ML19310A813
| ML19310A813 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/15/1980 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8006300543 | |
| Download: ML19310A813 (27) | |
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W Omaha Public Power District STATE 1623 MARNEY s OMANA. NE5RASMA 60102 a TELEPHONE 536 4000 AREA CODE 402 May 15, 1980 Mr. K. V. Seyfrit, Director U. S. Nuclear Regulatory Comission 5, { } =gg[~fil5V/i ~Ed.. l Office of Inspection and Enforcement Region IV
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611 Ryan Plaza Drive r :.!<:
Suite 1000 11 :
- ' ', ' I 9 1980
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Arlington, Texas 76011 ya -
Reference:
Docket No. 50-285 j
-1
Dear Mr. Seyfrit:
In response to IE Bulletin 80-04, the Omaha Public Power District submitted a letter to the Comission, dated May 8,1980, providing a comitment and schedule for the analyses required by the bulletin.
In accordance with our May 8, 1980, letter, the analysis for reactivity increase from a Main Steam Line Break with continued feedwater is attached.
Sinc,erely, j i f'\\
Vgpv W. C. Jones Division Manager Production Operations WCJ/KJM/BJH/TLP:jm Attach.
cc:
U. S. Nuclear Regulatory Comission Office of Inspection and Enforcement Division of Operation Inspection Washington, D. C.
20555 LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N. W.
Washington, D. C.
20036 8006360 f g
ATTACHMENT I The Steam Line Rupture (SLB) event was analyzed for Ft. Calhoun Unit 1, Cycle 6 using reload licensing assumptions and methods except for automatic initiation of auxiliary feedwater' flow in 3 minutes from initiation of the event.
The analysis assumed that the event is initiated by a circumferential rupture of a 26-inch (inside diameter) steam line at the steam generator main steam line nozzle. This break size is the most limiting, since it causes the greatest rate of temperature reduction in the reactor core region. The break outside containment cases were not analyzed since these result in less adverse reactivity transient because the break size is smaller due to flow venturis in each steam line.
For conservatism no credit was taken for the existence of the venturi flow restrictors in the analyses.
The SLB event was analyzed with the assumption of a three minute delay between the time of transient initiation and time when Auxiliary Feedwater (AFW) flow is delivered to the affected steam generator.
This is conservative with respect to the expected time of AFW initiation since the generation of the AFW signal actually occurs at the time of the low steam generator water level trip signal, and AFW flow is initiated three minutes following this signal.
The analysis assumes, therefore, that AFW flow is delivered to the steam generator sooner than the flow is actually available resulting in a conservative prediction of the resulting cooldown.
A conservatively high value of the AFW flow was calculated assuming that all auxiliary feedwater pumps are operable. An AFW flow of 10.5% of full power feedwater flow was used in the analysis. This value accounts for pump run-out due to reduced back pressure.
In addition, the analysis conservatively assumed that all the AFW flow is fed only to the damaged steam generator.
The analysis conservatively assumed that there is no main feedwater isolation when the Containment Isolation Actuat',c Signal (CIAS) is actuated.
- Hence, the main feedwater flow is ramped down to 5% of full power feedwater flow in 60 seconds (A more realistic main feedwater flow would be stopped in 20 seconds). This assumpt, ion is conservative because it prolongs the cooldown of the RCS and thus results in a more severe reactivity transient.
The two steam line rupture cases considered in conjunction with automatic initiation of auxiliary feedwater flow are:
- 1) 2 Loop - Full Load (1530 MWt)
- 2) 2 Loop - No Load (1 MWt)
Two Loop - Full Load The Two Loop - 1530 (includes 2% power measurement uncertainties) MWt case was initiated at the conditions listed in Table 1.
The. Moderator Temperature Coefficient (MTC) of reactivity assumed in the analysis corresponds to end of life, since this MTC results in the greatest positive reactivity change
during the RCS cooldown caused by the Stean Line Rupture.
Since the reactivity change associated with moderator feedbach varies significantly over the moderator temperatures covered in the analysis, a curve of reactivity insertion versus temperature, rather than a single value of MTC, is assumed
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in the analysis.
The mnderator cooldown curve assumed in given in Figure 1.
The noderator cooldown curve given in Figure 1 was conservatively calculated assuming that on reactor scram, the highest worth Control Element Assembly is stuck in the fully withdrawn position.
The reactivity defect associated with fuel temperature decreases is also based on end of life Doppler defect.
The Doppler defect based on an end of life Fuel Temperature Coefficient (FTC), in conjunction with the decreasing fuel tenperatures, causes, the greatest positive reactivity insertion during the Steam Line Rupture event.
The uncertainty on the FTC assumed in the analysis is given in Table 1.
The 8 fraction assumed is the maximum absolute value including uncertainties for end of life conditions.
This too is conservative since it maximizes the subcritical multiplication and thus, enhances the potential for Return-To-Power (R-T-P).
