ML20198J980
| ML20198J980 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/30/1986 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20198J977 | List: |
| References | |
| NUDOCS 8606030273 | |
| Download: ML20198J980 (71) | |
Text
_-.
i 5
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS e
2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or Low Flow t
2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER witi.
the reactor vessel steam done pressure less than 785 psig or core flow less than 10% of rated flow.
APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL P0WER and the reactor vessel steam done pressure less than 785 psig or core flow less than 10%
of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
Mb% b*
M *H tw o
THERMAL POWER, High Pressure and High Flow a
S4qll M b e le say icy I
UI& Sk% G WlaRon 19 oyitry
[2.1. 2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than H
v
( 1.06 withtthe reactor vessel steam done pressure greater than 785 psig and core flow greater than 105 of rated flow.
APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2.
I"55 ' b H I07-
- h( s;m y ggow 1..p oper=fi*n C*
ACTION:
I,. l p o p roib'e n With MCPR less than 1.06*and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE 2.1. 3 The reactor coolant system pressure, as measured in the reactor vessel steam done, shall not exceed 1325 psig.
APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.
I ACTION:
With the reactor coolant system pressure, as measured in the reactor vessel steam done, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
HOPE CREEK 2-1 8606030273 860530 DR ADOCK 0500 4
APR 1 1 1986
4 l
TABLE 2.2.1-1 m
g REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS N:
x ALLOWABLE FUNCTIONAL UNIT
- TRIP SETPOINT VALUES 1.
Intermediate Range Monitor, Neutron Flux-High 5 120/125 divisions
$ 122/125 divisions of full scale of fu11 scale 2.'
'A'v'erag'e Power Range Monitor:
s a.
Neutron Flux-Upscale, Setdown 5 15% of RATED THERMAL POWER
$ 20K of RATED THEMAL POER 4c.W(u-ou)+si/
o,g,(,(yg).g*I b.
Flow Blased Slaulated Thermal Power-Upscale
- 1) Flow Sia, sed q 0.55 "'5 5, with 5 4,46-W665, with i
a maximum of a maximum of
- 2) High Flow Clampec 1 113.5% of RATED
-< 115.5% of RATED THERMAL POWER.
TIERMAL. POWER
~
i A
FixedNeutronFluxpscale
-< 118K of RATED THERMAL POWER
< 120% of RATED c.
THERMAL POE R I
d.
Inoperative NA NA e.
Downscale 1 4% of RATED
-> 3% of RATED THERMAL POWER THERMAL POWER 3.
Reactor Vessel Steam Dome Pressure - High
$ 1037 psig 5 1057 psig 4.
Reactor Vessel Water Level - Low, Level 3 1 12.5 inches above instrument 1 11.0 inches above zero" instrument zero 5.
Main Steam Line Isolation Valve - Closure 5 8% closed 5 12% closed 6.
Main Steam Line Radiation - High - High
$ 3.0 x fu11 power background 5 3.6 x full power background
]
E (h,
- See Bases Figure 8 3/4 3-1.
h67.w ey mWM" !"Jo S
b Scm% N 6an vocies o w is 3 e b e.,J os 2.e d4bw.4e.iw i.Ao d c b N (in p w,.J Jria b vM pn&-5 f*M
- 6. k A E8 de w % core N lap %g sty @l< bog oaemnon o*r,.,M.,, M "f?"?y oo'sX b si f /-f o/'*A'7 ea 42)
I.e+w m
+w-40=0 m,..
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2.1 SAFETY LIMITS SASES
- 2. 0 INTRODUCTION The fuel cladding, reactor pressure vessel and primary systes piping l
}
are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these g
barriers during normal plant operations and anticipated transients. The fuel
.s.,
cladding integrity Safety Limit is set such that no fuel damage is calculated j,*_
to occur if the limit is not violated. Because fuel damage is not directly g
observable, a step-back approach is used to establish a Safety Limit such that 2
5[' g the MCPR is not less than h.06 d MCPR erseter tian 1.06 pepresents a con-ervative margin relative 1.0 we conditions req)f ree w maintain fuel cladding 4.
integrity. The fuel. cladding is one of the physical barriers which separate g
o the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom free perforations or cracking.
gg Although some corrosion or use related cracking may occur during the life of l
'O g the cladding, fission product migration from this source is incrementally i
s cumulative and continuously sensurable. Fuel cladding perforations, however, can result from thermal stresses which occur free reactor operation signifi-cantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related. cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.
