ML20214K877
| ML20214K877 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 11/24/1986 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19292G318 | List: |
| References | |
| TGX-003, TGX-3, NUDOCS 8612020496 | |
| Download: ML20214K877 (18) | |
Text
.. *,
ATTACHMENT 4 ST-HL AE # 815 PAGE I 0F / l, T6X-003 WESTINGHOUSE CLASS 3 NON-PROPRIETARY Responses to NRC Questions on RTD Bypass Elimination f
l 8612020496 861124 PDR ADOCK 05000490 A
-- _ ~.
d TACHMENT 2 i
HL AE. Ill8 AGE 9 OF 16 QUESTIONS ON STP RTD BYPASS ELIMINATION REMOVAL
REFERENCES:
(1) Letter from M.R.Wisenburg, HL&P, to V.Noonan, NRC, 10/16/86 (2) Letter.from M.R.Wisenburg, HL&P, to V.Noonan, NRC, 12/04/85 (3) Letter from M.R.Wisenburg, HL&P, to G.Knighton, NRC, S/2/85 4
The RS Branch has reviewed the above references from a thermal hydraulics viewpoint in regard to RTD bypass removal. Ref erence 3 is mainly concerned with the QDPS microprocessor based system which has the ability to average the hot leg RTD signals.
Reference 2 provided information on the out-of-range check performed by the QDPS and how it can compensate for an invalid sensor input. Reference 1 is a large package with only a fraction of the information pertaining to RTD removal.
It includes a number of plots showing transient times f or accident conditions.
The following questions are to gain further clarification and documentation related to j
I 1.
Because reference 1 contains only a fraction of its pages related to j
the RTD manifold removal we inf ormally asked f or a listing of the l
pertinent pages.
These were given as Pages:
15.0-20 1
15.2-20,01,& O2 15.4-7, 37, & 08 15.6-2, 24 Figures l
15.2-1 thru 8 1
15.4-4 thru 12 15.6-1 thru 3 l
j Please confirm!
i j
2.The RTD bypass removal has an affect on the accuracy of the hot leg i
temperature reading.
The hot leg temperature reading has the predominant i
Gffect when flow measurement uncertainty is analyzed.
Does STP plan to l
reduce the flow measurement uncertainty below 3.5*/. in the Technical Epecifications?
If so, a flow measurement uncertainty analysis is j
required.
We need to have the analysis well in advance (2 months) of when j
the draft Technical Specifications are reviewed.
Please provide the date l
when we will receive the analysis for review if you plan to use flow l
measurement uncertainty.
3.
From past experience it has been found that temperature gradients of i
7-9 degrees F can exist through the cross section of the hot leg.
The j
colution for i mproving non-uniform (temperature streaming) hot leg temperature accuracy was to use a Resistance Temperature Device (RTD) bypass system to obtain a representative sample of hot leg fluid and
,l measure its temperature.
Your new design known as RTD Bypass i
Modification, has many advantages including reduced radiation exposure, j
improved availability and reduced maintenance.
However, there is also an increase in response time from 6.0 to 6.5 seconds (Table 15.0-4) as a l
disadvantage.
Because this new method uses a single point measurement at locations where a scoop sampled over a length, it appears that some O
1 1
i 9.,-.4v-~.-.,ym.~.y.m.-wv..y.,.-.m_
-,mm.
m
.m.
,.- r,
.-,,.,--.m,,r..
ATTACHMENT 6L ST HL AE-18f 8 PAGE 3 OF IL tcmperature accuracy is lost.
Please provide information on the relative tcmperature measuring accuracy of the two methods.
If the accuracy is rCduced, indicate the amount and its affect on accident and transient i
cnalysis.
.4.
Because of the increase in response time of the new system a number of other times have been adjusted in the items shown in Table 15.2-1 (pages 15.2-20 to 22).
Please explain the adjustments made for these changes.
Are they purely from the results of the reanalysis with old settings or have settings been modified?
If settings have been modified, explain what they are.
i i
5.
The same request for the above question is requested for Table 15.4-1 (pages 15.4-37 & 38) and Table 15.6-1 (page 15.6-24).
1 i
6.
On pages 15.4-7 and 15.6-2, changes are shown with a value of +4.7 i
dcgrees F uncertainty for reactor coolant average temperature and -34 psi uncertainty for reactor coolant pressure.
Are these changes due.to the RTD Bypass Removal?
