ML19340C226
| ML19340C226 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/16/1980 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| IEB-80-18, NUDOCS 8011140271 | |
| Download: ML19340C226 (14) | |
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TENNESSEE VALLEY AUTHORITY-; :
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cHATUNOOG A. TENNESSEE 37401 400 Chestnut Street Tower II October 46',
1980
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Mr. James P. O'Reilly, Director Office of Inspection and Enforcement i
U.S. Nuclear Regulatory Coc=ission Region II - Suite 3100 101 Marietta Street Atlanta, Georgia 30303
Dear Mr. O'Reilly:
0FFICE OF INSPE ION-Ay'D ENFORCEIGNT BULLETIN 80 NRC-0IE REGION II j
LETTEE RII:JP 50-327 /SEQUOYAH NUCLEAR PLANT UNIT 1 - RESPONSE TO BULLETIN Enclosed is our complete response to your letter dated July 24, 1980, which trans=itted IE Bulletin 80-18 on Adequate Minitum Flow Through Centrifigal Charging Pumps. A partial response to the bulletin was subcitted on September 22, 1980. The enclosed response incorporates all of the informa-tion transmitted by our Septe=ber 22, 1980, letter.
i TVA employees have expended approximately 55 tanhours conducting the review and preparing the reports required by this bulletin. An additional 15 man-hours are expected to be expended to complete the required modifications.
If you have any questions, please get in touch with D. L. Lambert at FTS 857-2581.
Very truly yours,
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TENNESSEE VALLEY AUTHORITY L. M. Mills, Manager Nuclear Regulation and Safety Enclosure cc:
Mr. Victor Stello, Director (Enclosure)
Office of Inspection and Enforcement U.S. Nuclear Regulatory Co==ission Washington, DC 20555 i
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ENCLOSURE SE;UOYAH NUCLEAR PLANT UNIT 1 RESPONSE TO IE BULLETIN 80-18 ADEQUATE MINI!CM FLOW TO CE!.TRIFUGAL CHARGING PUMPS
-t Response to lten 1 of the Bulletin TVA has completed calculations to determine if the Sequoyah Nuclear Plant unit 1 charging system would maintain adequate pump flow during parallel safety injection operation and determined that adequate flow would not be taintained. The detailed calculations outlined by the Westinghouse Electric Corporation letter (NS-TMA-2245) are included as Attachment 1.
t Response to Itec 2 of the Bulletin a.
Modifications are planned for Sequoyat unit 1 as described under Interim Modification I of the Westinghouse letter attached to the bulletin. These codifications include:
(1) Verifying that the CCP ciniflow return is aligred directly to the CCP suction during normal operation with t.w alternate return path to the volume control tank isolated (locked closed).
(2) Removing the safety injection initiation auto =atic closure signal from the CCP miniflow isolation valves.
(3) Modifying plant emergency operating procedures to instruct the operator to:
(a) Close the CCP miniflow isolation valves when the actual RCS pressure drops to the calculated pressure for manual reactor 3
coolant pump trip.
(b) Reopen the CCP miniflow isolation valves should the wide range RCS pressure subsequently rise to greater than 2,000 psig.
These modifications are expected to be complete by November 15, 1980. In view of the startup test schedule, TVA does not believe this schedule for codifications has any significant safety implications.
b.
As indicated in the Westinghouse Electric Corporation safety evaluation (Attachment 2), if manual operator action is taken to close the CCP miniflow valves when the RCS pressure drops to the calculated pressure for manual reactor coolant pump trip (1,500 psig), no significant change in peak clad temperature (PCT) would be observed. 'Since tripping of the reactor coolant pumps is itself a manual operator i
action, it is our opinion that the additional requirement of closing the CCP miniflow valves (two handswitches) will not burden the operator and can be accomplished in the time necessary.
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The CCP tiniflow valves are supplied with shutdown power via the diesel
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generators. The site post-accident monitoring instrumentation (powereo by batteries and/or diesel generators) used to determine the reactor coolant punp trip pressure will be utilized to determine the need for opening or closing the CCP tiniflow valves, d.
