ML17276A957
| ML17276A957 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/11/1982 |
| From: | Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| GO2-82-26, NUDOCS 8201260202 | |
| Download: ML17276A957 (142) | |
Text
REGULIAITORY IVFOR>RATION DISTRIBUTIOVr SYSTEM(
(RIOS)'CCESSION'VSR'2912'60202 DOC', DATE: 82'/01/11 VO T'PRIZED e NO DOCI(EiTI IIi FACILI:50 397 HPPSS'ucil carr.,P'r orj act r Uni ti 2R Ha'shi ngtop>> Publ i c>> Pbwe 05000397 AUTH ~ NARKI AUTHOR= AFF>>ILIIATION BOUCHE>>Y>>G'.D. 'a'shingtop>>
Publiic Power" Supoly, System RKCIP; NAME" RKCIIPIIEN>>TI AFF>>ILIIATI OiV SCHHENCERR A ~.
Llicensing Br ahch 2.
p~
Fbrwa>>rds resoopses to! remaining Reactor I Sys Branch Quest(one" 211 ~,129R2'14,.031r 2'14.'.148' 211.209.ReVi Sed>>
FSAR OageS re ODYN cocle wi lilt be incor oorateQ= into> Ame'nd>> 23 ~.
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Washington Public Power Supply System P.O. Box968 3000George Washington Way Richland, Washington 99352 (509) 372-5000 January ll, 1982 G02-82-26 SS-L-02-CDT-82-008 Docket No. 50-397 Mr. A. Schwencer, Director Licensing Branch No, 2
Division of Licensing U.S. Nuclear Regulatory Commission Washington D.C.
20555 HECeveg 8
'~s 8 s >Beg~
~ l@ttt taame C406g~
tlag Dear Mr. Schwencert
Subject:
Reference:
NUCLEAR PROJECT NO.
2 RESPONSES TO REACTOR SYSTEMS BRANCH QUESTIONS AND ODYN ANALYSIS WNP-2 FSAR REWRITE
- Letter, R. L. Tedesco to R.
L. Ferguson, "WNP-2 FSAR-Request for Additional Information", dated June 8, 1981.
Enclosed are sixty (60) copies of responses to the remaining Reactor Systems Branch questions.
The responses to questions 211.129 and 211.136 are new.
The responses to 211.031 and 211.148 are rewrites of responses previously submitted to the NRC.
The response to 211.209 was also submitted as a part of the LRG appendix (RSB-3).
These responses will be incorporated into the FSAR in Amendment 23.
Also, enclosed are sixty (60) copies of the draft revised FSAR pages as a result. of the ODYN analysis.
These r'evised pages will also be incorporated into Amendment 23 to the WNP-2 FSAR.
Very truly yours, G.
. Bouchey Deputy Director, Safety and Security CDT/ct Enclosures cc:
R. Auluck NRC WS Chin BPA R. Feil
- NRC-Site 82pi2~p~
pgppp397 2-Sgpiii
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8 February 1980 Paae 1 o 4
Q. 211.031 (5.4 7)
In Table 5.4-3 of the
a signal indicat'ng reactor low wa"er level.
It appears tnat you have mislabelled these valves in this table as "recircula ion line suction" rathe" han as "RHR isolation".
Indicate whether this valve isolation siaral is based on the same signal as the RHR pump actua-'ion in the low pressure coolan-injection system (LPCI) moce ('.e.,
a water level wrich 's 1.0 oo above the active core).
If no-', indicate he wa"e" level in the reactor pressure vessel at which the isolation signal is genera"ed, tnereby isola"ing the RHR suc-
"ior. va3.ves.
Show the.
cool-'ng of the rea'ctor core can be maintained assuming
- a. pipe b=eak outside the containment.
Assuming a pipe break outside conta'nmen in the RHR system when the plant is in a, shutdown cool'ng mode, provide "he
. ollowing additional information:
a
~
Identify the sys" ems available or maintaining core cooling.
c
~
Indicate the maximum discharge rate resulting from ne postulated break and the time interval available for recovery based on the discharge rate and i"s ez ect on core cooling.
Iden ify the alarms available to alert the opera-tor in the even" of sucn a break and snow that sufficient time is available for operator action to prevent damage to safety-related systems.
d
~
e.
Indicate wha recovery procedures are available.
Following
- a. postulated break in a moderate energy line, the single za'ure criterion snould be applied in the manner discussed
'n Section 3.6.1 of the S" andard Review Plan (SRP)
HURFG-75/087, and in Branch Tecnnica'osi"ion APCSB 3-1, "Protection Aaainst Postulated Piping Failures In Fluid Systems Outside Containment,",
November 24, 1975.
Response
n
~
~
~
8201260202 211
~ 03$ -1
hh
Insert to Page 211.031-1:
Table 5.43 has been revised to indicate that the RHR isolation valves NOF008 and NOF009 are the "Shutdown Cooling Suction" valves.
F008 and F009 isolate at reactor water Level 3 which is 174 inches above the too of the active fue l.
LPCI is initiated at reactor water Level 1
(Reference, FSAR Figure 5.2-6).
The following items respond directly to the items requested above:
a.
should a pi pe
NHP-2 AMEND!4EHT HO.
8 February 1980 Page 2 'of 4
ailure occur, outside the containment, in the RHR system when the plant is in shutdown cooling, acceptable core cooling would be achieved by the core cooling systems.
The following core cooling systems would be available to maintain core cooling when applying SRP
- 3. 6.1 and ~ APCSB 3-1:
If the single active failure is HPCS the
~ >>:
ADIOS If the single'active failure is LPCS the
'1 ':;.
+.,
RDK.
If the single active failure is LPCI (not shut-down cooling loop) the following are available:
HPCS+
LPCS +
1 LPCl W PQS The maximum discharge resulting from the largest i
crack in the RHR piping outside containment is determined using the guidelines in BTP HEB 3-1 for moderate energy piping.
The maximum dis-charg e rate is es timated to be 1000 g pm
( to be conf irmed later
'n the ongoing pipe break and missile study).
This is based upon a pipe break in the pump discharge piping (18" Schedule
- 30) at the pump discharge
- flange, normal wate'r level in he reactor during shutdown cooling (approxi-mately 50 inches below the steam line nozzles),
reactor pressure of 135 psig, and the RHR pump, running at 7450 gpm (normal shutdown flow).
See "c" below for. the time int rval available for recovery.
The f I. ow rate used in t e
a a
e analysis referenced in part a) of this response was 1443 gpm.
T e fo owing alarms are avax a
e c
e opera-tor in the event of a pipe break in the shutdown cooling line outside containment.
l.
Low reactor water level alarm (Level 4, 198. 7 inches above active fuel).
2.
Low reactor water level (Level 3) to scram and isolate HOF008 and NOF009.
3.
Equipment area high temperature (Class lE) to isolate HOF008 and MOF009.
4.
High flow rate in the shutdown cooling suc-tion line to isolate tiOF008 and i~lOF009 (Class 1E).
211. 031-2
Insert to
. age 211'.031-2:.
A speciaL analysis for LaSaLLe was made of a hypothesized crack in the RHR suction Line outside o-primary containrent during operation in th shutdown cooling mode.
This analysis was performed with the standard GE LOCA models.
For this event the realistic or actual system condit'.ons are as follows.
No high pres restorationr RCi Cs but t h the start of psia and the decay heat i
4 hcurs have sure systems are ava i.e.r no feedwaterr e reactor water Leve this event.
Vessel YSiYs are c losed at s approxir ately 1/
o elapsed subsecuent ilable no CRO L is at pressu the st rated tc reac for water inventory fl owi no HPCSr and no normal elevation at re is Less than 150 art of this event.
The powers i.e.r approximately tor scram cr shutdown.
For a cons=rvative solution
~ o this hypo.hetical events the follow-ing sequence of events and conditions were assumed to exist or ensue from the hypothesized crack in the suction Line.
Crack occurs in the RHR Line; water'evel decreases to reactor vessel Level 3; then the RHR isolation commences and is comple ed 4:0 seconds Later.
b.
Systen pressure rises as a result of the isolation to where the vessel pressure reaches the SRV setpoirt thus causing them to opens blowdowni and reclose.
c.
inventory depletion results from blocdown and from Leakage out cf the cracked Line.
d.
The operator manuaLLy actuates ADS to reduce vessel pressure to where the Low prcssure ECCSs can replenish the water inventory.
e.
Mater Level is restcred to within normaL Limits to protect the core rom over.temperature.
ResuLts are presented in Figures 211.031-1 through 211.031-4 for a bounding calcu'.ation o
this event.
The standard Appendix K assumptions were used along with these conservative initiaL concitions:
a.
The ti-,.ing index "as started at the RHR isoLation (when Level 3
was attained) to neglect the time for the Level to fall from normal water leveL to Level 3 (about 2 minutes).
b.
An ini"ial pressure of 1055 psia was assumed to neglect the pressure rise
~ ime from the 150 psia (pressure permissive for shutdown cooling) uoon completion of RHR isolation to the 1055 pressure at.ainment.
This results in increased mass Loss during the 40 second isolation period due to greater driving pressure.
it also decreases the time increment needed for pressure
.o attain the relief valve setpoint.
Insert to Page 211.031-2 (Continued):
c The analysis assumes that scran occurs coincident with the start of the timing instead of 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> earlier.
This assumption naximizes the peak c Lad'emperature and steam production during the transient thus driving more fLuid from the vessel and prolonging the blowdown phase.
d.
Only one LPCS and one LPCI Loop were assumed to be availabLe throughout the event.
Operator action does not include possible diversionof the other two LPCI Loops fron the RHR mode.
e.
The crack area used in the analysis is defined consistently with the NEB 3-1 guidance for cr'ack size.
This crack area is consistent with FSAR postulates.
Results from this conservative ana lysis who that more than 20 ainu.es are available for the operator to depressurize the vesseL.
Once the system pressure is beLow the LPCI or LPCS shutoff head~
the reactor water level is restored to normal Limits very rapidly.
The naximum clad temperature is much less than the arbitrary 2200 F Linitat ion.
b.
The RHR system is a
Low pressure systems and alL of the piping outside of the prinary cooLant pressure boundary is classified as "moderate energy" piping ands according to BTP NEB 3-1i only
~
cracks (i.e.r not breaks) are considered in moderate energy piping.
Reactor vessel pressure must be decreased to below 135 psig before the RHR system can be connected to the reactor vesseL.
NNP-2 ANENDNENT NO 10 July 1980 Page 3 of 4
5.
Reactor bu'=inc floor drain sump level for leakage rate grea"er than 50 gp-:.
6.
Reactor building floor drain leakage rate alarm (5 gpm).
7.
ECCS pum room flood level instrumentation (Class 1E), installed to detect passive failures in the ECCS post-LOCA (Reference response to
"-SAR Question 212.003).
'Iotw'hstanding hese
- alarms, however, only about 13,000 gallons o= water will spill out of the
- break, before the reactor water level drops from the normal shutdown cooling level to Level 3, auto-matically closing j40r"0GS and l'1OF009 and isolating the break.
1'o single active failure can prevent isolation of the break.
-Bm '.
+~WE-eeR~~C S - n&RHC w-i-l-lne&aebxneL-iea~
l PG ~i'~~~A~.=i&+ SeVQ~
~~o d'or the largest pipe break (1000 gpm), which can only occur in the RHR A or B pump rooms, the flood level resulting from 13,000 gallons will not affect opera" ion of either RHR pump A or B.
In addition, the "looding would only affect the room in which the p'pe break occurred because of the watertight in"egrity of the RHR A and B pump rooms.
- herefore, no operator action is required to protect these pumps.
Pip breaks
'n tne RHR shutdown cooling mode which can affect other ECCS systems (LPCS, HPCS or RHRC) via flooding through the floor drain piping in the u"per portions of the reactor building will have flooding rates less than 1000'pm because they will be higher in the building (less static head),
have less driving head due to friction losses, and smaller crack sizes (smaller d'ameter pipe).