These conditions represent a signi-ficant departure from the condition intended by design for planned operation.
2.1.1 THERMAL POWER. Low Pressure or Low Flow The use of the GEXL correlation i.s not valid for all critical power calculations at pressures below 785 psig or core flows less than 10K of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will al s be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 1 lbs/hr, bundle pressure drop is nearly independent of bundle I
l i
power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x los 1bs/hr. Full scale ATLAS test data taken at pressures free'14.7 psia to 800 psia indicate that the fuel assembly criti-cal power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL PCWER of. sore than 50K of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 2iX of RATED THERMAL POWER for reactor pressure below 785lpsig is conservative.
i HOPE CREEK B 2-1 i
APR 1 1 1986
I i
i BASES TABLE B2.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT
- STANDARD DEVIATION QUANTITY
(% OF POINT)
Feedwater Flow 1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow Two Recirculation Loop Operation 2.5 Single Recirculation Loop Operation 6.0 Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5.0 TIP Readings Two Recirculation Loop Operation 6.3 Single Recirculation Loop Operation 6.8 R Factor 1.5 Critical Power 3.6 The uncertainty analysis tsed to establish the core 3
wide Safety Limit MCPR is based on the assumption of quadrant powe,r symmetry for the reactor core.
The valves herein apply to both two recirculation loop operation and single recirculation loop operation, except as noted.
e 4
HOPE CREEK B 2-3 1
l 4
3/4.2 POWER DISTRISilTION LIMITS 3/4.2.1 A^.
PtANAR LI Eaa HEAT gen ~iRATION RATE
~
LIMITING ColdITIINI FOR OPERATION AllAVEREGEPLANARLINEARHEATGENERATIhNRATES(APLHGRs)foreachtype 3.2.1 of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5.f APPLICAll!LITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equa' to 255 of RATED THENIAL P0hER.
ACTION:
With an APLNGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, t
3.2.1-4, or 3.2.1-5, initiate corrective action within 15 minutes and restore APLHGR to within the required Itaits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 255 of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
(yQ by;$$ of Fque S Q l-lj.5 3'l'%) S W ~3s SO'I~Y"<I'*'
' l ~'^ '
tecI*I*
l'* *)**NN y & ace <Q tw % valve ah ev. Bio H%% c$t.
sisy/n tectreeleNem hop
- mjwegm,
,;,,,;g w 4.a tu o
i SURVEILLANCE REQUIREMENTS l
I 4.2.1 All APLHGRs shall be verified to be equal to or less than the Itaits detamined from Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 and 3.2.1-5:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of__at least 155 of RATED THERMAL POWER, and c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL A00 PATTERN for AP'.HGR.
I d.
The provisions of Specification 4.0.4 are not applicable.
^
Y I
HOPE CREEK 3/4 2-1 l
~AeR 1 1 $66 1
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POWER DISTRIBUTION LIMITS 3/4.2.2 APSF SETPOINTS LIMITING C0fSIT10N FOR OPERATION 3.2.2 The APRM, flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux upscale control rod block trip setpoint (S RB) shall be established according to the following relationships:
TRIP SETPOINT ALLOWABLE VALUE w.e. d.41
~
S $ (0.6 1%)T S 5 (0.6 54%)T Q.fW Mr Sgg 5 (O.
+ 42%)T Sgg 1 (O.
+ 45%
where: S and S are in percent o ED THE
- OWER, g
W = Loo recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million Ibs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER (FATP) divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD). T is applied only if less than or equal to 1.0.
APPLICA8ILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or S andad$s,t5and/or5, gas above determined, initiate corrective action within 15 min to be consistent with the Trip Setpoint values
- within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS l
4.2.2 The FRTP and the CMFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, I
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at b.
i least ISK of RATED THERMAL POWER, and Jnitially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating c.
with CMFLPD greater than or equal to FRTP.
d.