As indicated in the Westinghouse safety evaluation, the flow available from the CCP's with the codification in place, along with the operator action indicated in item 2.b above, will have a negligible effect on the safety-related analysis (note Attachment 3 for UHI plants).
Since the results of the safety-related analyses evaluated in item 2.d e.
indicate the insignificant effects of the interi= codification and pro-cedure change, all technical specifications based on these remain valid.
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.4TTACHIENT 1 SEQUOYAH NUCLEAR PLANT UNIT 1
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MINIMUM CENTRIFUGAL CHARGING PUMP FLO'J DURING T40 FUMP PARALLEL SAFETY INJECTION CALCULATION FCR NRC IE BULLETIN NO. 80-18 i
Purpose Check capability to provide minimum pump flow during parallel safety injection with two centrifugal charging pumps (CCP's).
References 1.
I 2.
Letter from T. M. Anders6n, Westinghouse Water Reactor Division, to V. Stello, NRC, dated May 8,1980, No. NS-TMA-2245.
k 3.
Sequoyah Nuclear Plant Unit 1 Preoperational Test WG.lc data.
j Calculations I
Following the format suggested in Reference 2, using data from Reference 3.
Step 1: Maximum developed head pump flow = 2,600 psid = 6,006 ft. @
73.1 gpm -(pump 1A-1A) 4 Minimum developed head pump flow = 2,470 psid = 5,705.7 ft. @
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72.3 gpm (pump ' lB-1B)
Step 2:
Correction for testing error.
Test gauge accuracy =.25% x 3,000 psig = 7.5 psi (17.25 ft.)
+ 10 psi (23 ft.) reading accuracy = 40.25 ft.
I Maximum pump = 6,046.25 ft. @ 73.1 gpm Minimum pump = 5,665.45 ft. @ 72.3 gpm Step 3: From construction of pump flow curves, attached, minimum pump =
I 5,670 ft. @ 60 gpm Projection of weak-pump head point on strong pump operating curve
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shows flow of 224 gpm.
Total flow from both CCP's guaranteeing 60 gpm to tie weak pump is 224 gpm + 60 gpm = 284 gpm Step 4: Determination of injection piping head loss.
From Reference 3, runout head of pump 1A-1A = 480 psi runout flow of pump 1A-1A = 490 gpm l
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Develooed Head ph 1104 ft.
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g, (Runout Flow Rate)'
Q' (490 gpm)'
4.6 x 10
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The resistance of the injection piping (thf) at the total CCP flow required to maintain 60 gp= through the weak pump is:
Lhf = K02 = (4. 6 x 10- ft/gpm) (284 gpm) 370.86 ft.
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Step 5:
RCS head loss for 4-loop plant - 50 psid (116 f t.)
Step 6:
Determining elevational head loss RWST elevation 739' - 5 3/4" CCP suction elevation 672' - 11" RCS cold leg injection nozzle elevation '
697' - 1 13/16" Pressurizer safety valve elevation 757' - 2 3/16" RWST to CCP suction 66.56' Minus CCP suction to RCS
-24.23' Minus RCS to pressurizer S.V.
(60.03 ft. assuming a full pressurizer)
(Corrected for density dif ference)
-43.30'
- 0.97' Step 7:
Calculation of pressurizer safety valve pressure Note: 1% setting tolerance
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Relief pressurizer = 2,485 psig + 25 psig = 2,510 psig (5,798 ft.)
Step 8:
Determination of maximum RCS pressurizer pressure at which 60 gpm minimum flow is naintained to weak CCP.
Ibximus RCS pressurizer = (CCP developed head @ total CCP flow) -
(injection piping head loss) - (Head loss through RCS) -
(elevation head loss)
Maximum RCS pressurizer = 5,665.45 - 370.86 - 116
.97 =
5,177 ft. = 2,241.5 psig C nclusions 2
Comparing the maxi =um RCS pressurizer = 2,241.5 psig with the safety valve relief pressurizer = 2,510 psig, it is evident that the 60 gpm flow required for the weak CCP will not be maintained.
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WESTI::cnoUSE ELECTP.IC CORT 0 PATIO:: SAFETY EVALUATION.