These pipe breaks in the upper portion of the reactor building will be imme-diately detectec by the high flow (5 gpm) alarms in the floor drain downcomer piping.
Regardless of what the pipe break discharge rate is, the flood level resulting 'om 13,000 gallons is not capable o" affecting the operation of'he
- LPCS, 211. 031-3
NZP-2 A'l"liDYE'~7 hlO e 8
February 19i30 Page 4 of 4 HPCS or R:.RC pumps, assuming all oz the water is sp'11ed into each pump room.
- Again,
'o operator action is reauired to protect these pumps.
Zt should be no ed that the environ.ental effects (pressure, temperature and humidity) of pipe
.. breaks during shutdown cooling are being adcressed by ongoing pipe break and missile study.
t C~XAg 3.s %~GoBC~M-~u ~ &~~C ouipment remains unc" ional to automaticaH.y heave core covered at all tines.
Qn6 cold snu down ~~ocedure, i.e.,
containment
- heat, remo-
- val, needs to-"¹ =esumed. Ice p'e break d'saoles the co.;so shutdown cooldown suction line, cold shutdown~ca e assumed by the alter-na e shutdown coo3:aown pat
'scussed in 15.2,9.
Xr the p'p~ereah
'n the shutdocccooldown line is downstream oz the F006 valve, norma3 shutdown countdown csn be resuned us'ng the redundaw~RH'R e.
=or applica ion o= single failure criteria, see "a" above.
211.031-4
~
~
~
~
12
~ssv s scswsscn nttnrttG sttUTnotctt coot. itic; (tt,rcr
+ it.res t-nns)
~VESSEL PAESSUAE ADS ACTUATED
~'rr MCD' 0
UJ CX CA C/)
Cad CL
~vl C,s
~ Vo rs'I rl r 0
s-s Cl sil nm c:.
Vl
~ Cil m
I rsl rsl ~
~
Q IM rsvp m
sl CD XI
~-)
Ut rr >
(e (A-Z g
.g r.
l c>
t 0.<<
LU O.
0.<<
0.8 TINE (sEc) x lp~
1.2
F lj 0
i
Insert to Page 211,.031-4:
- d. lf a break should'ccur in one RHR shutdown cooling Loop cu side containment during shutdowns the foLLowing action is taken upon detection.and isolation.
Tne main steam iso-L'ation valves wilL be reopened and reactor excess steam wilL blow down to the main condenser untiL the shutdown cooling process via the other RHR loop is established.
The redundant shutdown cooling Loop components are also no.
assumed to fail under the cited NRC requirements of BTP APCSP 3-1.
If the pipe crack should occur in the common manifold supplying both redundant Loopsr the isolation mechanism is the
- same, as before'ut recovery would require reversion to the alternate shutdown configuration discussed in 15.2.9.
In this configurationr vessel water is circulated from the suppression pooL through the RHR heat exchanger to the vessel with return to the suppression pool via the ADS discharge Lines.
f
'I
l ~
0 ~
~ I
~ ~
~ ~
Q.
211.129 (4
~ 6)
The standby Liquid control system and the recirculation flow control system ar e reactivity controL systems.
Address or r'eference these systems in Section 4.6 and address all require" ments of Standard Review Plan 4.6.
Response
r The recircuLati'on 'flow control system
'is evaluated against the general design criteria as follows:
a.
Criteria 20'1'3 and 25:
Criteria 20m 21m 23 and 25 are applicable to protection systems only.
The recirculation flow control system is a reactivity controL system but is not a protection system.
b.
Criterion 26:
The recircuLation flow controL system is the second reactivity control system required by this criterion.
The requirements of this criterion,do not appLy within the system itseLf.
c s Criterion 27:
The recircula
~ ion flow control system is not intended to control reactivity following an accident.
Consequentlyi this criterion does not apply.
Criterion 28:
The tr ansient analyses in Chapter 15 evaluate the consequences of rea'ctivity events involving changes in reactor coolant temperature and pressure and cold water addition.
The results of these analyses indicate that none of these postulated events causes damage to the reactor coolant pressure boundary.
in additioni the integrity of the corer its support structures and otherreactor pressure vesseL internals are maintained so that the capability to cooL the core is assured.
The.evaluations with respect to genera l design criteria of the standby liquid control system can be found in Section 9.3.5."
The first paragraph of Section 4.6 has been replaced with the following:"
"Functional design of the controL rod drive system (CRD) is discussed below.
Functional designs of the recirculation flow control systen and standby Liquid control system are described in Sections 5.4.1 and 9.3.5r respectiveLy.
- Revised draft FSAR pages attached
NNP-2 l
LZ 5.4-3
."'".=~.'D!~~NT NO.
4 uune '979 A'eu'Is@
m~~ l g,-'+ 8'-3~I M
AND VZZ>-
D" SCR PTEON>>
Ac Eve/
cation Inactive Va1 ve Description Valve No.
Reference
~s 1 cure
"=KB. Vessel Zn Act've
- ct've act3.ve
- "1 2" 042 12~1 1 1 2a 5.4-13a
- 5. 4-13a
- 5. 4-13a and h
a.nd h anc h
~":3/recir-culation Line Xn Act'e
-"2=050
~i 1 2~0 1 9 5.4-13a and h
5.4-13a anc h
Act ve Inactive
=-12" 053
=-12" 1 12 5.4-13a and h
5.4-13a and h Head Spray Act've Active
="12" 019
=-1 2:-023 5.4-13a anc h
5.4-13a anc h
,P,~R
= "g~~uJ Al Suction Active Ac"ive Inac 'e r.12= 009
- -12= OOS
- "12 113
- 5. 4-13a and h
- 5. 4-13a anc h
5.4-3.3a and h RCZC Act've Vessel Out Active Act've Act've
= 064
~
"-53 "-063 "51 076 H51"-OOS 5.4-9a anc h 5.4-9a and h 5.4'-9a and b
5.4-9a and h
(Nuclear Inactive Boiler)
Reactor Vessel Head Inactive Inactive B22" 001 B22" 002 B22= 005
- 5. 1-3c S. 1-3c
- 5. 1-3c
"-eel'a er xn Ac 've Inactive Ac ive Ac 've B22= 010 B22:-013.
B22P032 B22" 065 5.1-3c 5.1-3c 5.1-3c 5.1-3c
'Sazetv Relief Act've B222'013 5.1-3c
- In adcition to the process valves 1'stec here'n, there are instrument test connections, a ain valves anc sa~pling valves less than one 'nch nominal size within the BCPB.
Refer to
'cures in 3.2 or the'=- locat'ons.
5.4-57
~ I
~
~
- 4. 6:"U? C""F.-~
D:-SXG'.. O. ~- CT-,-,~CON~HO-,
SVSTYS
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a QT Q i
gyes i V oc c, (ve co 6 i anc
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CSS nt ol syste.-..s cons'st o=. control ro s
=~~~~a~ r~a~vi~~
o
=o he
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VS -Ck--
c Dec in i 3
~ 5 4.6.1 OB.":-. ~
OY "-OB:-"-
CP.D vsT="N 8,,6.1.1 Con"rol Ro= Drive Svste-.;. Des'gn 4 ~ 6 '
~ 1
~ 1 C
1 1
6
~
~
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Des gn Bases I
G~
'6 a
D'6 wi Ba es 1
1 1
1
~ 0
~
~
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~
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Sa=.etv Design Bases The con ro sa.ety ces ro" cr'e
~ gn Oases:
I mecnan.ca sys emi. meets
-he
=oliow' g
a
~
- e cesign p Gv ces
- o a su:.ficien lv apic con-ro ro" 'nserticni -'..at no cruel carnage re-su's
=r"mi an'.
a" nor;.ia1 operating " ansient.
- e cesign
'nc uces pos't'on'ng cev'ces, each G: which inc'ifua v supports anc positions a cont o 1 roQ ~
c.
ac pL pcs't'on'ng device:
2.
3.
p even s its contvol
<<oc Ero ini iat ng s(" thcrawal as a result o= a sirigle mal-
- unc i.ion ~
Zs
~ ndivicuall y 6..-"g zed wneri rap roc 'nsert'cin (sera.-..)
's s'g..a'ec
=ail 're o
power sources external positioning cev'e coes not preven aosxt~cnxng cevxces'ontrol rocs insertec.
xc control so that to ~ne o"her Crom being Zs "nciv"c.a.
~ lv cperated.
so that a zailure cne pcs'"'oni.".g cevice Goes riot a,==ect the op@rat'n o= any other posi ioning cevice.
<.6.1.1.1.1.2 power Generation Des'n Basis The con"rol oc svs"em c 've esign prov'ces the control ro"s to control power genieration
""c= positioning the core.
4.6-1
Tl
Functional design of the control rod drive system (CPD) is discussed below.
Functional designs of the recirculation flow control system and standby Liquid controL system are described in Sections 5.4.1 and 9.3.5r respectively.
'J
WNP-2 Q.
211.136 (4.6.2)
Identify the specific common mode failure analysis and pro-tection from common mode failures referenced in Section 15A by Sections 4.6.2.1 and 4.6.2.2r respectively.
Response
Section 15A is the Plant NucLear Safety Operational Analysis (NSOA).
This analysis provides analyticalLy determinable'inits on the consequenc s of different classific'ations of pLant events i.e.r expec ed operationaL transients; unexpected operationaL transientsr and is thus an event-consequence oriented evaluation.
Event 53 Reactor Shutdown and Cooldown Without Control Rods satisfies the requirements of Regulatory, Guide 1.70 for Sections 4.6.2.1 and 4.6.2.2 r espectively.
In Section 15Am this event is discussed on page 15.A.6-36.
In additioni the scram discharge volume system has been eva lu-ated (as requested in Question 010.041) against the criteria enumerated in the Generic Safety Evaluation report "BWR Scram Discharoe System"r dated December 1r 1980.
With incorporation of the system modification described in the response to question 010.041m the scram discharge volume system was found to be in full compliance with the SER.
0
co-...ponent in. the SLC system and the act tnat nuc'ear system coolcown takes severa'ours while 1'cuid contro'olution injection takes approximately two hours.
Since this p obabil-ity is small, considerable time 's available for repairing and estoring the SLC system to an ope able condition. while reac or operation contin res.
Assurance that 'the system w'l still fulfillits function cur'ng "epairs is obtained by demonstrat'ng ope at'on o
the operable pump.
The SLC svstem is eva'"a ec against the appl'able General Des'n Cri"eria as ollows:
Cr'ter'on 2:
The SLC svstem is located in the a ea outside of the primary coil ainment and below the refuel'ng floor.
In th'
- location, it is protected by wal's rom external natural phenomena such as earthcuakes, tornadoes, hu icanes and floods and also from the effec"s o
'nte nal postulated accident events.
Cr'te 'on 4:
Qr
~
p I
ggg ~~ig y gyv M~
+5~
/ /
gem The SLC system is react'v'ty control system
~~
do not apply to the SLC system.
+ g~A( ~
Th's criterion is nest applicable/.
See the General Design Cr'e ia Section or c'scussion of combined capability.
Criterion 27:
The S
C system is designee
. o" the expected environment in the compartment in which ='s located.
In th's compartment, it is not subject to the conc'tions postulated in th's crit-e ion such as miss'les, wh'pping p'pes, and discharging fluics e ZO> 2/) 2'3 s d ZS 7 n ZSC.
Cr-s=~~~ applicab' to protection systems only.
The SLC system 's a reactivity control system and is evaluated gaxnst rx,~er~oa 29 (see below).
<>,27 ZP o g 2g Criterion 26:
/
C/. ".~"c
- 9. 3-30
Q.
211.148 (15.0)
Resolve the following items in Table 15.0-2:
a)
Nodify the values of vessel Level trip to agree with the values specified in Figures 5.2-6 and 5
3 2 (item 29).