If so, what were the previous values so that a comparison can be made of the effect?
7.
You have presented accident analysis plots for before and after the l
consideration of the change in response time of the RTD temperature.
These were given on.the following Figures I
15.2-1 thru 15.2-8 l
15.4-4 thru 15.4-12 l
15.6-1 thru 15.6-3 1
]
The DNBR is shown to be well above 1.00 for all cases except in Figure 15.4-6 for uncontrolled rod withdrawal from full power.
The original value appeared to be 1.4 and the new value appears close to 1.30.
Because of the crude scale, it cannot be determined accurately if the value is cctually at, above or possibly slightly below 1.30.
If it is at or elightly above 1.00, another consideration would be the measurement uncertainties.
Please indicate the DNBR value and explain if measurement uncertainties were included.
Since the new method for the hot leg i
temperature seems to possibly have higher uncertainties, explain this affect on the analysis.
Also highlight any differences in the analyses in f
the above figures that have significant changes that should be made known j
when compared to the old analyses.
J i
i l
i i
. se ATTACHMENT L Ill{
ST.HL-A E 0F If3 PAGE 5 NRC QUESTION #1 Because reference 1 contains only a fraction of its pages related to the RTD manifold removal ese informally asked for a listing of the pertinent pages.
These esere given as Pages:
15.0-20 15.2-20.21.h 22 15.4-7, 37, b 38 15.d-2, 24 Figures:
15.2-1 thru 8 15.4-4 thru 12 15.d-1 thru 3 Please confirm!
RESPONSE TO NRC QUESTION #1 Tho cbove pages are confirmed.
Additionally, pages 15.0-6, 15.2-5 and 15.2-6 should be included on the list.
.--.-___.._.-..___-._-_____-,_..-_~-_,m..
,-..--,,._,,.,___-_-,,,m--,_
ATTACHMENT d.
ST HL AE I&l6 PAGE $ OF 14, NRC QUESTION #2
.The RTD bypass removal has an affect on the accuracy of the hot leg temperature reading.
The hot leg temperature reading has the predominant effect when flew measurement uncertainty is analyzed.
Does STP plan to reduce the flew measurement uncertainty below 3.5% in the Technical Specifications'r If se, a flew measurement uncertainty analysis is required.
Ne need to have the analysis well in a&ance (2 months) of when the W aft Technical Specifications are reviewed.
Please provide the date when we will receive the analysis for review if you plan to use flew measurement uncertainty.
RESPONSE
Tho South Texas Project does intend to provide justification f or reduction cf the flow measurement uncertainty below 3.5%.
The methodology which is boing used to provide this justification is the same as that used for Shearon Harris Unit 1.
This methodology has been submitted on the Shearon H;rris Docket as WCAP-11168 Rev.1 (Proprietary) and WCAP-11169 Rev.1 (Non-i proprietary), entitled, RCS Flow Uncertaint.y For Chearon Harris Unit l",
l October, 1986.
The South Texas plant specific submittal will reference l
this methodology and provide pertinent plant specific summary tables f or l
NRC review.
This submittal is anticipated to occur on December 15, 1986 i
cnd will support a flow measurement uncertainty of 2.3%.
i
-,----,-----,...,y_...,_,------mm--r,-,-r-,---
l PAGE b 0F l SOUTH TEXAS PROJECT (YMPARTETON OF rwn aun tsv ler fir: 11MPERAMIRE MEA 51RD(ENT METIEI)S The new method of measuring hot les temperatures, with the therscwell REs used in place of the three scoops, has been analyzed to be slightly.more effective than the RTD bypass system, since thd streaming error oeused by imbalances in the scoop sample flows is eliminated. Although the new method measures tamperatures at one point, at the thernowell tip, compared to the five sample points in a 5-inch span of the sooop measurement, the thernovell tip [
).
The thernowell i
measurement say have a small error relative to the scoop measu ament if the temperature gradient over the 5-inch scoop span [
). +b,c.e Temperature streaming test data has shown that [
),+b,c.e Since three RE measurements are averaged, and the nonlinearities at each scoop
[
i I
J.* ' " # Considering that flow imbalances in the three scoop branch lines could introduce temperature measurement unoertainties of up to [
'), ' '* it has been concluded that the three thernowells will provide a more accurate measurement than the three sooops. The total temperature streaming uncertainty applied to the hot leg temperature measurement with thermowells has been established at [
]* ' # or the scoop measurement.
f
\\
ATTACHMENT A ST HL AE-1812
_PAGE_7 0FlL, RESPONSES TO NRC QUESTIONS 4-7 SOUTH TEX AS F'ROJECT HTD BYPASS ELIMINATION SAFETY EVAL UATION i
RESPONSE TIME The impact of the RTD bypass, elimination on the FSAR Chapter 15 non-LOCA cafety analyses is the increased response time associated with the fact response therrr. owe 11 RTD system.
