 0,4 0 I u

0.3 ° 0.2 0 ~

0.1 0 o.n o 0 200 400 600 8GG TIME est.CJ

FIGURE /\-*5 55 COrlT/\IllnrnT PrtESSUrtE vs TH1E SG 45 a:

( j) a.. 40

• w e:::

~

(j)

(j) w 35 e::::

CL f-z w

z . 30. I a:

f-z a

u 25

. 20 15 10 L--~~~-1-~~~~1--~~~---L-~~~~'~~~~----~~~__,

0 200 400 6CC 8CC lCCC TI ~l E t SEC l

6 0 0 _ _ _ ____,

FI GU.RE A-6 PRirtA.P-Y LOOP TE!1PEP..J\TURES

.!~ vs I *.

• TIME 480 COLD LEG OF ur*IAFFECTED Uil IT 420 HOT LEG u....

0 360

' .. - ..."""'"'* w 0

(/)

w c.c:: COLD LEG OF

=> 300 /

>-- AFFECTED unn a:

~

c.i:::

w CL

.w

>- 2. i 0 1

C:...

0 b

_J 180 120 60 0

0

• r r '*' :- r -:.' ;: r- 1

FIGURE f\-7 540 STEA11 GEllERATOR TEilPErJ\TlJRES .

vs ..

THIE ..

480 120 6G 0 _ __..______ L_ _ __.__ _ _ _---1._ _ _ ___,1.___ _ ___,

0 200 400 60C 8GG lOGG T I t*l E f ~ EC l

~------.,-.-.-**-*---* .. -~ * .............. J****

• J 1800 '-

16GC -

u FIGURE i3-1 w

(./')

'co l::

....J l 40C I ttAIN FEEO\IATER FLOVI VS TIrlE 3:

0

....J LL..

1200 ,__

a:::

c:

3:

0

.i.;

w lCGC .__

LL..

z cc l::

800 .__

6 ('\("I vv -

40G .__

200 I-AFFECTED u; i IT UNAFFECTED Ui:IT I

I 0 L-~~~~--~~~-1--~~~-l'~~~~-'L--~~~----~~-__J 0 200 400 sec 8GG

._ , . 'P 360 FIGUP-E B-2 320 E11ERGErlCY FEED\!i\TEP- FLO\.!

vs Tir1E rw FLO\-/ TO Uilfl.FFECTEO 280 UMIT u*

i..U Cf)

r:

co

--1 2 ,i 0 3:

0

--1

. LI...

ci:::

..u 200 I-.

a:

3:

0

~

w LL.

160 u

z tJ..j

(.'.)

ci:::

w

r:

..u 120 I

80 II.

40

... , r' FIGUr?E B-3 . '

REJ\CTIV ITY CHNIGES vs 8 TWE 6 t100ERATOR 4

I- 2 z

w u

0:::

w DOPPLER o_

0 -

I-I-

u a: -2 w

0:::

TOTAL TOTAL

' -4

-6 CEA

-8

-10 I ~

0 200 400' 600 800 1000 t20C TIME (ScCJ

FIGURE 3-4

.conE POUER 0.90 vs

' THIE .I o.i:o .*

o. 7 0 0.6 0 cc:

o* r.:O o .

UJ 3:

0 0....

w er:: 0 .40 0

u 0.30 0 200 400 600 BOG TIME f SECJ

FIGURE B-5 55 COtlTAHH1E11T PRESSURE vs TH1E 50 45 a:

Cf)

Q_ 40 LLJ c.c:::

Cf)

Cf)

LLJ 35 c.c:::

CL r--

z LLJ L:

z

..... 30 I c:::

r--

z a

u 25 20 15 10 L-..~~~-L~~~~Lf~~~~L-~~~--L-~~~----~~~~

0 200 400 6CG sec !OGG

• T HlE (SEC l

--....,.___....* ~-***'-,....._***--**-**. ~--*** *~-- .. _..._ *--. *-- ..... --.-*.... **- ---** ----- ................ _...... *****- -* *****--** - *... -* ....... .

• --***-*----*****--***""***------........ -*.----*-* ----*---***-OoPo,* oo ----... ---,---- -,, -**-~--* .. ------~---~*-.------*------

.. -*-~*,-- ......