C b)
Specify the maximum percent relieving capacity assumed in Chapter 15 for each node of SRV actuation (items 25 and
- 26).
c)
Provide th following information concerning the high flux trip setpoint used as input to the REDY modeL (item 29):
Explain why the high flux trip setpoint should not be increased to 122%
NBR prior to multiplication by the thercal"power correction factor of 1.043 to account for the setpoint plus calibration errors instrument accuracyi and transient overshoot specified in Table 7.2-4.
2)
Explain why the thermaL-power correction factor is applied to the high flux trip setpoint used in the REDY modeL.
d)
Provide the following information concerning the APRYi thermal trip setpoint used as input to the REDY model (item 30):
1)
Specify the highest flow-rated trip setpoint to be given in the Tech'nicaL Specifications and how this va lue is obt a ined.
2)
Is the 122.03 NBR setpoint equal to the setpoint to be specified in step d (1) times the thermaL power factor of 1.043 spec ifi ed in step c (1)?
e)
Table 15.0-2 does not contain alL of the input parameters used in the REDY computer code.
For each transient and accident analyzed in Chapter 15'rovide the following:
1)
A List of all input parameters.
2)
Justification that the input parameters are conservative.
WNP-2
Response
a)
The water Level setpoints 5.3-2 are consistent with 29 of Table 15.0-2.
The.a different elevation l.evel spec i f i ed in Figures 5.2-6 and those values specified in item pparent discrepancy is due to the each table is referenced to.
Water Level Set points In Reference To E leva t ion Above Reactor Vessel Zero Figure.5.2-6 Water Level inst.rumentation zero 527.5 in.
Fioure 5.3-2 Table 15.0-2 Reactor vessel zero Bottom of steam separator skirt 0
514.'0 in.,
Howevers for the purpose of consistency and clarityi Table 15.0.-2 has been revi sed to the same reference point as Figure 5.2-6 (bottom of the steam dryer skirt).
b)
The maximum relieving capacity for each mode of SRV'actuation assumed in Chapter 15 are:
R el i Safe Thes 91106 psig 91213 psig ef valve capacities ty valve cpaacities 101.8%
NBR 111.5%
NBR 1 s 1 s e values have been added to item 22'f Ta'ble 15.02.
c) 1)
The high flux s'etpoint sh'own in TabLe 7.2-4 is incorrect and a text correction is currently underway to make Table'7.2-4. consistent with Plant Technical Specifications.
A correct list of setpoint specifications will. be found in TabLe 2.2.1.-1 of.Chapter 16 when the WNP-2 Technical Specifications are completed.
In the Lists the trip setpoint column in Table 2.2.1-1 of Chapter 16 wiLL correspond to the setpoint column in Table 7.2-4r and
'imilarly the allowable values column to the setpoint
+
instrument dri ft column.
The neutron f L'ux (run model) shown in the new L i st i s 120%,of rated p'ower that inc Ludes instrument drift.
After accounting for calibration error and instrumentation inaccuracyr this setpoint to ~ aLed 121% of rated therma l power.
Item 27 of'able 15.0-2 has been corrected as follows,:
High FLux Trip NBR Analysis Setpoint (121 x 1.043) i NBR
= 12'6.20
WNP"2 A change in this high flux trip setpoint would cause no impact on tr ansient results since for each transient
~
analyzed in Chapter 15'he reactor was tripped by the direct scram prior to the high flux setpoint being reached.
2)
The thermal power multiplier (RST) is used to give a
conservative margin that is proportional to the core powers d) 1)
The maximum f low related Specification will be 115 (Table 2.2.1-1 of Chapter by adding 2% instrument d
trip setpoint of 113.5%.
used in the FSAR analysis
<instrument inaccuracy and trip setpoint given in Technical
.5% of rated therma l power
- 16).
This value is obtained rift to the maximum nominal The sa ety Limit setpoint is 117%r whi ch also includes cat ibrat ion error.
2)
Yes'he 122.03%
NBR setpoint is caLculated from mu ltiplying the thermal power factor of 1.043 with the 117% safety Limit as discussed in d)1) above.
e) 1)
Table 15.0-.2 was selected and provided to show the principaL parameters reLated to the transient analysis.
Providing a complete Listing of inputs would be impractical.
If some particuLar area of input is of speciaL interest it can be provided upon specific request.
2)
Parameters in which variation might have significant effect on the transient result were selected conservatively to bound the design values with uncertainty allowance-The Letteri R.
H. Buchholz (GE) to'.
S.
Che September Sr 1980'Response to NRC Request ODYN Computer Nodel" r.List the input paramet parameters coupled with Table 15.0-2 should conservatism of REDY to be completed.
REDY input parameters much the same values.
Qual computer code is documented in NED0-10802.
ck (NRC) r dated for Information on er s of ODYN.
These enable the review for and ODYN have as ification of the REDY
0
NNP-2 TABLE 15. 0-2 (Continued) 17.
18.
19.
20.
21.
22.
Void Coefficient (-) 0/0 Rated Voids Nominal EOC-1 Analysis Data or Power Increase Events Analys's Data, for Power Dec=ease Events Core Ave=age Rated Vo'd P-action, Scram Reactivity,
$ hk Analysis Data Control Rod Dr've Speed, Pos'tion versus time Je" P~~~p Ratio, M
Sa:ety/Rel
~ ef Valve Capacity,
% NBR 7.48 12.70
- 7. 065 41.32 P'gure 15.0-2 Figure 15.0-2 2.41 g/ p7 23.
24.
25.
26.
27.
':anufacturer.
Quantity Installed Relief Function Delay, seconds Relief :<<unction Response, seconds Se 'oints for Safety/Relief Valves Safety =unction, psig Relief Function, psig rDe of Valve Groupings Simulated Sa ety Punct'on, No.
Relief Punct'on, No.
High Flux Trip, 8
NBR /el Analysis set point (XZA. x 1.043),
NBR Crosby 18 0.4 O.l
- 1177, 1207,
- 1091, 1121,
- 1187, 1197, 1217 1101, llll, 1131
//yg JtFj
//97~/~
/>/ 7
//a~/ ///4
//p.C, //9e h~@
Safety valve capacity Re L ie f va t ve capac i ty 111.5 101.8 111.5 101.8
- 15. 0-19
bHP-2 A!1FNDYi"NT
~HO.
11 September 1980 28.
'29.
TABLE 15.0-2 (Continued)l.i=a'I Set Point si 1071 High Pressure Scram p
g
~acmic'5 Uessel Level Trips, ~~ Above fryer
&cger~~ Skirt Bottom Level 8
(L8),
r evel 4
(L4)
Level 3
(r 3),
Level 2 (L2), f-ee~
si. <
2-.x~
y ~.
5'0.
APRYi Tnermal Trip lI Set Point, 4 NBR 6 100% core flow 122.030 31.
"=.ecirculation Pump Trip Delay/
Seconds 0.140 Pecirculation Pump Trip Inertia t
me constant for analysis, seconds*
E
~~~~c l&D -5.
gp QDY/w vcr/~a~
ctrl cgcig~i~/~a/ ~i+g~
pg J
(
cy~A 4<
- Th inertia time constant is:defined by:the expression:
2 n Jon where t g T Jo inertia time constant (Sec).
pump motor inertia
( 1 b-ft2)
~
rated pump speed (rps).
gravitational constant
( t/sec2) pump shaft torque (lb-ft).
15.0-20
1 MNP-2 211.209 (5.2.2)
(6.3)
Provi (incl repre condi ADS v
relic plan" de assurance tha" uding.testing aft sen
~ ative of an e
tions) to support alves wiLL operat f valve ope rat i on sr should be inc L your re er being xtended your as e.
A qu r inc iud uded in Lief subj t iime sumpt arit i t irg s
thi s valve ected perio ion t ative inila evalu desi cn iis qualified to an environnent d at normal operating hat six of the seven history of safetyl r valves in other ation.
Response
Presently valves of a similar but earlier design are installed and have been operated in Chinshan 1
and 2
and two SRVs of a
nodi fi ed design are in Browns Fer'r y 3.
No unsat i s ac tory performance has been experienced on the modified design.
A spare SRV of the earlier designs which was installed, into Chinshan 1
and 2
was reported to have failed to fully reclose after a relief operation.
The Chinshan 1
SRV did reclose with no further anonLies noted.
A question exists as to whether gross leakage due to foreign material existed or if in fact the SRV did not fully reclose.
A direct means of
'etermining
.SRV.. position,was.not,-;used..-.-.4; Chinshan, 2.&f9/--
failed to reclose but did reclose fulLy after depressurizing the air inlet supply source.'he failure to reclose has been attributed to a fauLty solenoid and air valve assembly.
The design of the SRVs to be installed into Hanford 2 are a
modified version of those instaLLed in Chinshan 1
and 2.
The response to Question 211.051 contains
'additional infor" mation related to the qualification of Crosby valves.
REV1SED PAGES FOR ODYf< ANALYSIS
NNP-2 P~~XD.'-'"-NT No.
1 July 1978 ABLZ 4.4-(cont'nued)
Page 2 of 2 Gene al Oae at'nor Cond't'ons Average heat
="lux, Btu/h -sq f 145,100 Desicn operating min~
cr't'al power ratio (MCPB)
Co e 'nlet enthalpy at 420
=
>>Pv T, Btu/12 527.6 Core 'nlet temperature, at 420o~ bT(
o 533 Core maximum exit vo'ds within asse
'es, 76.0 Core average void rac 'on, act've coolan 0
maximum fuel temperature,
- 0.418 3,435.0 Act've coolan flow area per
- assembly, in.2 (BOL)
- 15. 824 Core average 'n'et velocityI ft/sec 6.88 Maximum inlet velocity, t/sec 7.28 Total co e pressu e drop, psi 24.74 Core suppor pla e p essu=e drop psi 20.32 Ave age or'ce pressure d op Cen--al reg'n, psi Pe 'pheral region, psi Maximum channel pr~ssure load'ng, psi 6.03 16.54 13.28 4.4-35
r
LIST OP FIGURES Nueher 5.1-1 Title Rated Operating Conditions of tpe Boiling Water Reactor 5.1-2 Coolant Volumes of the Boiling Nate Reactor 5.1-3a 5.1-3h 5.1-3c 5.2-1 g, g.-l h 5.2-2
- 5. 2-3 Nuclear Boiler System PAID Sheet 1
Nuclear Boiler Syst: em P&XD Sheet 2
(RP.D'I)
Nuclear Boiler System PGXD Sheet 3
fHSiu ClOSurC.
~a%Q
&leek SC<<~
DW54 ljeCf Sa4~$ /~'C~
~a~< s is tt f 1
~ ee ClmurC.> re'~l, gl~ gee (OOgM)
Simulated Safety/Relief Valve Spring diode Characteristic Used for Capacity Sizing Analysis Cont ol Rod Drive and Scram Reactivity Versus Time Character-istics 5.2-4 3 ~ 2 Peak Vessel P essure Versus Safety/
Relief Valve Capacity ri~E'EsPaesE a~ paEssuRE V8~~CI -Fox P(EQueWyIoe Pved75 V
~ ~
e ez.y
~siant a ve 5.2-5 5.2-7 Nuclear Boiler System P&XD Data Safety/Relief Valve Schematic Elevation
~ 2 8
Safety/Relief Valve and Steamline Schematic 5.2-9 Safety Valve Lift Versus Time Characteristic 5.2-10 Schematic of Safety Valve With Auxi3.iary Actuating Device 5-zv'
0
b.
The rated, capacity of the pressure relieving devices are sufficient to pzevent a rise in pressure w'hin the pzotected vessel of more than 110% of the design pressure (1.1 x 1250 psig
= 1375 psig) or the following events:
(1) generator load rejection (2) turbine trip (3) loss of condenser vacuum (4) turbine trip without bypass -
3.ow power (5) closure of a3.'a'n steam isolat'on valves (6) pressure regulator failure - fail open (7) loss of auxiliary powe (8) feedwater con"-olle failure - maximum fl'ow c.