The minimum CEA worth assumed te be available for shutdown at the t'1e of reactor trip at the maximun allowed power level is 5.81%a0, assuming that the most reactive CEA is stuck in the fully withdrawn position during a scran.
The analysis conservatively assumed that on Safety Injection Actuation Signal or.e High Pressure Safety Injection Pump and one Low Pressure Safety Injection pump fail to start. A conservative value for the boron reactivity worth of -1.0%Ao per 87 PPM was assumed in the analysis.
In addition, no credit is taken for any baron injected via charging pumps taking suction from the Boric Acid makeup tanks.
The conservative assumptions on feedwater flow were discussed previously.
The feedwater flow and enthalpy as a function of tine are presented in Figures 2 and 3 respectively.
Table 2 presents the sequence of events for the full power case initiated at the conditions given in Table 1.
of time is presented in Figure 4.
The reactivity insertion as a function event is given in Figures 5 through 9.The response of the HSSS during this The results of the analysis show the affected steam generator blows dry at 69.1 seconds and thus terminates the initial cooldown of the RCS.
The peak reactivity attained prior to delivery of Auxiliary Feedwater Flow is
.097%Ap which occurs at 72.6 seconds.
The corresponding peak Return-To-Power due to subcritical multiplication attained prior to delivery of auxiliary feedwater flow is 14% at 73.1 seconds.
The delivery of boron via the High Pressure Safety Injection pump inserts negative reactivity and the core power decreases to the decay power level as the reactivity becomes nore negative.
The delivery of auxiliary feedwater flow starting at 180.0 seconds initiates a further cocidown of the RCS which results in more positive reactivity -
insertion and causes the core to approach criticclity. However, the addition of boron via the high and low pressure safety injection pumps (one HPSI and one LPSI_.. pump are assumed operable) terminates the approach to criticality and the core remains subcritical.
The peak total = reactivity attained following AFU is
.32';An at 418.2 seconds.
Since the core never 64
reaches criticality, and, in the absence of suberitical multiplication, there is no Return-To-Power.
The results of the analysis show that the Re. turn-To-Power prior to AFil is less than the Return-To-Power for the 2 Loop-Full power case presented in the FSAR.
Since the critical heat flux was not exceeded in the FSAR case, it is concluded that the critical heat flux is not exceeded for the present case.
Hence, the consequences of tne event with AFW ara nn mnre advorse than the 2 loop full power tiSLB case presented in the FSAR, without AFW.
a Tuo Loop- !!o load The two loop-no load case was initiated at the conditions given in Table 3.
The moderator cooldown curve is given in Figure 1.
The cool-down curve corresponds to an end of life MTC.
An end of life FTC was also used for the reasons previously discussed in connection with the two loop - 1530 tiWt case.
The minimum CEA shutdown worth available is conservatively assumed to be 4.2'soo, assuming the most reactive CEA is stuck in the fully with frawn position during a scram. A maximum inverse boron worth of 87 PPfi/'..co was co:.servatively assumed for the safety injection during the no load The feedwater flow and the enthalpy used in the analysis are presented case.
in Figures 10 and 11 respectively.
Table 4 presents the sequence of events for the 2 loop-HZP case initiated from the conditions given in Table 3.
The reactivity insertion as a function of time is presented in Figure 12.
The flSSS response during this event is given in Figures 13 to 17.
The results of the analysis show that the affected steam generator blows dry at 121.7 seconds.
The peak reactivity attained durin;; this time period of.27';ao.
The addition of boron from the high pressure safety injection adds negative reactivity and thus the core reactiv' y becomes more negative.
At 180 seconds 'the auxiliary feedwater ' low is deliver'ed to the affected steam generator.
This' initiates a further cooldown of the RCS.
The cooldown of the RCS inserts more positive reactivity.
However, Low Pressure Safety Injection flow is initiated at 102.6 seconds which injects additional boron.
The negative reactivity added due to boron injection via the LPSI's more than offsets the positive reactivity inserted by the added cooldown of the RCS.
Hence, the core never reaches criticality af ter initiation of auxiliary feedwater flow.
The two loop-no load case attains a peab total reactivity of.27%an which occurs before AFW.
Hence, the results of 2 loop-no load SLB event with autonatic initiation of auxiliary feedwater are no worse than the 2 loop-no load case analyzed for the FSAR, without AFLI.
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TABLE 1
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s KEY PARAMETERS ASSU!1ED Ill THE 11AIrl STEAM LINE BREAK EVEtlT WITH AUT0t1ATIC IllITIATIO:1 0F AUXILIARY FEEDWATER (2 LOOP
' FULL LOAD C0flDITI0fl)
-Present Analysis Parameters Units Values Values t
Initial Core Power Level MWt 1420 1530 Initial Core Inlet Temperature 'F 547 547 Initial RCS Pressure psia 2100.0 2175.0 Initial Stean Generator psia 770,0 880.5 Pressure Low Steam Pressure Trip psia 478.0 478.0 Se.tpoint Safety Injection Actuation psia 1578 l'578 Setpoint High Pressure Safety Injection psia not available 1390.0 Flow Delivery Low Pressure Safety Injection psia 20l*
201 Flow Delivery CEA Worth at Trip
%Ap
-5.0
-5.81 lioderator Cooldown Curve
%ap vs. 'F Figure 1 Figure 1 Doppler Multiplier 1.20 1.15 Inverse Baron Worth PPfi/%ep 80.0 87.0 Feedwater Flow BTU /sec vs. Sec Figure 2 Figure 2 Feedwater Enthalpy BTU /lbm vs. Sec Figure 3 Figure 3 t
fio credit for Low Pressure Safety Injection was taken in the FSAR analysis.