Currently, the overall response time of the South Te:<as RTD bypass nystem assumed in the saf ety analy'ses is 6.0 secondt.
Fcr the f es.t recponse thertr:owell RT D syst err., the overall channel responne time will be 6.5 seccnds.
Thih iocrcaccd channel response time recultw in 1onger del hym ir om the time when the fluid conditions in the reactor coolant system (RCS) recuire en Over temper atere cel ta-T cr Ovt rpower ciel t as-T reccter trip until a trip
{
eignol it actually generat.ed.
Thieref ore, these transicents the.t rely or.
4 the above mentioried trapu must be evalucted for the 1onger responsc tiee.
Th ec affected tt.v*.s i e n t s i nc l uc.c the U1 controlled F:CC6 Withdrawal at Power, the Lc+u of.ond,Yurbine Trip, the Intovertent Cpening of a I'r css uri::er Ghicty cr Rt lie f Valve, the Ur.cor.trollod Doron Dilution at Power, and the Stcan line Ruotorr at F'ower evento and ere dir.cutr.ed in the 4cllowant:
parcarepnc, i
RTD UNCUC NN1 Y Allownnces are madce in the FLAR analycot, for t h r*
initig.1 average RCS I
temperaturo, pressure and powcr (an described in FEAfi 115. u ).
An a result of the RTD bypat.u el itni nat i ori, the f cllowing control anri protection syr.ttm paremntors have been offectod (by the change f r cm or.tr het leg RTD to three hot Icg RTDn):
the O <crt tempc'rcture dul to-T (UTDT), Overpoe.or delta-T (OPDT), and Lou RCS Flow reacter tr ip f unctions, Low Tavg and Low-Low Tavg control functions, RC C everage temperatur e mrracuremente use d f or input to the rod control syctom, and the measuroment uncerta*.nty for NCS flow, i
ATTACHMENT 7-ST HL-ISI PAGE OF l System uncertainty calculations have been performed for these parameters to determine the impact of the design change.
The results verify that f.cufficient allowence has been made in the reactor protection system setpoints to account for the increated RTD error.
Therefore, the current values of the nominal setpoints noted above as defined by the South Texas Project Technical Specifications remain valid.
However, the t ornperttur e c.ver aging system impl emented at South Texas for the averaging of the hot Icq temperatures adversely affected the RCS overagt-temper atur e m:?asurement input to the rod control system.
The affected value is the initial RC3 average temperature allcwance, which was b
increated by O.7"F from itL previous value of
- 4. 0=F, end has been appropriately noted in the South Texat FSAR.
J For the Chapte.- 15 non-LCCA eventt uncficcted by the increased channel responce time on the OTD1 and OPDT reactor trip functions. due to the RTD byptsc climinat4cn. the cifect of the incrcased initial RCS avercge temperature ciro-allowence has betii ascertained via coparato evaluations.
In (11 i ntlancc c.
the c.onclutionc presented in the South Te:ta. FS AR rom,*: r valid under this error allavance assumption.
i NCN-LO" A TR." N ~.,l ElrTS RE ANnL N ZED 1
All the rvm.tt-reatlyred uct d thc LO!' IRAN c.omuuter code.
LOFTRAN is a digital computer code, developec' to simulato t:anc)ont behavicr in a mul tiloop protour i::rM water reactor s y t.1 e m.
'I he pr ogr itm claulctet the noutron Linctico, the thermel-hydraulic conditions, the pretnuri:cr, steam j
generato c, reaitor coolant pumps, and the control and pr otect ion cyst em operation.
The secondary side of each *. team generator utili:cc a homogencout satur ated minture f or the thermal tr anni or't c.
Uncontr011ed RCCA !)enl W1thdrawal of Power i
The Uncontrolled RCCA Dani. Withdrawal at Power-Event is doccribed in Section 15.4.:' of the FSAR.
An uncontrolled RCCA bahl. withdrawal at power causes a poultivo reoctivity insortion which results in an increano in tht' l
core heat f l u...