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CEtiTRIFUGAL CHAEGIliG PUMP OPERAT10!i FOLLOWIliG SECO?iDARY SIDE HIGH EI?ERGY LI!iE RUPTURE I
Reference 1:
Westinghouse Letter 1:S-TMA-2245, 5/8/80
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Reference i notified the fiRC of a concern for ccnsequential damage of ene or more centrifugal charging pumps (CCP) following a secondary system high energy line rupture.
Reference 1 included a calculational method and sample calculation to permit evaluation of this concern on a plant
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specific basis.
Should a plant specific problem be identified, Westinghouse provided several reco=endations for the interim until necessary desi n 5
modifications can be implemented to resolve the problem.
These reco=enda-tions inci'uded two ' proposed interim modifications which included:
h.
Remove the safety injection initiation automatic closure signal from the CCP miniflow isolation valves.
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2.
P.cdify plant emergency operating procedures to instruct the operator to:
a.
Close the CCP miniflow isolation. valves when the actual RCS pressure drops to the calculated pressure for manual reactor coolant puqp trip.
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b.
Reopen the CCP miniflow isolaticn valves should the wide range RCS pressure subsequently rise to greater than 2000 psig.,
c.
Prior to making this reco=endation, Westinghouse evaluated the impact' of the recommended operating procedure modifications on the results of the various accidents which initiate safety injection and are sensit'ive to CCP flow delivery.
The accidents evaluated in detail include secondary system ruptures and the spectrum of small loss of coolant accidents.
The analytical results for steam generator tube' rupture and large loss ~of coolant accident are ilot sensitive t'o a reduction in CCP flow pf ^ the magnitude that results
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This letter functions to supplement Reference 1 and identify the sensitivity,of the accident analyses to the recomended modifications.
T,his evaluation,is generic in nature.
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t Secondary System Ruoture Sensitivity analyses have been performed for seccndary high energy line
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ruptures to evaluate the impact of reduced safety injection flow due to normally open miniflow isolation valves.
These analyses indicate an insignificant effect on the plant transient response.
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iEE A.
Feedline Ruptt.rt Folicwing a feedline rupture, the reactor coolant pressure will, reach the pressurizer safety valve sctpoint within approximately 100 seconds assuming maximum safeguards with the power-operated relief valves inoperable.
With minimum safeguards, the reactor coolant pressure will not reach the pressurizer safety valve setpoint until approximately 300 seconds. The time that the reactor coolant system pressure remain's
- b at the pressurizer safety valve setpoint is a function of the auxiliary feedwater flow injected into the non-faulted sisam generators and the timeatwhichtheoperatorisassumedtotakeactiEn. With the mini-
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flow isolation valves open,'the peak reactor coolant system pressure and the water discharged via the pressurizer safety valves are insignifi-cantly changed from the FSAR results..
i B.
Steamline Rupture E=F The effects of maintaining the miniflow isolation valves in a 'normally
, open position was also investigated following a main steamline rupture.
For'the condi' ion II " credible" steamline rupture, tht results of the
'.:5 transient with the miniflow valves open showed that the' licensing
[ criterion (no return to criticality after reactor trip) continues to
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,b_e me t.
The condition ~ III and IV main steamline ruptures were also reanalyzed' assuming the miniflow valves were open. The results of the analysis showed that, even with reduced safety injection flow
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.ip.to t,he core, no DNB occurred for any rupture.
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Small loss of Coolant Accidents d
Sensitivity analyses have been performed to evaluat'e the impact of reduced safety injecticn flow on small break loss of coolant accidents (LOCAs).
These analyses indicated that miniflow isolation can be delayed, but it 2
must occur at.sc=e time into the small break LOCA transient in order to
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limit the' peak clad temperature (PCT) penalt'y.
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The proposed modification delays miniflow isolation and reduces 51 flow delivered by approximately 45 gpm at 1250 psia during the delay time pericd.
The impact of this modification was evaluated based on two isolation times:
- 1) The tim'e equivalent to the RCP trip time, and 2) approximately 10 minutes in the transient, or just prior to system drain to the break for the worst The second time was evaluated to determine the impact small break sizes.
if the operator does not isolate miniflow within the proposed prescribed The spectrum of small break sizes are considered to encompass all
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time.
possible small break scenarios.