The provisfons of Specification 4.0.4 are not applicable.
"With CMFLPD greater than the FRTP " ~ ~
7 rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times CMFLPD provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.
fy See h*H. (S)' " 7%k a.2.s-1 HOPE CREEK 3/4 2-7 AP
'. *. GSS i
4
~
kce duich {
L _ 3.3.6-2 CONIROL R00 BLOCK INSTRUMENTATION SETPOINTS k
1 RIP FUNCIl0N TRIP SEIPOINT ALLOWABLE VALUE
[
1.
R00 BLOCK MONIIOR g
a.
Upscale
< 0.6 W 40%
< 0.66
+ 43%
p b.
Inoperative RA RA c.
Downscate 1
of RATED THERMAL POWER 1 3% of RATED THERMAL POWER 2.
APRM a.
Flow Biased Neutron Flux -
r
- * ' Ogiscale
< 0.
W + 42%*
< 0.'6 d
45%*
b.
Inoperative RA RA 1
c.
Downscale
> 4% of RATED THERMAL POWER
> 3% of RATED THERMAL POWER I
d'.
Neutron Flux - Upscale, Startup 312%ofRATEDTHERMALPOWER 314% of RATED THERMAL POWER 3.
SOURCE RANGE MONITORS l
a.
Detector not fuTT in NA NA 5
5 b.
Upscale
< 1.0 x 10 cps
< 1.6 x 10 cp, i
c.
Inoperative HA RA d.
Downscale 1 3 cps **
1 1.8 cps i
4.
?
U NA a.
Detector not full in NA b.
Upscale
< 108/125 divisions of
< 110/125 divisions of Y
Tull scale Tull scale I
E c.
Inoperative NA NA l
d.
Downscale
> 5/125 divisions of
> 3/125 divisions of i
Tull scale Tull scale l
S.
Water Level-High (Float Switch) 109'1" (North Volume) 109'3" (North Volume) i 108'11.5" (South Volume) 109'1.5" (South Volume) 1 1
6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.
Upscale
< 108% of rated flow
< 111% of rated flow b.
Inoperative HA RA c.
Comparator
< 10% flow deviation 1 11% flow deviation 7.
REACTOR MODE SWITCH SHUTOOWN POSITION NA NA c15o "The "c r u- *~ r-D : "^ i'M rod block ction is varied as a function of recirculation loop flow (W)j The trip setting ofg%46 function must e maintained in accordance with Specification 3.2.2.
May e reduced to 0.7 cps provided the signal-to-noise ratio is 12 k & igifgl C.% huoli
- 8 j
6 0 W l s. M in M k v
V iw w. he leh ma e)
Tk Arrup f=* h 'Ibi'*" bd BhE-c5 T.Lle. 2.;t.1-I,
s
.._. a =
. q.
P 3/4.4 REACTCR CDClaNT 5757EM
' 3/a.a.1 RECIRC'JLLTION 575TEP.
. ' RECIRCULATION LOOPS LIMITING CONDITION FOR 0?!KATION 3.1.1.1 Tw: react:r c:elant systa rt:irculati:n 1ceps shall be in cperation d th:
- z..
Tetz1 c:re f1cw gretter tat.. cr eRual t: 45 'cf rated c:ra fiew, c-L.
THEWAL PCVER ykth!'r the vere shneba' zent * / FYo s. c 3.4 I 1-1.
I..~
A; LIC:E LITY: CFERATICSAL C :'CITICH3 l' ar.d 2".
A TICN:
With cne res:::r c:clant systen re:ir:ulatten Ic:;: net in c;eratien:
)
a.
1.
Within 4 h:urs:
a)
F;a:t t.'e rztir:ulati:r ficw c: tr:1 systa: in the L:c2l l
Paws 1 4mede, and b)
Reduct THER"AL POW.R t: 170 % cf RATO THERMAL P.7.R and.
c)
In:rette the M*C.NM CRITICAL PCWER RAT!Q (MCFR) Safety p
Limit by 0.01 t: 1.07 per Specifica.icn 2.1.2, and, d)
ReSuce the Maxt u Average Planar Linear Heat Generatien Rate (PAPLMGR) lief t t a value of c.4/ times the t-o re,circulati:n le:p c;eration limit per' Specificati:n 3.2.1,
- and, e)
Reduce the Ave. age Power Range Menitor (AFRM) Scra: and R:d 81cck and lled BlulA.dter Trip' Setpcints and Allow-3 able Values to these a;plicable.for sirgle recirculatten Icep operatica per 5pecificatiens 2.2.1, 3.2.2, and 3.3.6.