Full account is taken of the pressure drop on both the inlet and discharge sides o
the valves.
All combination safety/relief valves discharge into the suppression pool through a
discharge pipe from each valve which is de-signed to achieve sonic flow conditions through the valve, thus providing flow independence from discharge piping 3.osses.
Table 5.2-7 lists the systems which could init'ate during the design basis overpressure event.
5.2.2.2 Design Evaluation 5.2.2.2.1 Method of Analysis To design the pressure protect'on for the nuclear boiler
- system, extensive analytical models representing all essen-tial dynamic characteristics of the system are simulated on a large computing facility.
These models include the hydro-dynamics of the flow loop, the reactor kinetics+, the thermal'haracteristics of the fuel and its transfer of heat to the
- coolant, and all the principal contro13.er features, such as feedwater flow, rec'zculation low, reactor water level,
- pressure, and load demand.
These are represented with all the'r pr'ncipal nonlinear features in mode3.s that have evolved through extensive experience and favorable compari-son of analysis with actual BWR test data./ra A detailed description of~ mode35 %~documented in licen-sing topical report NEDO-10802 (5.2.6, Ref.
5.2-1IIW.
Safety/
relief valves are simulated in the nonl'near representation,
- 5. 2-5
NNP-2 K
hh and the modeMithereby allows full invest'gation of the various va3.ve response times, valve capacities, and actuation setpo'nts that are available in applicable hardware systems.
The typ'cal capacity characteristic as modeled is xepresented in Figu e 5.2-2 for the spring mode of ope ation.
The assoc-iated turbine bypass, turbine control va3.ve, and main steam isolation va3ve charactexistics are also simulated in the mode2$
K5.2.2.2.2
System Design
A parametric study was conaucted to determine the xeauired steam f'ow capacity of the safety/ eliex valves based on the following assumptions.
5.2.2.2.2.1 Operating Conditions jo4.zs 4'-
a.
operat'ng power - 346g PWt (MMof nuclear boi3.er rated power)~
-RDl'.
vessel come pressure 1020 psig, and c.
steamflow - 14.98 x 10 lb/hr.
(105% of nuclear boi'r rated steam flow)
These cond:tions are the most seve e because the maximum storea energv exists at these conaons.
At lower power conaitions the transients woula be less severe.
5.2.2.2.2.2 Transients The overpressure protection system must accommodate the most seve e p essurization transient.
There are two major t ans-ients, the c'osure of a3.1 main steam isolation valv nd a turbine-gene ator trip with a coinciaent cJea~
f the Va'.lung turbine steam bypass system valves, that represent the most seve e abnormal operationa3. t ansients resulting in a nuc3.ear system pressure rise.
The evaluat'n of t ansient behaviox'ith inal plant configuration has shown that the isolation valve closure is slignt3.y more severe when credit is taken only =or indi ect derived scrams; therefore, it is used as the overpressure protect,'on basis event and shown in Figure 5.2-3..
Table 5.2-10 lists the seauence of events of the var'us svstems assumed to opexate dur'ng the main steam line isola-on c'osure with flux'c am event.
l-JESSE:
y~E'8 5.2.2.2.2.3 Scram Lovye: 5~c <'g~~ S:Z-ig b.
control rod drive scram motion - Figure 5.2-3 5.2.2.2.2.4 Safety/Re3.ief Valve Transient Analysis Spec ications
~
\\
Ch V'
n A
r tn i~
Pi
~,
4a n's h
c z "~ c c ln>
~
h Ln a.
Valve groups:
Spr'ng-action safety mode - 5 groups 5.2-6
/
I i(T.n C. o 5C.
c ~
~PP<uVhi r.')C! y
/( s~)
NHP-2 At".ENDHENT NO.
11 September 1980 b.
Pressure setpoint (maximum safety limit):
Spring-action safety mode -
1177 - 1217 psig The setpoints are assumed at a conservatively high level above the nominal setpoints.
This is to account for initial set-point errors anted any instrument setpoint drift that might occur during operation.
Typically assumed setpoints in the analysis are 1 to 28 above the actual nominal setpoints.
High conservative safety/relief valve response characteristics are also assumed.
5.2.2.2.2.5 Safety Valve Capacity Sizing of the safety valve capacity is based on establishing an adequate margin from the peak vessel pressure to the vessel coce limit (1375 psig) in response to the reference transients in 5.2.2.2.2.2.
5.2.2.2.3 Evaluation of Restults 5.2.2.2.3.1 Safety Valve Capacity The sequence of ~ven investigated to the pressure relief H
in Table 5.2-1
~
~
meet code requirements and to evaluate system exclusively.
The required safety/relief valve capacity is determined by analyzing the pressure rise from a NSZV closure with flux scram transient.
The plant is assumed to be operting at the turbine-generator design conditions at a maximum vessel dome pressure of 1020 psig.
The analysis hypothetically assumes the failure of the direct isolation valve position scram.
The reactor is shutdown by the backup, indirect, hich neutron flux scram.
For the analysis, the spring-action safety setpoints indicateg that the design valve capacity is capable of main-taining adequate margin below the peak ASHE code allowable pressure in the nuclear system (1375 'psig).
Figure 5.2-/ X snows curves proauced b ~ analysi~
ts Under the General Requirements for Protection Against Over-
~pressure as given in Section III of the ASNE Boiler and
,Pressure Vessel Code, credit can be allowed for a scram from
.l the reactor protection system.
In addition, credit is also
'taken for the protective circuits which are indirectly de-rived when determining the required safety/relief valve capacity.
The back-up reactor high neutron flux scram is conservatively applied as a design basis in determining the
! required capcity of the pressure relieving dual purpose
- safety/relief valves.
Application of the direct posit'on
'al r:gc e s.2-g~
sg ~s +~ +~<>
) )'",.'F;"*
Vf. SSSSCt lf&l fLVK 4
S)Di llfk LQH C
E Sl l SlE4ll 1 OM J Vf S l SSSS il g I';t Q
+
ISS I
Ul C3 ill 4
SS SS C)
I T)IIE (SE IS C)
ASS ISS LXYCIl IIICU'Stl'ES SKISI i l a '4S
)
V l.t't<~lgl Sl ISS Clll:
la.~
SS lil g
4S
) SQ/i 5 f~llfr IS
.r t
'lttlE (SEC)
CORE FLOII ('C) pygmy,~ >fry/ pfw~ PIG<f-+
NASIIIIIOIMPISI.IC POII(R SUPP(V SVSI(II OV(RPR(SSURC PROlfCIIOII AUAI.VSIS (IISIV RUCL(AR PROJ(CI RO.
2 CIOSIPR( MIIIIIIIAIf1 UX SCRAII IRIP).
fICUR(
b.?- I
. 0
o IN.
b zulrsl r',It I I 2 rfa. IUI
) Ilil: SM i I[tnotill S VLML 4 (otlflt1E'li
" IKJLL)UZ t(ttt Eftl fLIII 0.
I VESSEL I-'f5 IIISE IPSII 2 Slit I.t:.C'IITS r'lsc I:SII
~jntt.I~Pl il: rL('A IZI il liILILI P~iVL I I~4-<ZI
~Oirn:5 V/V.VE ILm IZI 5 Iat.ltl. IUtlfIf'.IZI ry I' I ti W
'I.i;.:Q I".t I1 g;q n
( "J n.0.
V, 0.
IR
'tltiC tSCCI IG.
~Ig).
U.
II, IZe tltiC I"KCI l
I CIXiE litdt SUII lnIIIA,I'I 2 H A SinIEO ttvCLIIICiElt II II SlliSEO tVC I:ZIESI
)70K IlttfI fLlll I Z I 5 Pitve ILI.I I I Z I I Iehaal fLIit 2:ttr A:E IChf I' I 1I 0.
8 Oa.
glfi q9.
jl Q
i i.
-IC}.0.
0.
I2.
IIIE ISICI 0.0.
G ~
LOT fLH IXI IC3-Fi9ure$.2-l HSlV Closure With Flux Scrant lllstalled Safetj/R"..1!ef Valve Capa"ity (PCnf)
ll I
)50.
I IN-:Itf)NNI I LIIX 2 I'l'ln Itrlovn rUrt. Aool
. ll.fjV[ i%Ill[i I:t.firn't f [.UX 5 VLSSII. 5 Etltl FI.OII I vrssrt. pt 2 "i)H I Itlf 5 ltft'AS5 Vt 6 TUINIIHE !
Csi AISF. II'5ll I'IN'5 A)Sf II'Sl)
I,V(I.IN) fo)
I V) I'I(NI tol LVC I'LOII fat, tEAtl FLU)I Ii)
~ Ion.
lg ui 50, LI lliil
- 200, 0.0.
20 I)~
6.
1 ftIE ISEC) 8.
0, 123 0.
~l IJ, 6..
TIIIE ISCCI
@III.
0.
200.
I I.EVEL lit)I 2 H A SEIISI 3 H fl SFIISI ft QIII:IITCI 5 INIIVE FLI II-flff-SEI'-Stf)AI' LEVELIIIICIIESI 0
LEVEL I IHOIES)
- MCSD, I
II )
I
~ I I VOIO Afg 2 nflPI'ISA I 3 Sr)IIN Urr O'Aii HE T.IVITT EON VITT rT)v CT II)t 0.
-100.
U.
liII
%Oil II'IIOIii'LSD II.
6.
TIHE ISECI tlslv CLOSUAEo lt)6)I FLUX SCAI)tf (ofi tO')
I
-2.
0.8 PIC.
- 5. 2 - I.tx IN)CHIIECIIFI01-1.6 TIHE I5EC)
- 2. II 3.2
ill O
n Im c
~ rM
~~O O m m C/l O
a
~M I
4/l n
Cl O
~r m
O CO I/l~R W(
mm n>
%o 0 I/l R~
no 1.0 o L 0B Z
IIIL I
z 2 III K a.
I 0.6 0
z I
II.
III O
IL 0.4 02 670 SCAAM AOO OAIVE END OF CYCLE I scnAM cunvE
'8<
/S 8b To ILT/oi) aF'q IIA~
Cu1IIIE UsEb /4 bbYhl RO/II( yp/f SCAAM CUAVE UTILIZED IN OVEN PAESSUAE ANALYSIS 0
0.6 2B 2.6 TIME Ilocl 3.0
I
't K
C t
A!!:-ND.'!E."T NO.
1 September 1980 1550
/
251 BWR/5 ~ MSIV 14$ /% VOID COEFFICIENT
/
1500 1450 g
1400 I
I0a r
1350 C
C i
I i'
/
////////
/
1300 1250 1200
~
/
J FLLXSCRAM
/
///,
/
I I
I I
I I
I I
SITION SCRAM
,40 50 6.
70 '0
/
SAF FTY I
I I
I I
90 100 ALVECAPACITY ~ 5 NB RATED STEAMFLOW I
I I
I 110
'20 130 0
150 160 170 7
8 9:3 11 12 13 <'f4 15 18 17 18 NUMBER OF OPERATING SAFETY VALVES REPLACE un-H /JEtu FlGI/QE s4ASIiItiGTOil PUBLIC PO/iER UPPLY SYSTEI<
IUCLEAR PROJECTI
'O.
2
?EAY. VESSEL PRESSURE VERSUS SAFETY/RELIEF CAPACITY FIGURE 3e 2
III 0
I.I C ~
~ t
~
t L
l
-. ~ "f h tl T
J I,
1 as/
a..
P X
I/I As O
A Pl I
/V 1400
~Il A O O m
Pi O
~
cs
'cs
.Sssl g
1300 I/
~s s
~'
Pl 7C 3" (/I PT
~m Pl P A 0 U Al Wm A I/I
~ t/I I C-M A7 Pl I/I N Pl r I/I Cl C:I/I
-I C/I Pl mw W X7 IJI IJI a
UI; K
Isl =
II>
ts:
0r0 sa IU 1250 1200 1150 1100 1050 0-s 251 OWR/5 14 g/ls VOID COEFFICIENT TIME Iseel 10 12 14
,."./i h1SIV F LUX SCRAM,FOR 18 SAFETY VALVES I/I Pi IIs
~ r m iJ
~ I0'n IQ g s/
I-r.