Since the steam gennrat or lagt bchind tht-core cover generation, there is a net incretace in the reactor ceclant t er.ptrature.
ATTACHM 4T JL-
~.
ST HL g JI PAGE Y 16 Unlect tern.inated by manual or automatic action, the increase in coolant temperature and power could result in DNB.
For this event, the Power Range High Neutron Flux and Overtemperature delta-T reactor trips are assumed to provid2 protection against DNE.
Therefore, thic event has been 2
reanalyred with the increased channel response time to show that the DNB limit is mot.
Methode The escumptiores used are consistent with the FSAR in that initial power, pressuro, and RCS average temperature are assumed to be at the nominal valuco ccrrespondirig to 10",
60%, and 100% power plus the appropriate revised uncertainty allowancec.
Doth minimum and maxirnum reactivity f eedback cases were reanalyzed with the increated ti rt.e recponce vc,li..c.
The anal ysis wac donc using the LOFTPAN computer ccdc.
Re s u l_ij:
F or both r.n r.i m in. and ma;,1 mum reactivity i nser ti or.c, et tne vor ious power 1evelL ar.21y ed. the D!lbR 1imit 1 c n.et.
A calculetcd sequence of (vents for a fact ud L1on intertion red.e for ectb activity f cedbac k actiutt.pti or, i t. p.-ener.ted on the revised FSA9 Tablo 15.4.1 for fu11 cowcr.
Rev:. sed FSAR Figurcc 1 D.1 -- 4 through.10.4-9 chow r esJ1ts f or a fcct inocrtien caw and n slow i ncer t. l on caso ccer cupondir.; to the 100% powc caso cr.d both reattivity ascumntions.
The plots of mi n i rnum CNBR vor s.ut, reactivity intertio-ri.to et (.1 1 three power lcvels are shown in revised FS W Fi gur et 15.4-10 through 15.4-12.
Conclusionc The lirtit DNDR is tatt, and thrrofere, thc conclucient pretented in the
}
FSAN remain valid.
ATTACHMENT.2 ST HL AE 1818 PAGE / OOF &
Qus of Lodd/ Turbine Trio The' South Texas FSAR only explicitly analy:en the Turbine Trip Event which is presented in Section 15.2.3.
This event relies on any of three reactor trips for~ primary protection:
High Pressuri:or Pressure, Low-Low Steam Gencrator Water Level, and Overtemper eture delta-T.
Thus, the increase in f
channel retponce time may have nn effect on the results of this transient.
Me t h oj;'s.
The escumptionc used are conca stent with the FSAR in that initial power, pretnure, and KCU average tarrperature arc assumed to be at the notainal vclues c.c'er ec sondina to 100% power plus the appropriato revised uncertainty allowtncec.
All i out-casus pensunted in the FSAR were -
reanalyicd i nc or por at i ng t he ctsun.ptions of the RTD bypasc elimination.
These are Decinning-of-Life ( E;OL) cnd End-of-Life (EOL), w.th and without pr et:sur e contr ol (p r essur i n er cprev erd POrNs).
The aqel ysit wac don?
using the LOrTRAN corr.puter code.
Mhl.il_.t_q For c)1 coat.i nM i ens of reactivity feedbock and pressure control, the DNER limit is met.
lhe cecultn cf thoto four caten are presented in revi sed FSAF: Figures 15. 2-1 ttrovoh 10.2-0.
A cal:ulnted seqaence c;f evente as shewn in rovi ur d FT.R Tot. l c 15.2-1.
Revised Figures 15.2-1 and 15.2 -2 show t he r ecpor tet. 4 or a turbine tr3p evant with n i ni.T um r cactiv) ty feedbaci (Low) scsuming operability of pressurtror sprays and PORVr..
The reactor is tripped by the High Pr ecsurizer Pressure trip f uncti on.
The DNDR increancs t%roughout the trancient and never drepas below the init:a1 value.
The pr eucur i. er safety valves are actuated and primary system pressure ren.ainc below tha 110% de :;i gn value.
Revised Figures 15. T-0. ar.d 15.2-4 show the respcinses f or a tu:-bine trip with maximum reactivity feccbsch (EUL) and pressure control, The reactor is tripped by the Over temperature delta-T ti-1p iunctien and the DNDR never drops below the initial valut.