,L,i.,.t H < sped a / He.p/ rdJ pq sp l, a draby recireu f) fha, se eguals.1o*4
- see Special Test Exception 3.10.4..
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N e
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-.. ::::; _: = --_ -
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R s
h ION (Centinuedl
_1.1MIT MG CONDITION FOR OPERAT t.cTION:
(continued) g)
Perfo'r= Surveillan:e Requirement 4.4.1.1.2 f f THERMAL POVER 1
is < 30'."a of RATED THERW.AL POVER or the recirculation 1 cop floi in the operating Icep is < 5C7."" of rated Icep flow.
s o
i
- 2. Tn p. ev,* sis.'s el Spesil,eal. con 3 0 a re sto l o p p loi le..
- 3. b tf.e r is. L e ir, s & l<* st Hor :>rerpcJH :,,:.'i ;, i.:e,,,,g. ; 2 4,ars,
- b. L.';rle,. o r en t.r a,,.le ! ey sia.- re circ ula.ba,. l..ps i., eyera b,,,, ;.a do.kly b e r e du c e To.'H:m p o s s a e.u c l, 6.* e.t, l
,*.s p..! w,'t h S A e
}
in ol,<.:,e a eH* n re el e*d el z o s e o f f,g ue, 3.u. t. s - t
- .% 2 S we
- , a n el,,,;t,, te,,n carsa.,,1,,,
he pled 4 : e u sis i., n a fe.as t 2 7elpTW witi..., & hau. :
LA:,n b l, c n e s h (o h o a,' c.
S f ** bt f l* b *
SYo j
U cneof WithAtwo reactor coolant systa jrs:irculatica Iccps in operatten and total core flew less than 4EYof rated c:re ficw and THERMAL PCWER c.
hAere<6tride.( zooe e#
- Figure 3.4.1.1-1
withid Determine the APRM and LPR.Y" ncise 1e tels (Surve'111ance 4.4.1.1.I,.):
1.
a)
At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and Within 30 minutes after the c::pletien of a THEt'.AL power b) increase of at least,5". of RATED THERMAL POWER.
With the APRM or LPRM" neutron flux noise levels greater than three times their established baseline noise levels, witb4 ff,.davtesi 2.
~
initiate corrective action to restore the noise levels within
\\
the. required limits within 2 h urs by increasing core flow j or by r,, educing THERMAt. PCWER.
ci j
Q,y.gir
=
- " Detector levels 'A an'd C of one LPRM string per core octant pl.us detecters A and C of one LPRM string in the center of the core should be monitored.
Final values to be deter ined during Startin Testing based
- "' Initial values.
upon the thresheid THERMAL PCWER.a.nd recirculation loep flow wat.:h will y
sweep the cold water frcm the ve,ssel bottom head preventing stratification.
- U lu' U 'o~ k. e
- t 5 ~r b lh d e d A vr %. $ w es.u p
,p,cp,,,.(C,le/4w{,
bat 4
'~
ree eec I. t,.,, p.,
,a,..,,,,, o,,,, p,,,p, g M y-2.
w N.
- l 2 U n
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~,l,.........,,_,
. ~. ~ ~. -..... _
~- - - -
-. 6 L.,...,. 4
- - - - -. ~..
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+
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f g-1 CR CX'L1NT SYSTU'
., x
' _SURVEll, LANCE REQUIREMENTS w3 m
1 bs With one resciar c:sclant syste: retirculation loep net in operdion,j 1
4.4.1.1.1 at least once per 12 heurs verify that:
s Res:ter THEFAL PC' AIR is (7c: cf RAID THEFAL PChTR, c.