IO Q co
~
Cl
i
.I) (540) (E)
CCNCRAL CLKC)IIIC COll
. ACIICIICC'IAOY.ll~ Y.. U.S.A.
)OOX)40 7/IOO Ine jleeeeeone S.2-5.
e I
I' I
~
- il.
~ ~
'l I
e
~
,' i!i li,.
-I'Il,'L I.i e
' "I e
~
j l
~ e
- j. ~
'I I
~
~
II-A'I
'e'lil iiiI r
- I,I I P
~ I Ilii li:e Illi
~
~
~ I. ~
i l.l e i I
~ e
Table Number
- 15. 0-1
- 15. 0-1A 15.0-2 g>,03 TABLE OP CONTENTS (Continued)
TABLES Title Results Summary of Transient Events Applicable to WNP-2 Summary of Accidents Input Paramete s and Initial Conditions or Transients S.E+u<AE.Q opBRAT><< ">T 1/alua5 PacCe 15.0->4 15.0-17 15.0-18 gg, 0 - goo.
C
\\
I
TABLZ OP CONT~TS (Continued)'ZGUR"S Pi re Number 15.0-1 15.0-2 gy 0-3 Title.
Typical Power Plow Map Scram Posit'on and Reactivity Character'st cs ops+~( I+$ Cf ~ Ll Lt.
ceca.m
~P~
Pace 15.0-21 15.0-22 gg,O--~
15-iv
1 w
AMENDMENT NO.
8 February 1980 TABLE OF COl!TE!!TS (Continued)
FIGURES 5!umber Title 1S. 2-1
- 15. 2-2
- 15. 2-3
- 15. 2-4
- 15. 2-5 DD~
v7 Qfpggg pMi-ow Generator E,oad Rejection 15.i2-14 QC C pe eje I
urbine Trip, Trip SCRAM, Bypass and
- 15. 2-26 RPT On
> ~TH o~~
Q ~PAD 7'P'~>
Turbine Trio D"-~
15 227 Three Second Closure of All Main Steam
- 15. 2-36 Line Isolation Valves with Position Switch SCRAM Trip
- 15. 2-6 Loss of Condenser Vacuum at 2 inches Pe Second
- 15. 2-45
- 15. 2<<7
- 15. 2-8
- 15. 2-9
- 15. 2-10
- 15. 2-11
- 15. 2-12
- 15. 2-13 1S. 2-14
- 15. 2-15
- 15. 2-16 Loss of Auxiliary Power Trans ormers Loss of All Grid Connections Loss of All Feedwater Flow Summary of Paths Ava'lable to Achieve Cold Shutdown ADS/RHR Cooling Eoops Activity Cl Alte nate Shutdown Cooling Path Utilizing RliR Loop B
RHR Loop C
RHR E.oop B (Suppression Pool Cooling Mode)
Activity C2 Alternate Shutcown Cooling Path Utilizing RHR Eoop A
-Vessel Tempe ature and Pressure Versus Time (Activity Cl.b.l or C2)
- 15. 2-56
- 15. 2-57
- 15. 2-63
- 15. 2-72
- 15. 2-73 1.5.'2-77
- 15. 2-78
- 15. 2-79
~
- 15. 2-80
- 15. 2-80a 15-vxii
lvNP-2 15.0.3.2.1.3 Single Active Compo..ent Pailure or Single Operator Error Analysis a.
The undesired action or maloperation of a single active component, or b.
Any single operator error where operator errors are defined as in 15.0.3.2.1.2.
15.0.3.3 Core and System Pe formance 15;0.3.3.1 Introduction
)
The various fuel failu e mechanisms are cescribed
'n 4,.4, "Thermal and Hydraulic Design"..-'.voidance of unacceptable safety limits 1 and 2 (4.4.1.4) or incidents of mocerate frequency is verified stat'stically with considerat'on given to date,'alculation, manufacturing, and operating uncertain-t'es.
An acceptable criterion was determined to be that 99.9%
of the fuel rods in the core would not be expected to exper-ience boiling transition (see GET%3 Refe ence 15.0-2).
This c 'terion is met by demonstrating that incidents of moderate frequency do not result 'n a minimum c-iti a o w~~
(HCPR) less than 1.06.
Tits r+c"gr s a
-s~te CPR op rat-5 tKPR(d c use an a-n t v~Vwe in en h
M< t is h t The HCPR during signifxcant abnormal events zs calcu atea
~
using a transient core heat trans e
ana'ysis computer pro-g am.
The computer program is based on a multi-node, single channel thermal-hydraulic model which equires simultaneous so'u"'on of the partial different'al ecuations for the con-servation of mass, ene gy, and momentum in the bundle, and which accounts for axial variation in power generation.
The pr~~ arv inputs to the model include a physical description of the bundle, and channel 'nlet flow and enthalpy, pressure and power generation as functions of time.
A detailed description of the analyt'cal model ma be found in Aaoendix C of Reference 15.0-2.
T W'a
~'.'-.a 6u M its
~aidan g,,g
- tion rate 1%i (13.$ XA/M) d a th Wk Piheap' a 'qrg-e4'ait 4)'K 15.0-8
INSERT B
Deteraination of the steady-state operating limit is accomplished as fol1ows:
1)
The change in the critical power ratio (hCPR) which would result in the sa,ety limit CPP. (1.06) being reached, is calculated for each event.
These values are shown in i&le 15.0- l.
2)
The bCPR value i then added to the safety 1'imit CPR value (1.06) to result in the event based HCPR except for events whose 4CPR is calculated using ODYH.
3)
For events whose hCPR is determined by ODYH (all rapid pres-surization events) the event based HCPR is determined in conjunction with correction factors, the ACPR and the safety limit CPP,.
Those correc ion factors are explained in detail in Section 3/4.2.3 of the Technical Specifications.
These results are given in Table 15.0-3 and Figure 15.0-3 for the limiting transients.
The opera ing limit HCPR is the maximum value of the
//
event MCPRs calculated from the transient analysis.
The maximum calcula+ed transient MCPR is depicted by the solid line in Figure 15.0-3.
Maintaining the CPR operating limit at or above this operating limit assures that the safety limit CPR of 1.06 is never violated.
r f/626 2/4/81
I'AULE 15+0-I DESDLTS SUHHARY OF TIINISIEIIT EVEWTS APPLICNILE TO Itlp-2 Para-graph Flguro I.D.
I.O.
~Doser I t 1 on Haxlmum Haxlmum Ileutron Domo Flux Prossuro
~IIQll ~sl ltax Imam Hex lmum S loam Vossol Line Prossuro Prussuro
~sl
~sl Hex lmum Coro Avorago Sur lacn Iieet Flux 5 of A.CPA~
+Frerluoncy Inlt lal Cato~or Duration ol Olowdoxn Oura-IL>. ot tlon Valvos of 1st Dlov-Olow-dovn dovn soc 15+ I DEOIEASE Ill ONE COOLAIIT TEHPEIIATIIIE
- 15. I. I
- 15. 1-2 Loss of Foodvator Hoator, tlanual Flex Control 124 '
1030 15+i 2 15'-3 Foedxator Cntl Fall-
/yv'..Z
- Iitrp, uro, f4x Oomand 476<0-fl4F 1070 1001 l/7E l/I o H 70-H24 117. I
/nfI.P'40.8 o.oB 0) fr09 p)ga 15.1 3
15 1-4 15+ I ~ 4 Prossure Regulator Fall-Open Inadvortont Opening of Safoty or Do-Ilof Valvo 104.3 1090 1117 1097 100. I
~ o.o+ +
SEE TEXT 2
6.4 15+ 1.6 15.2 15 2.1
.15e2o2 15+2-I 15+2.2 15.2-2 DIA Shutdovn Cooling Halfunctlon Docroas-Ing Tomporaluro SEE TEXT IIICDEASE Ill DEACTOII PIIESSQIE Prossuro Aogulator Fall - Closod SEE 15.2.2 and
- 15. 2.3 X/Bypass on
)Zo z fit4 A95-1148 Gonerator Load Ao- 'g/r.8
]/rr/5 J oct Ion, Bypass-Oni ~i ft57 Generator Load Ao-
- Joctlon, Oypass-Off lo7'-8 HOr5 a
(<)
O.otf 445-a n.OQ f'r) h05
&7 I0
-B.Q.
15.2.4 15.2-5 Inadvo 'ent HSIV Closuro IB6. 2 1154 15.2.5 15.2-6 Loss ot Condonser Vacuum Loss ot huxlllary Power Transformers 157. 5 1135 104 3 1153 15.2.3 15.2-3 Turb lno Tr Ip, Oypass-On 147+ 5 1136 15.2,3 15.2-4 Turblno Trip, Bypass-2/8. 3 rl 13 Ol I 233K 4165 1163 1121 12 o/
(/L9 I+17 1191 1146 1162 1120 1172 1149 101 ~ 7
/oC.Ir IOB-.9 100,0 102.6 100,0 o.of LAD a
Cr>
o.og 4'
o.o'/
a
< o.l7
~8~
a oo aP a
IO 5.5
<<7 IB 4M IB 5.7 IB 5.4 0
0
TAULE 15.0 ICon.
Dod)
Para-
<graph Figaro I.D.
- I.O.
Descrlptlon Max lmIBB hLBx lmIBB Max lnuBB Neutron Domo Vossol Flux Prossuro Prossuro
~ISfl ps lg Haxlmum StoDBI Llno Prossuro
~sl 11 Haxlmum Coro Avorago Surtaco Itea I F I ux httntBnwn-S ot ACI'IIB"I Froquency Inlt i el Ca In~<or" Duration of Dlowdown Dura-Ito. of tlon Vnlvos of 1st Dlow-Dlow-down down soc 15.2.6 15.2-8 Loss ol Al I Grid Connocllons 144 ~ 2 '157 1177 1137
< D.o9 IO I ~ 6 1.1-.15~
0 IO 5.5 15.2.7 15.2 Loss ol all Food-wator Flow 104 3 1095 1107 1095 100+0
~D OC1 +
4rhcRC 0
2 6.0 15 2.8 Foodwater Plplng Oroak SEE 15.6.6 15.2.9 15.3 Falluro ot AIIA Shut-down Cooling SEE TEXT DECAEASE Bl AEACTOA COOLANI'YSTEH FLOII-AATE UI ED I
Ul 15.3.1 15.3-1 Trip ol One Boclrcula-t lon I'ump Motor 104.4 1021 15.3.1 15.3-2 Trip ot LIth Boclrcu-I at lon Pump hlotors 104.4 1104 1.5.3.2 15i3-3 Fast Closure of Ono Hain Aoclrc Volvo 104 3
IIOI 15.3.2 15.3-4 Fast Closure of Two HaIn Aoclrc Valves 104.4 1105 1061 1116 1115 1115 994 I 100 1097 1100 100. 0 100.1 100.0 100.0
-o.~'">>
a o.oo" ~
a
~D.04+I 0
0 0
6 53 2
6.8 6
5.4 5;
rg IM 15.3.3 15.3-5 Sol zuro ot Ono Aoclr-cul at lon Pump 104.3 1105 1117 100 IDOL2
~o ao cksIC~
c 6
5+4 15.3.4 Aeclrc Pump Shaft Oroak SEE 15,3.3 15,4 15.4 I ~ I 15.$. 1.2 15.4 2
15.4.'3 BEACTIVITY AIS POWEA D ISt A IOuTIoil AIIOI4ALIES BWE - Aofuol lng
'EE TEXT AWE - Startup SEE TEXT AWE - At Power SEE TEXT Control Bod Hls-opuratlon SEE 15.4.1 and 15.4.2 Ul ~
ta'0 W cc A Ia t3 8 R 0 tu O H Pg
<D O co
~
O 15.4.4 15.4-6 Abnormal Startup of Idio Aocirculatlon Loop 94 ~ 2 981 995 970 g.I7o~ I'~)
146.6 ~&
0
TABLE 15.0-1 (Continued)
Para-graph I ~ D.