• ,., \

• • 600 J

I FlGUP-E B-6

.:! P~Il1AP-Y LOOP TH1PERATU:;Es 540

. ,*vs

.Til1E 480 COLO LEG OF UW\ffECTED _WjJl._

420 *,

HOT LEG/

u..

0 36-0 w

0 en w

cc

J 30G COLO LEG OF ,.

~

cc AFFECTED Wl IT./

e:::

w CL l::

'-1.J I- 2.4 0 Ci_

.o 0

_J 180 120 SC 0 Gr-.....

0 200 400 sec *t n v

... r '*' ~ ' ~* ~ ('" ~

~- vvv FIGURE B-7

-~

540 STEA11 GHIER/\TOr- TE11Pcr:1\TURES vs TH1E 480 420 LL..

c.:>

w urtAFFECTED Utl IT a

w 360 c.t:::

. I-a:

c:::

w

- 0..

E: 300 i..LJ I-c::: AFFECTED un IT a

I-a:

c.t:::

w 2 Ir 0 z

UJ

(...:>

l::

a:

~

I- 180

(/)

120 60 I

I 0

0

___ _.__ _ _ _ L__,_*__J__

200 400 600

_ __ L _ _ _

8CG

....___J' TI r1 E 'r SEC J

lBOG FIGUP-E C-1 t*1AHI FEEn'.IATEP. FL0\-1 115 T Ir1E 1600 1400 u

w (f) r: 1200 xi

_I 3:

a

_I LL- 1000

~

..Ll 1--

a:

3:

Cl

.lJ w soc LL.

z a:

z:

600 400 I 20G AFFECTED Ur!IT UNJ\FFECTED U~I IT I.

I

_.__.1-1 BCO

--"-----.J

CGG 0 200 400 600 T f ~l E ( SE. C J

-r v v ..--------- * - - - - - - - - - - - - - - -

360 320 FIGURE C-2 vs tlO FLO\/ TO UtlAFFECTED Uri IT 80 40 0 400 6GC TI~IE (5ECJ

FIGUnE C-3 8

• RE/\CTIV !TY CHMlt;ES vs TH1E 6

1*10DER/\ TOR 4

I-

• 2 z

w u

e.r::

w DOPPLEf!.

CL 0

I-I-

u BORDrl cc -2 w

Ct:: .

-4

-6

-8 j

CEA

- 1Q .___ _ _ ___,___ _ ___.______ _L I

____ _J,;

400 60G SOC  ; OGG 1 ~,.....,....

0 200 " c:.. \J * -

T I ME .r 5 t: C J

0.9 0

• FIGURE C-4 CORE POUER vs *.

1 TI11E 0 .80

o. 70 Q,60 0.5 0 c.c::

w 3:

a CL w

~ 0. L()

0 u

0. 3 0 0.20 0.10 o.oo ~==-,,___l_'~~~~..J_~~~~'~~~~-l-~~~--'-~~~--i 0 200 400 6GO 8CO l OGG l 2CC TI~IE (SECJ

FIGURE C-5

.... COtiTAiilt1ErlT PRESSUf;[

55 vs Tir1E 50 45

. er:

Cf)

CL 40

.w C:::::

=>

Cf)

Cf) w 35 c.c:::

CL I-z w

~

z 30 er:

I-z 0

u 25 _;

20 15 I

I

_L _ _ _ __JJ .,.;

10 - - - ' - - - - - - - _ :_ _ _ __.1.-. _ _ _ _

0 200 **400 6CC 8CG 'en'"'

-.JU T I i*J E ( S t. C I

0 lJU e

(;

FI r;u r,E c - 6 PRil1/\P.Y LOOP TE11PEP-J\TUP.ES 540  !

vs TH1E

~

480 420 COLD LEG OF UtlAFFECiED UtlU__

lJ...

0 360 L:.J Cl

. Cf) w LI:::.

J 30C COLD LEG OF I-a: AFFECTED UN IT/

~

w CL L:

i..U I- 240 CL a

C)

_J.

180 120 6G 0 200 400 6CO 8 uUr.

("1 lOGO TIME (Sf.[J

" bUU - - - - - -

FIGURE C

• 7 540 STEN1 GENERATOR TD1PER/\TU~ES vs.