/
The recir:ulation flow c ni'01 !.ystas is in the Local Manual b.
t
- e::de, ar.h
~-
'J,q rt g.. r.
i 15a 90h' The spd) posf +ed.of He ofero6'j recorcubbott f.A 3 is Ienf na I,
t c.
e t* r d e systa. re:ir:ulaticn Icep net in c;eratien, L
~
s U.,;in3.7. k'ita cr.e res:::r c::la..:
4*i,
ei..e'r THERuAL FCw!R in:. ease or rq:ir-l!
c r.:rt th:. 1E ninetes prier :
in:reste, verify tnat tne f: lie ing cif'eren {al tea.;erature, ;
-ulatica Icep ficw cf RATED THER*AL PChTR'er the i
re:uirements are e:e* if THE?.uAL PCwE; is i 3t%*-in tne cperatir.; r :f r:ulaticn le:p of rated redr:ulatten ic:p ficw
}
lo:p f*ow:
< mi *F betnetn rascur vasta*. s. san spaci c: lant and b % :m hea:
]
~
a.
crain line c:eiant, li
< u*F det,een t.e rytcur c::W. Minin tne let: net,in c;e*atics 3
~-
p t.
ine tne c::la.t in t. a ret:::r pat ters vessel, anc V
rea:*.:t c::la.; wi.in tr.e ic:p n:: in o=eratien fi
< EG*F be een tr.
c.
a.c ine c; erat,ing 10:;.
a.A.l.1.25. and c.
g Tne diarentini in;e ssure ri uire-ms of S:ect'icatien a
ce net a;cly when the lecs n in c;e-nien is isolaue f t:m the recur i
pressure vessel.
- 4. 4.1.1. 3 Each pomp HG set se' cop tube mechanical and electrical step shall be demonstrated CPERASLE with overspeed setpoints less than or equal to 105% and 102.5%, re,spectively, of rated core flow, at least once per 18 months, g
neutron flux noise value within 4.4.1.1.4.
Establish a baseline APRM 'and LPRMaa I
the regions for which monitoring is required (Specification 3.4.1.1, ' ACTION c)
I i
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering the region for which monitoring is required unless l
baselining has previously been perforr.ed in the region since the last refueling.'
t outage.
e
(
I t 0:
- tsbdhs ed cloo'ry %mp-fag,.am_ (a,, jj,, an g,g,
]
I e a an pump a t,..,6 u,,, pe p g,o ed).
7 c
Final values to be deter =ined during startup Testing tassif
^
ill l
upon the tnreshold THEf>L PCkIR and recir:ulaticn Initial values.
i
(~'
~
y "Datector levels A and C of one LPRM string per core cctant plus detectors A
/ and C of one LPRM string in the center of the core shculd be ronitored.
di h
- =w J
o.
=..
a g=, ; *
.~.
INSERT A With one or two reactor coolant system recirculation loops in operation and total core flow less than or equal to 39%# and THERMAL POWER within the restricted zone of Figure 3.4.1.1-1, within 15 minutes initiate corrective action to reduce THERMAL POWER to within the unrestricted zone of Figure 3.4.1.1-1 or increase core flow to greater than 39%# within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
INSERT B d.
Core flow is greater than 39%# when THERMAL POWER is within the restricted zone of Figure 3.4.1.1-1.
W 8
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REACTOR C00LANT $YSTEM i
JET PtmPS LIMITING COW ITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERA 8LE.
APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With one or more jet pimps inoperable, be in at least HOT SHUTOOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS *
- c. s. 2.
A ll :yet p m ps, :9igli be Seecm'e CPcAA h F as SIN.
0.0.1.2-Each of the above required jet pumps shall be demonstrated OPERA 8LE Q.
prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by determining recirculation loop flow, total core flow and diffuser-to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur when the recirculation pumps are operating in accordance with Specification 3.4.1.3.
l,[ The indicated recirculation loop flow differs by more than 10% from the established pump speed-loop flow characteristics.
((
The indicated total core flow differs by more than 10% from the established total core flow value derived from recirculation loop flow measurements.
3/
The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from the established patterns by more than 105.
b, During single recir:ulation loep operation, each of the above req,viteJjec, pumps shall be demonstrated OPERA!LE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> b that no two of the following conditions occur:
l l
l.