15'.5 Figure
- l. D.
ri5.4-7 Doser~i') tion Fast Openilug of One Hain Recirc Valve Maximum Haximum Neutron Dome Flux Pressure 202.9 980 Haximum Vessel Pressure 1000 Haximllm Steam t,ino Pressure
~)sl g 971 Haxlmllm Core Ayerage Surface Beat Flux L of loirial 141.0 Hhabilum ACPR >
Frequency C~aro or
~o.c c+" C<)
06 A*
a Duration of Blowdown Durr1 No. of tion Valves of 1st Blow-Blow-down down
. sec 15.4.5 15.4-0 15.4.7 15.4.9 C)
- 15. 5 Fast Opening of Both Hain Aecirc Valves Hisplaced Bundle Accident Rod Drop Accident INCREASE IN REACTOR COOLANT INVENTORY 222.2 977 SEE TEXT 1000
-o.oos>
~'>)
134. 6 WM~
a 15.5.1 15.5.3 Inadvertent NPCS Pump Start 104.3 1020 1061 993 100.1 BWR Transients SEE APPROPRlATE EVENTS It/ 15.1 and 15.2 a:
Moderate Frequency bl Infrequent Incident c:
Limiting Fault Estimated Value CyrC C~SP4V4f lunar~
DCPi4 i~ Ba~ t oP AP iu~ <~
+
Y
~ (]) ggyg g~)~(.l 5
~.-)L~
4 ~
aJ )~l.l r ej 7
~4C~e<5 L,
l.pl (S,.'...,
.a 9 ~-r.,/or tl.~.i. VI.
1A C.l' -'Cl5 LJ c-h oDY'N Agent Jvfdo-zvls~l <<or, zvizyg I
Dig f'jQ, rt
~l hoor'~I Ã ( ei '
TABLE 15. 0-2 ZNPUT PARAY~TERS AND ZNZTZAL CONDZTZONS FOR TRKNSZENTS 1.
Thermal Power Level, MWt Warranted Value 2.
3.
Analysis Value Steam Flow, lbs pe hr Analysis Value Core
= low, lbs per hr Feedwater Flow Rate, lb per sec Analysis Value 5.
=eedwate Temperature, F
0 6.
Vessel Dome Pressure, ps' 7.
Vessel Core Pressure, osig 8.
Turbine Bypass Capacity,
% NBR 9.
Core Coolant Znlet Enthalpy, Btu per lb 10.
Turbine Znlet P essure, psig 11.
Fuel Lattice 12.
Core Average Fuel Cladding Gap Conductance, Btu/sec-ft F
~2 0
3323 Sile 14.98 x 10 108.36 x 10 4161 424 1020 1031 25 529.3 975 8 x 8
0.1667 332 3 3s t'~ (i) j j.99)L/o~
lo9.5 g!O (I)
SIC I QM Cj
/0~0 jo3I 25 pe.3 g75
- oggqg, C>)
o.isa~ (>)
13.
14.
15.
16.
Core Leakage
Nominal EOC-1 Analysis Data 11.84
- 1. 24 ('I) 1.06 0.227 0.215
>l SY
(<)
/ o@
oi&7
- o. X~S
- 15. 0-18
0
NNP-2 TABL=- 15.0-2 (Continued) 17.
Void Coe ficient (-) 0/% Rated Voids Nominal >>OC-1 Analvs's Data for Power Zncrease Fvents 7.48 12.70 18
'9.
Analys's Data for Power Dec=ease Events Core Ave age Rated Void.
- Fraction,
% (Sv=-go~i sl ~rc)
Scram Reactivity,
$ hk Analysis Data 7.065 41.32 Figure 15.0-2 (s3 q].37 ())
Control Rod Dr've Speed, Pos'tion versus time 20.
Figure 15.0-2 F /gu~t'.
2.41 Pal 111.5
/~!.g Crosby 18
//I.0
/o/ Y CcZ.o58 /
lQ Quantity Xnstalled Relief Function Delay, seconds Relic Function Response, seconds Set Points for Safety/Relief Valves Safetv =unction, psig 0.4 23.
24.
0.1 25.
!: i, lip'7 ll9'7)
Iao7
/ lgl7 illl',~ l)24(l)
/l34~ llQG
- 1177, 1187,
/'"7g 1207, 1217
- 1091, 1101, llll,ll21, 1131 Simulated Relic
"-unction, psig Number of Valve Groupings Safety Funct'on, No.
Relief =unct'on, No.
HLgh Flux T io alysis set point (129'BR 26.
27.
x 1 ~ 043) I
)p/ gp
)ag,ko 21.
Jet Pump Ratio, H
22.
Safetv/Relief Valve Capacity, NBR Y
g/,-,c7/ vow" e~<"'
12l 3 osig F gg ~f t P,/AC /7Y (8 //OC PSI t~anu acturer
- 15. 0.-19
HNP-2
&1ENDYa"NT NO.
1 1
September 1980 TABLE 15.0-2 (Continued) 1071 29.
Vessel Level Trips, Feet Above
'K.YpQ 28.
Hign Pressure Scram Set Point, psig 0DV<
Io 7I Level 8
(L8 ), g~ i~c~es Level 4
(L4), fee i~cNeS Level 3
(L3)i ~ )~c+c Level 2 (L2),
~a t i~eke.5 30.
APRN Thermal Trip 5'5. 5 31 5 i 7 l".5
(~~A C-)
(8)
(6')
Set Point, 8
NBR 6
100% core flow 122.030 f24 G30 31.
Recirculation Pump Trip Delay, Seconds 0.140 32.
Re circul ation Pump Trip Incrtia time constant for analysis, seconds (z)
S (7)
Me rES:'I) co> ('ac" Q.J +
J te5<
d~S('g~ da4~.
Ml 2} p$ %'g R 're pre Su.at genic re.frog<'4 Pgn5 iv WXP c rrap.
(3) 4 here< S~ r i+ pa~ Jgc.
~ av cc (5<<~ gee> )
ig et> C ~ +1,<
f8 $+Q l J.
pl'c f'l'45kr1 $40i+~
~v We Sup I l s~g l OWcL g-.) AM (q) ge~
Tat.J<
a' 3
(53 CS 'IQ VALUE5 g g,q Ca/cu.la~e.J
,t,e goJe
$'ar
~ id <$ Cgcle.
i (t',)
Para.me,-,'er no-;
u,se d The inertia time const i>)
2 IL Jon t=
where t g To 0
n'.
/S.0~0 To
+t e ODYL AVALVSI S, ant is defined by the expression:
inertia time constant (Sec).
pump motor inertia (lb-ft2).
rated pump speed (rps).
gravitational constant (ft/sec2) pump shaft torque (lb-ft).
GErr ERALo ELEgrp~g NUGAE*R ENERGY Dl rsloNS
/
TABLE 15.0-3 REQUIRED OPERATING LIMIT CPR VALUES Pressurization Events:
Load Rejec ion without Bypass Turbine Trip without Bypass Feedwater Cortroller Failure Load Rejection CPR (Option A
CPR (Oo<
1.20 1,12 1.19 1.11 1.19 1.16 1.15 1.07 Hon-Pressurization Events:
CPR Rod '<<fithrawal Error Loss of Feedxater Heater 1.24*"
1.22 Includes adjustment factors as specified in the HRC safet report on QDYN, HEDQ-?4154 and NEDE-?4154P Required OLCPR using either Q tion p son A or Opt~on B adjustment factor y
g y of the turb>ne-generator trip events r
ess o
r equenc cate or
1.0 S40.21
~ 537.05 40 z
0.8 z
08 C
ZO v
0.4 0
X 04 OV 678 SCRAM ROO DRIVE ENO OF CYCLE 1 SCRAM CURVE SCRAM CURVE USED Ik AkALYSIS 30 20 10 0
s IU lib C
C:V tO TIME Isccl PCV S~g Por+ crhf oM 'EC.vg ~ c ~gJ i~
O>VA ~o 1yZ, s 15.0-22 WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.
2 SCRAl< POSITION ANO REACTIVITY CHARACTERISTICS FIGURE 15.0-2
rvwuncv or~rre4 CI'R lt~l'pf ghlf p5fhl.
CLA9l. LJ49Y g.z5 lRO
/.20 l.s5 8P ~
- a. +5 vrto l.I5 l.so
~le ~
gP.
/./0 AC6
@vs Ea Ez@ =Z'g p]g. /g.g-~
g'n'~
Opardhrrg
~PA D~i<
w.
Sc~a<
Sa-.ed
-2 I
reactor core isolation cooling svstem and the high pzessu e
core spray system'o maintain long-term water level control following tripping oz feedwate pumps.
15.1.2.2.3 The "-fzect of Single Failures and Operator
=-rrors Xn Table 15.1-3 the first sensed event to initiate correct've act'on to the transient is the vessel high water level (L8) trip.
Multiple level sensors are used to sense and detect when the water level reaches the L8 set point.
At this point in the logic a single fa'ure w'll not initiate" or prevent a, turbine trip signal.
Turbine trip signal trans>>
m'ssion however is not bu'lt o single fa'lure c iterion.
The esult of a failure at this point would have the effect of delaying the pressurization "signature."
However, high moistuze ente ing the turbine will be detected bv high levels in the turbine's moisture separators which trip the unit.
Xn
-add't'on, excessive moisture entering the turbine will cause vib at'n to the point where it too will trip the unit.'cram trip signals from the turbine are designed such that a single failure "will neither in'ate nor imped'e a reactor scram tr'p initiation.
See 15A for a detailed d'scussion of this subject.
15.1.2.3 Core and System Performance 15.1.2. 3. 1 Mathematical Model The o p+~ ~a lj&spi Qi-. lp.lp.p'.1++ p'sp'd 15.1.2.3.2 Xnput Paramete s and Xnitial Conditions These analyses have been pe ormed, unless othe wise noted, with the plant conditions tabulated in, Table 15.0-2.
Fnu o-cyc e
a 1 rods out) scram characteristics are assumed.
The sazety-relief valve action is conservatively assumed to occur with higner than nominal set points.
The transient is simulated by programming an upper limit ailure in the feedwater svstem such that 1358 feedwater flow occurs at tne design pressure of 1060 psig.
- 15. 1-12
t
IdzeaT g The predicted dynamic.beh>vior has bccn dete~<ncd using a cerquter simulated, analytical wdeI of a generic direct-cycIe S'~'R.
This radeI is described in detail in Reference l~. )-3
~
This co~uter model has been i~roved and veri-fied through extensive ccnyarison of its predicted results xith actual 8'~R test data.
'The nonlinear cc'refuter simulated analytical model is designed to predict asso"lated transient behavior of this reactor.
Se'c of the significant features of the ~e3 are:
f
, e.
An integrated one-di.<<nsional core model is ass=..ed xhich includes a detai led description of hydraulic feedbac" effects, axial poxer sha,",e chaqes, and reactivity feedbacks.
b.
The fuel is represented by an average cyiindr ical fue1 and cladding
~del for each axial location in the ccrc.
l c.
The stem Tines are cedeled by eight pressure nodes incorporating aass l
and centum balances xhich xill predict any xave phencaena present in the stma line during pr essurization transient.
The core aver'age axial xater density and pressure distribution is calculated using a singIe channel to represent the heated active flox and a single channel to represent the bypass fiox.
A model, represent'.ng liquid and vapor aass and energy conservation and mix-ture ~ntum conservation, is used to describe the thermal-hydraulic behavior.
Changes in the flow split hebr n the bypass and active charm I flox ar accounted for during transient events.
e.
Principal controiie~ functions such as feeCxat r f1~, recirculation J
flox, re'actor xater level, pressure and load Caaand, are represented together
<<1th their dominant nonl ineyr characteristics.
f.