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_ TABLE 2
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Sequence of Events for the Main Steam Line Break Event with Automatic Initiation of Auxiliary Feedwater Flow (Full Load, Two-Loop Condition, Nozzle Break)
Time (sec.)
Event Safety System Initiated Setpoint or Value 0.0 Initiation of break j
3.5 Low steam generator Reactor Protection System 478 psia Pressure trip signal Main Steam Isolation System occurs, MSIS initiated and Main Steam Isolation Valves begin to close.
4.4 Trip breakers open 6.9 CEAs at 9G; Insertion Reactor Protection Sys tem 8.4 Complete clo3ure of Mair Steam Isolation Valves to terminate blowdown from the intact steam generator 13.8 Low RCS pressure, SIAS Safety Injection System 1578 psia Initiated 14.5 Pressurizer empties 21.8 High Pressure Safety Safety Injection System 1390 psia Injection flow Initiated 64.4 Main feedwater flow
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completes ramp down to 5%
69.1 Affected steam generator liquid inventory depleted and beginning of blowdown of feedwater only I
73.1 Peak return-to-power
- 3 4,~
occurs with a peak reactivity of.097"no return-to-power includes decay heat and subcritical multiplication I
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TABLE 2 (Continued)
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l Time (sec.)
Event Safety System Initiated Setnoint or Value 135.4 Boron. from safety injection reaches core mid-plane 180.0 Auxiliary Feedwater flow to affected stean generator initiated 418.2 Low Pressure Safety Safety Injection Systen 201 psia flow initiated i
418.2 Peak reactivity post
..32%Ao auxiliary feedwater
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delivery I
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9 e
4
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TABLE 3_
KEY PARAfETERS ASSU!iED Ill THE 11AItl STEAf1 LIllE BREAK EVEllT llITH AUTOiATIC IllITIATION OF AUXILIARY FEED;lATER (2 LOOP - NO LOAD C0flDITION)
FSAR Present Analysis Parameters Units Values Values Intial Core Power Level Milt 1
1 Initial Core Inlet Temperature F 532.0 532.0 Initial RCS Pressure psia 2100.0 2175.0 Initial Steam Generator psia 900.0**
900.0 Pressure Low Steam Pressure Trip psia 478.0 478.0
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Setpoint Safety Injection Actuation psia 1578 1578 Setpoint High Pressure Safety Injection psia not available 1390.0 Flow Delivery Low Pressure Safety Injection psia 201*
201 Flow Delivery CEA llorth at Trip
%ap
-2.4
-4.2 Moderator Cooldown Curve,
%ao vs. 'F Figure 1 Figure 1, Doppler tiultiplier 1.20 1.15 Inverse Boron llorth PPM /%ap 80.0 87.0 Feedwater Flow BTU /sec vs. Sec Figurei0 Figure 10 Feedwater Enthalpy BTU /lbm vs. Sec Figure 11 Figure 11 4
110 credit for Low Pressure Safety Injection was taken in the FSAR analysis.
approxir. ate value e
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' TABLE 4
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Sequence of Events for the flain Steam Line Break Event with Automatic Initiation of Auxiliary Feedwater Flow (No Load, Two, Loop Condition,flozzle Break)
Time (sec.)
- Event, Safety System Initiated Setpoint or Value 0.0 Initiation of break
{
1 3.9 Low stean Generator Reactor Protection System 478 psia Pressure trip signal f4ain Steam Isolation Systen occurs, MSIS initiated and flain Steam Isolation Valves begin to close.
4.8 Trip breakers open 7.3 CEAs at 905 Insertion Reactor Protection System 8.4 Complete closure of flain Steam Isolation Valves to terminate blowdown from the intact steam generator 10.4 Pressurizer empties 13.4 Low RCS pressure, SIAS Safety Injection System 1578 psia.
Initiated 21.4 High Pressure Safety Safety Injection System 1390 psia Injection flow Initiated 102.6 Low Pressure Safety Safety Injection System 201 psia Injection Flow Initiated e
i 115.0 Doron fron sa'fety injection ----
reaches mid-plane 121.7 Affected steam generator liquid inventory depleted and beginning of blowdown of feedwater only 124.5 Peak Reactivity
.27%As
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TABLE 4 (Continued) l
' iime (sec.)
Event Safety System Initiated Setpoint or Value 180.0 Auxiliary Feedwater-
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