The indicated recirculation loep f1 the operating loop differs by more than IC: from the establishe
- / speed /..f /4 s cWecler/36Er..
l O
"During tne startup test program, data shall be recorded for the parameters Itsted to provide a basis for establishing the specified relationships.
Comparisons of the act'ual data in accordance with the critaria listed shall commence upon conclusion of the startup test program.
HOPE CREEK 3/4 4-4 APR 1 1 1988
,.7
h I
O I 4.f,2, b. _ (cM6'
~
- 2.
- The indicated total core flow differs by more than 15 from the estab-lishetMatal core flow value derived from single recirculation loop flow sensurements.
3 The indicated difference-ta-lower plenus diffgantial pressure of any individual jet pue:p differs from establishetf'Ungle recirculation
, loop patterns by a, ore than 13.
r:visier.s of Spe:ificzti== 4.0.4 are n:t a:;1icable provided
.[t this surveillanca is perfer d within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after ex:ending 2J of g;,;E3 Tiii.:iu/,1. PCWIK.
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i
.c ACTOR COOLANT SYSTEM RE:IRCULATION LOOP FLC'.*
DITION FOR OPERATION
_IuTTIN; CON i d within:
3.4.1.3 ' Re:ir:ulation le:p flew mismat:5 shall be mainta 5% cf rated cere ficJ wi 1:7C: cf rated c:re ficw.
I';; E E *
- IC II* d iYh *?~"~*!I*C CD'* ll*'
1 % cf rated Cc!t b.
rated c:re ficw.
10:7 OFE:.ATIONAL C3NOITIONS 1** and I"during two re:ircu FFLIC 3!!!U:
i:;: era :en.
i
- limits,
{
with the re:ir:ulatien le:; fiews diffe e..i by ::re.than the spe:if A~ TION:
eithe.:
witr.in the spe:ified' limit Rest:re the re:te:ulatien 1::; fiews :
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=
e SURVEILLANCE REOUIREVENTS i hin the Re:ircula' tion'loep flew mismat:5 s..all be verified to be w t f
~
4.4.1.3 limits at least once per 24'heu-s, l
i L
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REACTIVITY CONTROL SYSTEMS 8ASES 3/4.1.3 CONTROL R005 The specifications of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) limit the potential effects of the rod crop accident. The ACTION statements permit variations from the basic requirements but' at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
Damage within the control rod drive mechanism could be a generic problem, therefore with a withdrawn control rod immovable because of excessive friction or mecnanical interference, operation of the reactor is limited to a time perioc which is reasonable to determine. the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.
Control rods that are inoperable for other reasons are permitted to be 1
taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.
1 The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.
4.g,j cl J I g8e
- hge, The control red system is designed to bring the actor ubcritical at a rate fast enough to prevent the MCPR from becoming less han during the limiting power transient analyzed in Section 15.4 of the FSAR. This analysis shows that the negative reactivity rates resulting the scram with the i
average response of all the drives as given in th specifications, provide the required protection and MCPR remains greater than The occurrence of scram times longer then those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval l
is reduced in order to prevent operation of the reactor for long periods of l
time with a pot'entially serious problem.
(
I The scram discharge volume is' required to be OPERABLE so that it will be available when needed to accept Bischarge water from the control rods during a y
reactor scram and will isolate the reactor coolant system from the containment when required.
Control rods with inoperable accumulators are declared inop able and i
Specification 3.1.3.1 then applies. This prevents a pattern of inoperable l
accumulators that would result in less reactivity insertion on a scram than I
has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the i
I accumulator ensurek that there is a means available to insert the control rods l
even under the most unfavorable depressurization of the reactor.
HOPE-CREEK 8 3/4 1-2 APR 1 1 1996
1 l
POWER OISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) b.
Model Chan'oe 1.
Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.
2.
Incoporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOO pressure transfer when the pressure is increasing was changed.
A few of the changes affect the accident calculation irrespective of CCFL.
These changes are listed below.
a.
Input Chance 1.
Break Areas - The D8A break area was calculated more accurately.
b.
Model Chance 1.
Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation.