The ability ta simulate necessary reactor protection syst~ functions is prov id&.
1 g.
The controI syste"...s and reactor protection system models are, for the most part, identical to those eryloyed in the point reactor model, xhich is Cescribed in detail in Reference l5. l-l and used in analysis for other transients.
4
4 1
15.1.2.3.3 Results WNP-2
//77 The simulated feedwater, controller transient at 105%
NBR steam flow is shown in Figure 15.1-3.
The high water level turbine trip anc feedwater pump trip are initiated at approximatelv '
sec.
Scram occurs simutaneously from stop va ve closure, and limits the neutron flux peak and fuel thermal transient so that no fuel damage occurs.
R4 f D
~h A
J OO 1
The turbine bypass system and the main steam sa ety/=e ief valves open to limit the peak vessel bottom pressure to~ psig, and the nuclear system process ar 'e pressure immit is not encanqered.,
The bypass valves subseauent'y close to re-establish oressure control in the vessel dur' shu down.
Thegle el 'l 'gr ua op phq lo~<e el solprio. r e
e g po'n ac iv C
gpgggg~ svat
. s or '
e 1
el coht 1.
p Pb 15.1.2.3.4 Cons'deration of Uncertainties All systems ut'1'd for protect'on 'n this event were assumed to have the most conservative respon:
(e.g., rel'f setpoints, scram stroke time and work characteristics).
Expected plant behavior 's, therefore, expected to lead to a less severe tran -ient.
15.3..2.4
,Ba rie Performance As noted above, the conseauences of this event do not result in any tempe atu e or pressure transient in excess of the crite ia for wh'ch he fuel, pressure vessel or containment a
e des'gn c; the efore, these barriers maintain their intearity and function as designed.
15.1.2.5 Raciological Conseauences While the conseauerce of this event does not result in fuel failure it coes result in the discharge of normal coolant activity to the suppression pool via SRV operation.
Since this act'vity is contained in the primary containment, there will be no exposure to operating personnel.
Since this event does not result in an uncontrolled release to the environment the plant ope ator can choose to leave the activity bottled up 'n the conta'nment o
discharge it to the environment under optimum meteorolog'cal and release conditions.
Ef puraing of the containment is chosen the release will be in accorcance with established technical specifications;
.the"efore, th's event, at worst, would only result in a small
'ncrease in the yearly intergrated exposure level.
?
Cf p 0c (0
9 - ~
yo In
- 15. 1-13
0
Events caused by low water level trips, including Ir%
V
~
m~ closure of main steam line isolation valves, and initiation of Hp~~ and PZIC core cooling system functions are not included in the simulation.
Should these events
- occur, they ~ill follow s~time after the primary concerns of fuel therral margin and overpressure effects have occurred, and are expected to be less severe than those already experienced by the system.
TABLE 15
~ 1-3 SEQUENCE OP EVENTS "OR rZGURE 15.1-3 Time-sec Event 0
//. CSl~
Initiate simulated failure of 146$ upper limit on feedwater flow.
7~e-g;yC.
~fpkK gqi uPS C-. ae'~
L8 vessel level set po'nt t ips main turbine and feedwate pumps.
Turbine bypass operation initiated.
~~ sKzs X8-;B5 /ggg Reactor scram t ip actuated from main turbine stop valve position switches.
Recirculation pump trip (RPT) actuated by stop valve position switches.
~
1
)
e e
e e
~ ~
8 IYsV Recirculation pump motor circuit breakers open-causing dec ease in core flow to natural circulation.
Group 1-5 relief valves actuated due to high pressure.
- 15. 1-14
)
I/l O
n ImD I-Mn CO m ~
n I/I O
~
QU
~~
I C
I/I III m
150.
I tJEUIAIXf 2 PEAtf FIKI 3 AVE Q)AF fi FEEMHE 5 VESSEL S LUX CEtfIEB TEHP tEAT FLUX
'R EAH FLOH 100.
0.
PAES AI PSI)
BES Al IPSI)
.liE f'.
1%)
FLOff 1%)
AH FLf%f 1%)
2 STH LIHE 3 TIMI)IE I ff AEL1 5 NPASS 6 TUI)OItlE I VESSEL Pf 5 BISE I ll mm M III QO o D m
n C)
C)
I IIll nD I
m 0.0.
10 100.
0.
20.
30.
T)HE ISEC)
I)0.
I COAE It)L 2 H H SE))Sf 3 tf ft SEt6!
ff L4 IN 5 Of)IVE FL r M) fOTun.O) 0 LEVEL(lt)DES) 0 I.EVE).flt)C)ES)
I FCSA 2) 11%)
-100.
0 120.
IIO.
10.
I tIEU 2Q 3 Ff.
20.
30.
TIHE fSEC)
AfZI FLUX ACE )IEAT F t ACF ftt P)i
')NEI) Y)kl UX SOS)L 161Eh I)0.
CJI
~
Q I~ m 0
10.
20.
30.
I)HE fSEC)
I)0.
0.0.
25 50..
)5.
COAE FLOH 1%)
100.
150.
1 I ttottrmi I
- . >1( Ctli lt J IIVC:0 I'I CL LL
>I I.l 5 VII"'
)
I.l>X i'IIVI Fl>FL A(10>
I.I, t>>.>>l tet>I(
I >>.>(I rl>>> I'Lo>t 125.
VI "..~'.I. IIIl.
>'IXII ir>>st i
.'iil I it:.'! >".
r.'I.";c.>r..l>
I ill:.i."f!CI";I'> ~liit ilg>ttL6>
5 ti>:I. Ir>jl> I!I,'I>. g>.LIi>ht>>
<<I~>" "
!iL>ir~-ai>'1I.)
~
100.
IaI Iit 50.
lit 75, 25.
0,II.
l2.
ll<<sect'.
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I
15.1.7 1 5 ~ 1-,1 15.1-2 fs'-g REFERENCES R.B. Iinforp, "Analytical Nethods
~
Evaluations for the General Electri.
Reactor," April 1973 (HEDO 10802).
"Plant Design Assessment Report for ZOCA (Revis'on 2)," Washington Public Power Sup.
transmitted to HRC as Amencment No.
6 to the September 19, 1979.
> DA R 5~Tp'/y El/g /g p$ 44
~>
Sped <<c /
c/ee4:c.
Toy,:eq/ >ego.~:
Quoi:4'ca~:ow a0 /Le
<u~
) ~ > ebs o~q /
co ~a
~~~si ew7
~<Je/ g,>>
pa./'>g gqQ>>
Peg'g 4vs'" 4 F~ -zg igq'EIE z-J Y~5$lp Z~~
~J 3
I/80.
- 15. 1-31
15.2.2 GENERATOR LOAD REiECTXON 15.2.2.1 Hen ificat'on of Causes and Frequency Class'ica-tion 15.2.2.1.1 Identification of Causes Fas closure of the turbine control valves (TCV) is initiated wheneve electrical gz'd disturbances occur which result in significant loss oz elect ical"'oad on the generato=.
The'u bine control va3.ves a
e equ'red to close as rapidly as possible to prevent excessive ove speed of the tuzbine-genezator
('T<<G) oto Closu e of the main turbine control valves will cause a sudden reduction in steam flow which re-sults in an increase in svstem pressure and reactor shutdown.
15.2.2.1.2 15,2.2.1.2.1 Frequency Class'fication Generator Load Rejection This event is categorized as an incident of mode ate fre-quency.
15.2.2.1.2.2 "Generator Load Rejec 'on with Bypass Failure
~ g J-g
.<<eguqp gy qveuP This event is categorized as
. =th the following cha aeter't's':
F equency:
0.0036/plant year MTBE:
278 years Frequency Basis:
Thorough searches of domestic plant ope a-ting ecords have revealed three instances of bypass failure dur'ng 628 bypass system ope ations.
This gives a pzobabi-lity of bypass ailuze oz 0'.0048.
Combining the actual frequency of a generato load rejection with the failure rate of the bvpass y'elds a
recuency of a generator load rejec-tion with bypass ailure of 0.0036 event/plant year.
15.2.2.2 Secuence of Events and System Operation j
15.2.2.2.1 Sequence of Events 15.2.2.2.1.1 Gene ator Load Reject'on
>> Turbine Control Valve Fast Closure A loss of generator e3.ectrical load from high power condi-.
tions produces the sequence oz events listed in Table 15.2-1.
15.2-4
NNP-2 15.2.2.2.3 The E zect oz Single Failures and Operator Errors Mitigation of pressure
- increase, the basic nature" of this transient, is accomplished by the reactor protection system functions.
Turbine cont ol valve trip scram and recircula-tion pump trip (RPT) are designed to satiszy the single failure cr'e 'on.
An evaluation of the most limiting sin-gle failure (i.e., failure oz the bypass system) was consid-e ed n this event.
Details oz single failure analvsis can be found in 15A.
15.2.2.3 Core and System Pe formance 15.2.2.3.1 Mathematical Model "'S. i. z.S.l simulate, th's event.
15.2.2.3.2 Input Parameters and Initial Conditions These analyses have been performed, unless otherwise noted, with the plant conditions tabulated in Table 15.0-2.
The turbine digital elec" ohydraulic control. system (DEH) powe /load imbalance device detects load rejection before a
measurable spe d change takes place.
The closure characteristics of the turbine control valves are assumed such that the valves ope ate in the full arc (FA) mode and have a full stroke closure time, from fully open to fully closed, of.15 seconds.
Auxiliary powe would normally be independent of any turbine-gene ator overspeed effects and continuously supplied at rated frequency as automat'c fast transfer to auxiliary power suppl'es normally occurs.
For the purposes of worse case
- analysis, the ecirculation pumps are assumed to remain tied to the main generator and thus increase in speed with the TK overspeed until t ipped bv the recirculation pump trip system (RPT).
The reactor is ope ating in the manual flow-control mode when load rejection occurs.
Results do not significantly differ if the plant had been operating in the automatic flow-control mode.
The bypass valve opening characteristics are simulated using the specified delay together with the specified opening char-acte istic recuired for bypass system operation.
', 15. 2-6
Alt. ugh th sure of ;. in steam 'lat'on v " es as ca edby owq ter el rip
(
) i incl ed i th s'-
la ion, e flaws f om initiati of IC d HPCS c
e c oling yste zun tions pre n
incl ded s nodal tar up a d ac a io can take u6 to 3
seconds b fore~effe ts e
eal'.
f th se events o~ur, timey w'll allow ometQn fte~ the ri y concerns
~ fuel/marg'n and/ over res@or cts )eve assed ard a~
expected result in ezfeqt
/-
ss seve chan thos reacy ex e~'enced b'"
react r svs"'em.
15.2.2.3.3 Results 15.2.2.3.3.1 Genexator Load Rejection with Bypass
=igu e 15.2-1 shows the results of the generator t ip rom rate" powe Peak neutron flux rises above NB rated conditions.
<n.ar gez.9 ~
The ave age surface heat lux peaks at W~W of the initial value and HCPR does not signi"icantly decrease below its
'nit'al value.
15.2.2.3.3.2 Gene ator Load Rejection with Failure of Bypass z=~
Figu e "15.2-2 shows that, for the t"ase oz Bypass fa'lure, peak neutron flux reaches about ~~ of rated, average surface heat flux reache~ ~~ of its init'al value.
t i e
e
's 1
z e
n 'nf au tt it o
~
m' b
G T r'te 'a and th NCP 1',it p
t~ dto f 1 bl h
s y
mi fo s
fAo e e f e e
HC.
ac s
.05 d'or th erI 15.2.2.3.4 Considerat'on of Uncextainties The ull st oke closure t'me ox the turb'ne control valve oz
.15 seconcs is conservative.
Typically, the actual closure t'me is more like.2 seconds.
Clearly the less time it takes to c'ose, the more severe the pressu ization ez ect.