Tir1E 480 120 6C I

I 0 '-----~----L----__.__ _,______..__J

!OCG 0 200 400 6CO 8GO T IM E r Sf C l

' .*. ~uuu FIGURE 0-1 1800 t1AIN FEED\*!ATER FLQ\-J vs TH1E

• 1600 l 4 00 u

i...i.J Cf)

......... 11"'}

~u~

nr

~

a:i

. ......I

~

a

......I lJ... lOOC er:

.LI I-a: .

~

Cl

...u 800 w

Ll...

z a:

l::

6GO 40C I 200 AFFECTED UrlIT I

UNAFFECTED UN IT .

I 0 -----~--'-~~~~-'--~-'-~~----~- -~_!._'~-1...~~-'-~~~--il 0 200 400 6GO 8 GnU , Q 1 'U,..... C '* -->.:. ,-.......; ~-..

TfMF IC:.r!l

.. \'.,

400 .-----~

360 FHiURE D-2 320 Ef1EnGE!lCY FEEDUATER FLO!/

vs TH1E rm FLmJ TO 280 UflAFFECTED un IT u

w en L:

co ,!

_J 2 *iO ~

v 3:

0

__J tJ....

er::

~ 200 1--

G::

~

0

.;.J w

lJ...

160 u

.z i..LJ er::

w

I:

i..LJ 1

l~

no I 80 40 0 200 400 6CC *err-

• I.JV TTMi=" f',1-!1

ti' I ** ' FIGUr.E 0-3 REJ\CTIV ITY CHNIGES vs TH1E 8

6 M_ODEP.ATOR 4 ,

I- 2 z

w u

e::: DOPPLER w

CL 0

I-

~

u a: -2 w TOTAL e:::

TOTAL

-4

-6 CEA

-8

-10 .____ _ __..__ _ _ _.___ _ ___. ____ ----1._ _ _ _ _.._I_ _ ___,

200 400 600 800 1000 1280 TH1E (SECJ

i'J

.v

• "' I U

,~

e

.~

0.9 0 FI GU BE D-4 CO Rf PO\IER vs 0.80 Tir1E I -

I I

I

o. 70 '

0.60 o.so er:::

w 3:'.

0 a_

w e:::: 0.40 0

u 0.30 0.20

n. 1o 0.00 0 200 400 600 800 'er.('

• ._iv T H1E ( :5E CJ

55 FIGUP.E o~s CONTAIW1Ei1T *PRESSUf!E vs Tif1E so 45

. a:

I

(/)

a:...

40 w

c:::

(/)

(f)

w. 35 u::::

u..

I-zw L':

z 30 a:

I-z 0

u 25 .;

20 15 I

I 10 _ J _ _ _ ___.__ __.____ _.__1 _ _ _,____J 0,..,vU('

0 200 400 sec sec 1 T Ir*\ E ( SE. C J

,. " r '\*

• FIGURE 0-6

'PRIMARY LOOP TH1PERATURES vs TH1E 480 COLD LEG OF UtlAFFECTED U~l IT 420 u.. '

(.J 360

~

0

(.(')

w ci::: COLD LEG OF

J 3CC /\FFECTED UtlIT /.

1-- ----~

a:

~

w ...

CL

• L:

LlJ 1-- 240 CL 0

0

.....J 180 1 '? ("')

1 '- ....

60 0 L __ _ _-L---"--~'---_:,__J,__ ____ L.-_ _ __,. _ _ _-"-____,

0 200 400 600 BOG lOGG TIME (Sf.CJ

,'4

. i;.<

,,...,* 600

'W '~ ..

~ (,

• ..... FIGURE 0-7 540 STEN1 GEi!ER/\TOR TE~1PERATUP-ES vs TH1E 480 420 u.

C..')

w UNAFFECTED Uri IT 0

w 360 Cc:::

I-a:

Cc:::

w CL l.: 300 AFFECTED i.l.J I- UtlIT Cc:::

a I-

. a:

w CY- 2 iO 1

z w

Q

. :r:

c:

..u

>- 180 Cf)

I t

I 120 I I

I I

SC 0

0 200 400 6CG 80G lCCG

_J

~ ~ ,-.. .-

.. '- v -*

TfMF ( Sr. r I

\ l \. ~