A list of the significant plant input parameters to the loss-of-coolant 1
A accident analysis is presented in Bases Table B 3.2.1 $1e // YAP.fG pbd freHow wl4 %9/eMl-is 3 2./-2; 3 2. I-3, 3 3.t-9 med M.p tx/rcuNicH /w he
- Qnt S s~f we h
O.3b y be( cd M
mit 3/d2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at TED THERMAL POWER. - The flow biased simulated thermal power-upscale sc as setting and the flow biased neutron flux-upscale control rod block trip ints must be adjusted to ensure that the MCPR does not become less than&B6 or that > 1%
I plastic strain does not occur fit the degraded situation. The scram setpoints 3
and rod block setpoints are adjusted in accordance with the formula in Specifi-cation 3.2.2 whenever it is known that the existing power distribution would cause the design'LHGR to be exceeded at RATED TNERMAL POWER.
Nc. Cees %f w(py-O.M I$
b s % M T "; 2 T I N S,,
7 c+ M LgcA emIo,m3,b HOPE CREEX 8 3/4 2-2 APR 11 1986
I e
3/4.4 REACTOR COOLANT SYSTEM SASES
.[
3/4.4.1 RECIRCULATION $YSTEM M yrt. D k
Oy... i T.,
.6: rr ~
culati s
p prohibit ntil valuation o erfo
... i.. t -
loop ned to be se(e'ptable.
opera n has n performed. a
-t:
e Q
An inopera'le jet pump is not, in itself, a sufficient reason to declare b
oj a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.
d Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.
-b d
ecirculation n un speed mismatch limits are in compliance with the ECCS J
LOCA analysis design criteri b The limits will ensure an adequate core flow f.
coastdown from either recirculation loop following a LOCA. rmer eF 3
In order to prevent undue stress on the vessel nozzles and bottom head p
region, the recirculation loop temperatures shall be within 50*F of each other a
iprior to startup of an idle loop. The loop temperature must also be within
/f j 50*F of the reactor pressure vessel coolant temperature to neavent therma 1EnSerb
%./ shock to the recirculation pump and recirculation noules.6 ini.. Um ei ni, in
- {.. httg vi i{.....i is ei e i-,1
, _r.suru Umn it c.ehnt h tb g-p='-
......,. - -- -- a s rw 3 3 vn Unw... wi wouiu resuis if sne i y.r.;. m O "e-Me m g _-ter tM.1","F.
~
The objective of GE BWR plant and fuel design is to provide stable operation I
with margin over the normal operating domain. However, at the high power / low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g.,
rod pattern, power shape). To provide assurance that neutron flux limit cycle e.111ations are detected and suppressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region.
1 Stability tests at operating BWRs were reviewed to determine a generic region of the power / flow map in which surveillance of neutron flux noise levels should be performed. A conservation decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs. This generic region I
has been determined to correspond to a core flow of less than or equal to 45Y. of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.,
Plant specific calculations can be performed to determine an applicable region for monitoHng neutron flux noise levels.
In this case the degree of conservatise can be reduced since plant to plant variability would be eliminated.
In this case, adequate. margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.
9 HOPE CREEK B 3/4 4-1 APR 1 1 1986
[..
La
- +
g
~ INSERT E
/
The impact of single recirculation loop operation upon-plant safety is assessed and shows that single-loop operation
'is permitted if~the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in. Tables 2.2.1-1 and 3.3.6-2, respectively.
MAPLHGR limits are decreased by the factor given in Specification 3.2.1, and MCPR operating limits are adjusted per section 3/4.2.3.
Additionally, surveillance on the pump speed of the operating
-recirculation loop is imposed to exclude the possibility of excessive core intervals vibration.
The surveillance on differential temperatures below (30*F)* THERMAL POWER or (50*F)* rated recirculation loop flow is to mitigate the undue thermal Stress on vessel nozzles, recirculating
. pump and vessel bottom head during the extended operation of the single recirculation loop mode.
INSERT F In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single. recirculation loop mode.
INSERT G Sudden equalization of a' temperature. difference > 145*F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.
l^ b
- Initial values; the final values are determined during startup testiny based upon the threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from thejvessel bottomhead, preventing saturation.
t i
l l
I s
I.
d I
o w
Bases Table 8 3.2.1-1 h
SIGNIFICANT INPUT PARAMETERS TO THE I
- 1 LOSS-OF-COOLANT ACCIDENT ANALYSIS I
Plant Parameters:
Core THERMAL-POWER.................... 3430 Mwt* which corresponds to 105% of rated steam flow Vessel Steam Output................... 14.87 x 108 lbe/hr which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure.............