All systems utilized fox protect'on in this event were assumed to have the poox'est,allowable esponse (e.g., relief 3.5, 2-7
1 closure of rrzin steam line isolation valves,
'nd initi~t;on of pygmy )PE~ and,.CIC core cooling system functions are not included in the simulation.
Should these events
- occur, they sill follow sometime after the primary concerns of fuel ther.al rargfn and overpressure effec.s have occurred, and are exp c od tc be less severe than these already e"perienced by the system.
NNP-2 set points, sc am stroke time and work characteristics)
Expected plant behavior is, the efore, expectec to reduce the ac"ual severity of the transient.
15.2.2.4 Barrier Pe formance 15.2.2.4.1 Generato Load Reject'on Peak p essu e remains within normal operat'ng range and no "hreat to the barrier exists.
'5 ~ 2 ~ 2 ~ 4.2 Gene ator Load Rejection with :ailu e of Bypass
/l~ 8 Peak pressure at the valves reaches ~ psig.
The peak nuclear svstem pressure reaches ~ psig at the bottom of the vessel, well below the nuclea barr'e transient pressure limi of 1 375 ps',
/Zo 2 15.2.2.5 Rad'ological Conseauences While the conseauences of this event coes not esult in fuel failu es, i" does result in th discharge of normal coolant ac"'vitv to the supp ess'cn pool via SRV ope ation.
S'nce this activity is contained in the pr'mary containment, there will be no exposure to operating personnel.
Since this event coes not result in an uncontrolled release to the environ-
'ment, the plant operator can choose to leave the activity bott'ed up in the containment or d'scharge it to the environ-ment under optimum meteorologica3.
and release condit'ons.
Zf purcing of the conta'nment is chosen, the release will be in ac'ordance w'th establ'shed techn'cal spec'fications; the efore, this event., at worst, would on3.y esult in a small increase in the yea ly, integrated exposure level.
- 15. 2-8
HNP-2 TABLE 15.2-1 Page 1 of 2
S QUENCE OP EVENTS POR PIGURE 15.2-1 Time-sec Event
(-) 0. 015 (approx.
)
Turbine-generator detection of loss of elect=ical load.
Turbine-'generator power load unbalance (PLU) devices
" ip o in'tiate turbine control valve fast closure.
0 Turbine-generator PLU trip initiates main turbine bypass system operation.
Past cont. ol valve closure initiates scr am trip.
Past control valve closure in'iates a
reci culation pump trip (RPT).
- 0. 07 D. ll o.!9 Turbine control valves closed.
Turbine bypass valves start to open.
Rec'rculation pump motor circuit breake s open causing dec ease 'n core flow to natural circulation.
Group 1 relief valves actuated.
Group 2 relief valves actuated.
Group 3 relief valves actuated.
Group 4 relic valves actuated.
C
- 15. 2-9
TABLE 15.2-1 (Continuec)
Page 2 of 2
Event
<<JC J
'I m.
v C
~
15.2-10
TABLE 15.2-2 Page 1 of 2
SEQUENCE OF EVENTS FOR FIGURE 15.2-2 Time-sec Event
(-)0.015 (approx.)
Turbine-generator detect'on of loss of electrical load.
Turbine-generator power load unbalance (PLU) devices trip to initiate turbine control valve fast closure.
Turbine bypass valves fail to operate.
Fast control valve closure initiates scram trip.
Fast control valve closure initiates a recirculation pump trip (RPT).
0.07 0.:L-4-c./ I Turbine control valves closed.
Recirculation pump motor circuit'reakers open causing dec ease in core flow to natural circulation.
Group 1 relief valves actuated.
Group 2 relief valves actuated.
Group 3 relief valves actuated.
Group 4 relief valves actuated.
Group 5 relief valves actuated.
- 15. 2-11
TUEL=- 15.2 (Continue') Page 2 of 2
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15.2.3 TURBINE TRIP 15.2.3.1 Ident'cation of Causes and Preauency'lassification 15.2.3.1.1 Ident'fication of Causes A var'tv of tu b'e or nuclear svs tern malfunctions w'l
'n'tiate a " rhine t ip.
Some examples are moisture separa-to" and heate drain tank hich levels, large vibrations, operator lock out, loss of control =luid pressure, low con-denser vacuum and reactor hign water level.
15.2.3.1.2 Frecruency Class'ficat'on.
15.2.3.1.2.1 Turbine Trip Th's trans'nt is categorized as an 'ncident of moderate fre-ci~ency.
In def'ning the frecruency o" this event, turbine tr'ps wh'ch occur as a by-product o
other transients such as loss of concenser vacuum or reactor high level trip events are not incluced.
However, spur'ous low vacuum or high level tr'p signals which cause an unnecessary
'turbine trip are in-cluded in de ng the freauency.
In order to,get an accurate event-by-event freauency breakdown, this type of division or" in't'ting causes
's reauired.
15.2.3.1.2.2 Turbine Tr'p with Fa'lure of the Bypass I
Preauency is expected to be as follows:
gQpgp cf Preauency:
. 0064/plant year MTBE:
156 yea s
Precruency Basis:
As ciscussed in 15.2.2.1.2.2, Gene ator Load Rejection with Bypass Failure, the failure rate of the bypass is 0.0048.
Combin'g this with the turbine trip re-cruency o
1.33 events/pl'ant vear y'elds the freauency of
. 0064/plart vear.
- 15. 2. 3. 2 Secuence of Events and Systems Operation 15.2.3.2.1 Seauence of,Events 15.2.3.2.1.1 Turbine Trip
'I Tu bine "r'p at high power produces the secuence of events 1'sted 'n Table 15.2-3.
- 15. 2-15
0
I 15 ~ 2. 3. 2. 3. 2 Tubine Trips at Power Levels Less Than 30%
NBR Same as 15.2.3.2.3.1 except RPT and stop valve closure scram t ip is normally inoperative.
Since protection is still provided by high flux, high pressure, etc.,
these will cont'nue to function and scram the reactor should a s'ngle failu e occur.
15.2.3.3 Core and System Performance 15.2.3.3.1 Mathematical Model The compute model described in 15.3..1.3.1 was used to simu-late W~ +Ac.
+u~L ~c. +'9 u +4 J y~sz
<~+~'w ~ ~c
~4~~)
~ 3)
Suhsccki~d ls. I.z.3. f <4(
~ scJ
%a~ +z.
ur6'a~ +r:p
<+.m~
.q',]vrc.
o
~yp gg~
g vp<+a 15.2.3.3.2 Xnput Parameters and Initial Conditions These analyses have been performed, unless otherwise noted, with plant conditions tabulated in Table 15.0-2.
Turbine stop valves full stroke closur'e time is 0.1 second.
A reactor scram is initiated by posit'n switches on the stop valves when the valves are 908 open or less.
This stop valve sc am trip signal is automatically bypassed when the reactor 's below 30%
NB rated power level.
Reduction in core recirculat'on flow is initiated by pos'ion switches on the main stop valves, which actuate trip cir-cu'try which t ips the recirculat'on pumps.
15.2.3.3.3 Results 15.2.3.3.3.1 Turbine T ip
/
A turbine trip with the bypass system operating normally is simulated at 105%
NB rated steam low conditions in Figure 15.2-3.
Neutron lux increases rapidly because of the void reduction caused by the pressure increase.
However, the flux increase is 1'mited to 147.5% of rated by the stop valve scram and tne RPT system.
Peak fuel sur ace heat flux does not exceed 101.7$ of its initial value.
- 15. 2-18
15.2.3.3.3.2 Tu bine Tria with Failure of Bypass A turbine trip with ailure of the bvpass system is simulated at 105$
NB rated steam flow conditions in Figure 15.2-4.
zl8. 3 9 Peak neutron lux reaches ~~~ of i"s rated value and peak
~ -ue censer "empeZatye xpcrepses ppr xzm-e y/
ance th's even~ is cljssi"ied/as an in rea ent ncident, it 's no't limited dy se psTM crjrerik an th KCPi 1'm t i pe~itt6d I to fall,~beAw ""ae ~afetv lid.it for mcidynts 6f m erzke ~e-aueacv~
Ewwa"er, @De >CSR/for is Crane'as
>. is /l. 07/whj8h
>is/cyst aPiove the/safety P~zi for incidents'f moderate'fre-cdencv and the efor'=/s thd de~ign basi is sati8fieg.
15.2.3.3.3.3 Turb'ne Trip with Bypass Valve Failure, Low Power This transien='s less severe than a similar one at high powers Below 30% of rated power, the tu bine stop valve clo-sure and tu bine cont ol valve closure scrams are automati-cally bypassed.
At these lowe power levels, turbine first stace pressure is used to initiate the scram logic bypass.
The sc am which terminates the transient is initiated by high vessel pressure.
The bypass valves are assumed to fail; therefore, system pressure will inc ease until the pressure relief set points are reached.
At this time, because of the relatively low power of this transient event, relatively few relief valves w'll open to limit reactor pressure.
Peak pressu es are not expected to greatly exceed the pressure relief valve set po'nts and will be signi "icantly below the RCPB t ansient 1'm' of 1375 psig.
Peak surface heat flux
'nd peak uel center temperature remain at relatively low values and HCPR remains well above the GZTAB safety limit.
- 15. 2. 3. 3. 4 Considerations of Uncertainties Uncer ainties in these analyses involve protection system sett'ngs, svstem capacities, and system response characteris-t'cs.
In all cases, the mos conse vative values are used in the analvses.
=or exam le:
- 15. 2-19
a.
Slowest allowable cont ol rod sc am mot'on is assumed.
b.
Scram worth shape for all-rod-out concitions is assumed.
c.
Minimum specified valve capacities are utilized for overpressure protection.
15.2.3.4 Set points of the sa ety/relief valves include errorsy'(high) for all valves.
gPl~y+q',4+ g5, Barr'er Pe formance 15.2.3.4.1 Turbine Trip Peak pressure in the bottom oz the vessel reaches 1163 ps'g, which is below the BSHE code 1'mit of 1375 psig for the reacto cooling pressure boundary.
Vessel dome pressure does not exceed 1136 psig.
The severity oz turbine t i s from lower 'nit'al power levels decreases to the point wne e
a sc am can be avoided 'z auxiliary power is available f om an exte nal source and the power level is within the bypass capab'lity.
15.2.3.4.2 Turb'ne Trip vith r"ailure of the Bypass The safety/relief valves open and close sequentially as the stored ene gy is cissipated and the pressure falls below tne set points of the valves.
Peak nuclear system pressure reaches ~H psig at the vessel bottom, therefore, the over-pressure boundary transient pressure limit of 1375 psig.
Peak dome p essure does not exceed 53 psig.
~
I/7Q 15.2.3.4.2.1 urb'ne Trip with Failure of Bypass at Low Power Qualitative discussion is provided in 15.2.3.3.3.3.
15.2.3.5 Radiological Consequences While the consequence of this event does not result in fuel failure it does result in the discharge of normal coolant activity to the suppression pool via SRV operation.
Since 15.2-20
1 11 1i i
TABLE 15.2-4 Page 1 of 2
SEQUENCE OP rMENTS FOR FIGURE 15.2-4 Event Turbine trip initiates closure of main stop valves.
- 0. 01 Turbine bypass valv'es a'1 to operate.
Hain turbine stop valves reach 90%
open position and initiate reactor scram trip.
- 0. 01 Ma'n turbine stop valves reach 90%
open position and initiates a recircu-lation pump trip (RPT).
0.10 Tu bine stop valves closed.
~~ Q./f Rec'rculation pump motor circuit breakers open causing dec ease in core flow to natural circulation.
Group 1 reli'ef valves actuated.
Group 2 relief valves actuated.
Group 3 rel'ef valves actuated.
Group Group 4 relief valves actuated.
relief valves actuated.
O V' kJ j
15.2-24
TABLE 15.2 (Continued)
"-er.e 2 of 2
Event
'4 4
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~
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wgee pe 4 >e
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~
1 15.2-25
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n Im CÃJ In O
. mn m
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1
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p
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