The prossurt:er safety valve lift set pressurr is not reached.
i
ATTACHMENT F ST HL-AE f f M PAGE // OF 16 Reviced Figurec 15.2-5 and 15.2-6 show the responses for a turbine trip with minimum reactivity feedback (BOL) and without pressure control.
The reactor is tripped by the High Pressuri:er Pressure trip function and the DNDR never drops below the initial value.
The pressuri:er safety valves are actuated and insintain system pressure below 110% of the design value.
Revised Figurec 15.2-7 and 15.?-B show the responsen for a turbine trip with rnanimurr. reactivity f eedbach (EOL) and without prescure control.
The reactor is tripped by the High Pressure Pressure trip function and the DNDR never drops belew the initial val ue.
The pressuri er safety valves are actutsted and me.intain systein pressure below 110% of the design value.
Conelutaonc
]
The DNLP 1ir.il val ue 2s met in al1 4our cason, end thereiore, the conclu:1ont, presented in tha FSriR remain va.1id.
int d c Ltp_it__Qpenin_ _of a Prencuriper SM c t v_or Rel,l e4 - V.nLvpe i
t The Iriadvertent Opr_ning of a Prescuriner Saf e t y or Rolivi V41ve Cvona. 1c dec:ribec in Sc tion 15.6.1 of t ht: Gouth Tc::tc FSTR.
At cecidental depr es suri:ati on ci the RCE could occur a t, t.
r esul t of cn i nac'vtr t oni cponinc; of t p r ew.ir i z er relief or usfety velve.
Si mco et s..f e t y v 4.l s ;.e in sitt.d to r r li eve opproni matel y twice t% ctuam flowrate of a reliuf val vt:
and will therefore allow a much more rcpid deprescur2:nttor upon openin;;,
the met.t s.ever e core conditions, res.ulti ng f rom an accidental dep r et.t.ur -
1:ation of the RCE are acsociated with in inadvortent oper.ing cf a pressurirer saiety velve.
This event rol1eb ore either of two reactor trips for prir.ar y ;)rotection:
Law Precuuri;;:er Pret. cur e and Overtem-peretture deltec 1.
lhuc, the increase ir, channel r e c p o n c e t i rt.t. vany have or offect on the renults of this trannient.
Met hodg.
The arr.,s.umpti ons used are corici stent with the ISAR i t.
that i n i t i c.) power, preucure, and RCS average temperature are as.cumed to be at tt--e nomi nal values, correnoor ding to 100% pc wo-olut the aop: opr a ct r reviced uncertainty allowancec.
The analysiti m.n dann uct riu U.e LOF il W, coriputer code.
ATTACHMENT W 1
ST.HL AE ItlIb PAGE/5kOF ECEP)t5 For the inadvertent opening of a pressurizer safety or relief valve, the DNDR limit is met.
A calculated sequence of events in shown in revised FSAR Table 15.6-1.
Revised FSAR Figures 15.6-1 through 15.6-3 present the results of this transient with the increased channel response time.
Gpnc 1 us i onc,
}
The DNDR limit value ic, met, and therefore, the conclusions presented in the FSAR romain valid.
2_ign et Power Ungontro!)ed Daron D 1.1 t
The Ucrcn C11ution at Fowcr Event in descr ibed in Section 15.4.6 of the FSAR.
The sequence of ovents fer this transient, whor. in manual contiol, is eccent i all y identical to tt.c Uncont rolled RCCA Dani: Withdenwol at Power Event.
The bor on dilution at powne tranciert has boon analy::cd based upon the r esults cd the uncontrolled RCCA banl withdrawal at power analys,1c incl udi ng the increaced channel response ti m<2 and the increased temper o t u-t-unc ertainty allowarico.
'I t.c r esul te o f thic onelynis thow that the conclurians precentud in the revised FGAR Section iti.4.6 remain valid, g
i.e., there is greater then 15 minutet eve ilcbl e f rom the tims of al ar m until the total lots of plant chutdown margin.
6.kSMJ _1 ne Ryrd,rej at P,py_s_r.
The Stoan.line Ruptur e at Power transient wac encly:cd consistent with WCAP-9226-R1.
The analynin included the increased chenqcl responce time cnd the increar.ed tempor ature uncerteinty allowance.
For thic ovent, the het bec'n met.
} dont gn bact s as ciencribed in WCAP-9226-H1
ATTACHMEN A ST.HL AE /
PAGELS OF Gonclucions The impact of the RTD bypass climination for South Tenas Project Units 1 cnd 2 on the FSAR Chapter 15 non-LOCA accident analyses has been ovaluated.