Only cold leg break locations are considired since they will continue to be 1,imiting in terms of PCT.
Very small breaks that do not drain the RCS or uncover the core, and A.
maintain RCS pressure above secondary pressure (< s2" diameter).
I For these break sizes, it is quite possible that the operator may never isolate the miniflow line, since the pressure setpoint will ',
not be reached, and continued pumped SI degradation will persist.
.f However, this'will have no adverse consequences'in terms of core i
uncovery and PCT.
No core uncovery will be expected for the degraded The SI case, similarly to the base comparison case with ful.1 SI.
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only effect would be a slightly lower equilibration pressure for a
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given break' size.
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Gill breaks that drain the RCS and resul.t in the maximum cladding
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m temperatures (2"< diameter <6")..
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l This range of break sizes represents the worst small break size for 1
' At tachment 2 most plants as determined utilizing the currently approved October 1975 Evaluation Model version, as sho..n in KCAP-E970-P-A.
.If miNiflow is isolated at the RCP trip setpoint rtther than the "S" signal, a reduc-tion in safety injection flow of less than 45 gpm results, averaged for the approximately 50 second period of time separating the two events.
This reduction in RCS liquid inventory 'results in core uncovery less than ene second earlier, and has a negligible impact on PCT.
If mini- -
ficw is isolated at the. time of core uncovery, or approximately 10
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minutes for break sizes in this range, a greater reduction in RCS liquid inventory results in a core uncovery 10 seconds earlier in the transients resulting in less than a 10 F PCT penalty for the worst size small break.
This w'ould not' result in any present FSAR small break analysis becoming limiting than the corresponding large break LOCA FSAR analysis.
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If minificw isolation does not occur at any time into the transient-for this category of small LOCA, a PCT penalty of 200*F or more could occur.
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C.
Small break sizes larger than the worst break through the intermediate.
break sizes (> 6" diac.oter)'.
Break sizes in this range have been determined to be non-limiting for small break utilizing the currently approved Octeer 1975 Evaluation Model, WCAP-8970-P-A.
If miniflow isolation' occurs at the RCP trip time for these break sizes, the negligible effect at PCT presented.
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above also applies.
Similarly, if isolation occurs prior to core uncovery, the~ small (< 10*F) PCT penalty will resu3 ras well.
- However, for these larger break sizes, the time of first core uncovery occurs prior to 10 minutes.
If miniflow isolation is not grformed until
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1,0 minutes, reduced SI will be delivered during the care uncovery time, which can have a greater impact on PCT.
Studies in5cate a potential PCT penalty of 40*F resulting for these non-limiting break sizes if miniflow'is not isolated until 10 minutes.
This is not expected to l.:g=
shift the worst break size to larger, breaks, since etese breaks are l
typica-11y hundreds of degrees'l..s then_ snaller liaii.ing.smil bred
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analyzed with the currently approved Evaluati.on Mo:irit.
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' Attachment 2
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For all F5AR saall LOCA analyses, one ccmplete train failure is. assumed.
It l'
is clear that two charging pumps without minificw isolation provides more flow than ene pump with miniflow isolation.
The impact presented in this evaluation maintains the one train failure and assumes no miniflow isola-tien for the remaining pump.
If both pumps were operating, the PCT results would be much icwer than present FSAR calculaticns even if miniflow isola-tion is not assumed to occur for the two pump case.
In this situation, the plant FSAR small break calculations remain conservative.
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These sensitivity stu6ies form the basis for the recommended interim modifications to the emergency operating procedures.
The accidents evalu.
ated are relatively insensitive to the recomended modifications.
- Further, the accidents evaluated will give results that satisfy acceptance criteria".
as.long as the CCP miniflow is' isolated within 10 minutes of event initiation.
However, small LOCA sensitivity studies with one SI train operating confirm
.that small LOCA analyses require miniflow isolation within 10 minutes.
To comply with the reco=en'ded modifications, the operator can isolate mini-flow at any point in the depressurization transient prior to RCS pressure reaching the RCP trip setpoint., Should a repressurization transient occur, E;.:.
the operator can open CCP miniflow at any point between the RCP trip set-point and 2000 psig.