For the events impacted by the increase in the channel it has been demons + rated that the conclusiont presented in 7
response time, the FSAR remain valid.
For the remaining Chapter 15 non-LOCA events, the offect of the increased initial RCS average temperature error allowance has been ascertained via separate evaluations.
In all instances, the conclusions, presented in the South Texas FSAR remain valid under this arror allowance assumption.
A---
- - - - ~
~
ATTACHMENT A ST.HL AE-lill PAGE14 OF lL, Ouestion 4 8ecause of the increase in response time of the new system a number of other times have been adjusted in the items shown in Table 15.2-1 (pages 15.2-20 to 22).
Please explain the adjustments made for these changes. Are they purely from the results of the reanalysis with old settings or have settings been modified? If settings have been modified, explain what they are.
Question 5 The same request for the above question is requested for Table 15.4-1 (pages 15.4-37 & 38) and Table 15.6-1 (page 15.6-24).
Resnonse to Questions 4 and 5 This increased channel response time results in longer delays from the time when the fluid conditions in the reactor coolant system (RCS) require an Overtemperature delta-T or Overpower delta-T reactor trip until a trip signal is actually generated.
Therefore, those transients that rely on the above mentioned trips must be evaluated for the longer response time.
Tt.e affected transients include the Uncontrolled RCCA Withdrawal at Power, the Loss of Load / Turbine Trip, the Inadvertent Opening of a Pressurizer Safety or Relief Valve, the Uncontrolled Boron Dilution at Power, and the Steamline Rupture at Power events.
System uncertainty calculations have been performed for these parameters to determine the impact uf the design change.
The results verify that sufficient allowance has been made in the reactor protection system setpoints to account for the increased RTO error.
Therefore, the current values of the nominal setpoints noted above as defined by the South Texas Project Technical Specifications remain valid.
i All the above events analyzed in support of the RTO bypass elimination include revised error bands for initial RCS average temperature and initial pressurizer pressure (increased error in pressurizer pressure is not as a result of RTO bypass elimination). Therefore, not all events experience trip signals affected by the Overtemperature delta-T response time.
1 l
l l
l 1845n:39/80L/ll-85
ATTACHMENT 2, ST HL AE I&ll PAGE 150F LL Question 6 On pages 15.4-7 and 15.6-2, changes are shown with a value of +4.7 degrees F uncertainty for reactor coolant average temperature and -34 psi uncertainty for reactor coolant pressure. Are these changes due to the RTD Bypass Removal? If so, what were the previous values so that a comparison can be made of the effect?
Response to Question 6 The temperature averaging system implemented at South Texas for the averaging of the hot leg temperatures adversely af fected the RCS average temperature measurement input to the rod control system.
The affected value is the initial RCS average temperature allowance, which was increased by 0.7'F from its previous value of 4.0*F, and has been appropriately noted in the South Texas FSAR.
The increased error in initial pressurizer pressure (from 130 psi to 134 psi) is not as a result of RTO bypass elimination.
1 i
1845n:40/80L/11-85
ATTACHMEN P--
ST HL AE-if PAGE 160FL42 Question 7 You have presented accident analysis plots for before and after the consideration of the change in response time of the RTD temperature. These were given on the following Figures:
15.2-1 thru 15.2-8 15.4-4 thru 15.4-12 15.6-1 thru 15.6-3 The DNBR is shown to be well above 1.30 for all cases except in Figure 15.4-6 for uncontrolled rod withdrawal from full power. The original value appeared to be 1.4 and the new value appears close to 1.30.
Because of the crude scale, it cannot be determined accurately if the value is actually at, above j
or possibly slightly below 1.30.
If it is at or slightly above 1.30, another consideration would be the measurement uncertainties.
Please indicate the DNBR value and explain if measurement uncertainties were included. Since the new method for the hot leg temperature seems to possibly have higher uncertainties, explain this affect on the analysis. Also highlight any differences in the analyses in the above figures that have significant changes that should be made known when compared to the old analyses.
3 Response to Question 7 1
The impact of the RTO bypass elimination for South Texas Project Units 1 and 2 on the FSAR Chapter 15 non-LOCA accident analyses has been evaluated.
For the events impacted by the increase in the channel response time, it has been demonstrated that the conclusions presented in the FSAR remain valid.