Such operator actions will ensure that plant accidents
- atisfy acceptance criteria and protect the CCPs from consequential damage y,
during the repressurization transient that accompanies a secondary system N
h'igh energy line rupture at high initial power levels.
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i k'ESTI::GHOUSE ELECTRIC CORPOR/iTIOi! SAFETY EVALUATION
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CENTRIFUGAL CHARGIliG pVP.P OPERATION FOLLOW!! G SECCdDARY SIDE HIGH EliERGY LIllE B.REAK (UHI PLAfiT SUPPLEM 11 The small loss of coolant accident (LOCA) section of the main report was generated primarily for plant applications which do not include upper
' head injection (UHI) as part of the ECCS design. This supplement pro-vides additional small LOCA information for UHI plants and, together s
with the main report, assesses the impact of delayed miniflow isolation for small LOCAs for UHI plants.
T he model utilized to determine the SI sensitivities andlo identify the worst small break size discussed in the main report was the October 1975 Model (WCAP-6970-P-A) version of the Evaluation Model.
This model is not yet approved for UHI plant analyses.
UHI small break analyses are performed with the December 1974 small break version.
However, sensi-h tivity studies performed to determine the effect of pumped SI on small break LOCA PCTs utilizing the December model yielded nearly identical 5 results as presented in the main report. This is expected since the model changes included in the October model do not affect the basi-vessel inventory and core boiloff relationships that determine the impact of changes in pumped safety injection on PCT.
An important difference in UHI plant small break analysis results as compared to similar non,UHI piant analysis results is the small break-2::::.:.
r size resulting. in the highest PCT.
This break size is generally greater for UHI plants than for similar non-UHI plants because of the additional safety injection flow provided by the UHI accumulator at relatively high
. RCS pressures.
The worst small break size for UHI pTiants may be a
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six inch diameter break or larger.
The main report identified breaks
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of this size and larger as non-limiting small break sizes. ' While this is true for non-UHI plants, it is not accurate for typical UHI plant small
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break analyses. Therefo.re, the stated 40 F potentiall pe'nalty for
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six inch breaks applies to the worst break for UHI plants for. tne case n'
where miniflow isolation is delayed untii 10 minutes.
It is Westinghouse's opinion, however, that the stated? penalty of 40*F is conservatively high and bounding for UHI plants, for the following 5
reasons: a) The 40 F penalty was based on sensitivity studies performed assuming an approximate 20's reduction in total HPI flow.
However, the anticipated 20'; reduction actually applies only to the charging pumps.
F Intermediate head SI pumps are not affected.
Therefore, total HPI for plants with intermediate head SI pumps, which includes'all UHI plants, will result.in less total degradation, and thus a smaller PCT penalty.
The high pressure accumulator on UHI plants has a similar effect of reducing' the total HPI degradation due to the delay in miniflow isolation.
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b) The UHI accumulator is a significant source of liquid mass inventory for breaks greater than or equal to six inches in diameter.
This addi-tional mass delays the core uncovery time as ' compared to the same size break occuring on a similar non-UHI plant, since more liquid mass must
)
exit.from the break prior to core uncovery.
The delay in core uncovery" results in clad heatup at a 1.ower power level caused 'by the decay in residual core heat.
Therefore, clad heatup rates are slower which"also tends to reduce the sensitivity to changes in HPI delivery rate.
1 In conclusion, the sensitivity provided for six inch diameter and larger break sizes in the main report represents the worst break size range for UHI plants. The stated 40*F PCT penalty for breaks of this size ',
"yst resultant from a 10 minute delay in miniflow isolatio.n is a conservatively highandboundiIigvaluefortiHIplants,forthereasonsstatedabove.
If miniflow is isolated at the time of RCP trip, the negligible impact 1...,._
on PCT discussed in the main report applies for UHI plants as well.
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The <10*F penalty resultant if miniflow isolation occurs prior to core uncovery also applies to UHI plants, with the added benefit that this
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event occurs later in a UHI plant transient than for a non-UHI plant
' transient of the same break s.ize, allowing more time for t.he operator g.g
- N to act.
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