For the remaining Chapter 15 non-LOCA, events, the effect of the increased initial RCS average temperature error allowance has been ascertained via separate evaluations.
In all instances, the conclusions presented in the South Texas FSAR remain valid under this error allowance assumption and the DNBR limit value is met.
I j
l l
1845n:41/8DL/11-85
-~ ___.
ATT ACHMEhT3 ST-HL AE 416 STP FSAR PAGEI OFA Plant characteristics and initial conditions are further discussed in Section 15.0.3.
Except as discussed above, normal reactor control systems and ESF's are not required to function. Several cases are presented in which pressurizer spray and power-operated relief valves are assumed, but the more limiting cases where these functions are not assumed are also presented.
The RTS may be required to function following a turbine trip.
Pressurizer safety valves and/or steam generator safety valves may be required to open to maintain system pressures below allowable limits. No single active failure will prevent operation of any system required to function. A discussion of ATWT considerations is presented in Reference 15.2-2.
Results The transient responses for a turbine trip from 102 percent of full power operation are shown for four cases:
two cases for minimum moderator feedback and two cases for maximum moderator feedback (Figures 15.2-1 through 15.2-8).
For the minimum moderator feedback cases, the core has the least negative moderator coefficient of reactivity. For the maximum moderator feedback cases, the moderator temperature coefficient has its highest absolute value.
The calculated sequence of events for the accident is shown in Table 15.2-1.
Q prtMe'&& phe<.
Figures 15.2-1 and 15.2-2 show the transient responses for the turbine trip with minimum moderator feedback, assuming full credit for the pressurizer spray and pressurizer power-operated relief valves.
o credit is taken for the turbine bypass. The reactor is tripped.hy7thej ::rt. F rat u r M :.r 4 l 18' 43 signal. The minimum departure from nucleate boiling ratio (DNBR) remains well above 1.30.
The pressurizer safety valves are actuated and maintain primary l18 system pressure below 110 percent of the design value. The steam generator safety valves limit the secondary steam conditions to saturation at the safety valve setpoint.
43 Figures 15.2-3 and 15.2-4 show the responses for the turbine trip with maximum moderator feedback. All other plant parameters are the same as the abovegetcept @
The DNBR increases throughout the transient and never drops below its initial value. Pressurizer relief valves and steam generator safety valves prevent overpressurization in primary and secondary systems, respectively. The pres-surizer safety valves are not actuated for this case.
The turbine trip accident was also studied assuming the plant to be initially operating at 102 percent of full power with no credit taken for the pressuriz-i er spray, pressurizer power-operated relief valves, or turbine bypass. The 43 l
reactor is tripped on the high pressurizer pressure signal.
Figures 15.2-5 l
and 15.2-6 show the transients with minimum moderator feedback.
The neutron flux remains essentially constant at 102 percent of full power until the reec-tor is tripped. The DNBR increases throughout the transient.
In thfs case the pressurizer safety valves are actunted and maintain systen pressure below 110 percent of the design value.
Figures 15.2-7 and 15.2-8 show the transients with maximum moderator feedback with the other assumptions being the same as in the preceding case. Again, the DNBR increases throughout the transient, and the pressurizer safety valves are actuated to limit primary pressure.
@ &rd fN. rtAder b kph h h. cot.dtAifadwet OT Nif %
+
15.2-6 Amendment 43
ATTACHMENT
~
STP FSAR ST HL AE 181 PAGEgLOF.t TABLE 15.2-1 TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Accident Event Time (sec)
Turbine Trip 1.
With pressurizer Turbine trip, 0.0 l43 pressure control loss of main (minimum moderator feedwater flow feedback)
F.8 l18 C a g :::::::
^L M
Ih D#'#
tri? E^irt 522;h d
"~'
h CtOY peta Rods begin to y,8-h9 18 h,p p. m t cesc. k d drop
" Initiation of
-0;s.
18 70 steam release l-from steam gen-erator safety I, valves Minimum DNBR (1) occurs 9V Peak pressurizer
-4:6-I pressure occurs l
2.
With pressurizer Turbine trip, 0.0 pressure control loss of main
'3
+
(maximum moderator feedwater flow feedback)
(.. $
Overtemperature AT
- h e-18 trip point reached Rods begin to 6.3 -he l18 drop Initiation of 70470-l18 t
steam release
~
from steam gen-erator safety valves 15.2-20 Amendment 43 L