1055 psia
~
Design Basis Recirculation Line 8reak Area for:
a.
Large Breaks 4.1 ft8 2
b.
Small Breaks 0.09 ft,
Fuel Parameters:
PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNOLE GENERATION RATE PEAKING POWER i
FUEL TYPE GE0 METRY (kw/ft)
FACTOR RATIO Initial Core 8x8 13.4 1.4 1.20N A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 6.3.3 of the FSAR.
- This power level meets the Appendix K requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE. limit.
N k br sigle rec:whMon opm Hcw, joss ch hvcbH2 i
4 is Qhed of o l. s.eccac/ ob LOCA 9,o/le3s c) wad Mc?c f.
F' 8 3/4 2-3 HOPE CREEK APR 1 i 596
,c.._..
,m
.r POWER DISTRIOUTION LIMIT 5 OASES 3/4.2.3 MINIMM CRITICAL POWER RATIO The required operating 11af t MCPRs at steady state operating. conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR s% and an' analysis of abnomal operational transients. For any abnormal operating transient analysis evalua-tion with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit.MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.
To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational tra.1sient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
't iI tti-t :--f xt y'e!* +' 1==?+ dalta '
/"*"
Sn - f fl i. 6ne 5.iesy Limii r i.;; M 1.^', th g_!M
-iai="-
ce--t! ;; !!:it ;;;i; vi Liiis iivu 5.2.5 i.
i.l.J.
The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-3 that are input to a GE-core dynamic behavior transient computer progr g The code used to evaluate pressurization events is described in NE00-24154 and the program used in non pressurization events is described in NED0-10802(2)
The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundlewitht%singlechanneltransientthermalhydraulicTASCcodedescribed in NEDE-25149 The principal result of this evaluation is the reduction in MCPR caused by the transient.
The purpose of the K factor of Figure 3.2.3-2 is to define operating limitsatotherthanrate$coreflowconditions". At less than 1005 of rated flowtherequiredMCPRistheproductoftheMCPRandtheK{edduringaflow factor. The K factors assure that the Safety Limit MCPR will not be viola increase transient resulting from a motor generator speed control failure.
The K factors any be applied to both manual and automatic flow control modes.
g The K, factors values shown in Figure 3.2.3-2 were developed generically and are ap611 cable to all BWR/2 8WR/3 and BWR/4 reactors. The K factors were l
derivedusingtheflowcontrollinecorrespondingtoRATEDTHERMA[POWERat i
rated core, flow.
For the manual flow control mode, the K factors were calculated such that f
for the maxion flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POWER along the rated flow control line, the Itaiting bundle's relative, power was adjusted until the MCPR changes with different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the K.
f 8 3/4 2-4 HOPE CREEK APR 1 1 1986 l
C:
i 9
ATTACHMENT II a>
t J
e e
9 0
f o
s MDE 15-0186 DRF No. A00-02541 HOPE CREEK SINGLE LOOP OPERATION ANALYSIS FEBRUARY 1986 Prepared for PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK GENERATING STATION 4>
Prepared by GENERAL ELECTRIC COMPANY
}
liUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA 95125 p.
O e
o
- 8 HCGS FSAR APPENDIX 15.C TABLE OF CONTENTS P_ age 15.C RECIRCULATION SYSTEM SINGLE-LOOP OPERATION 15.C.1-1 15.C.1 INTRODUCTION AND
SUMMARY
15.C.1-1 15.C.2 MCPR FUEL CLADDING INTEGRIT) 1AFETY LIMIT 15.C.2-1 15.C.2.1 Core Flow Uncertainty 15.C.2-1 15.C.2.1.1 Core Flow Measurement During Single-Loop Operation 15.C.2-1 15.C.2.1.2 Core Flow Uncertainty Analysis 15.C.2-2 15.C.2.2 TIP Reading Uncertainty 15.C.2-4 15.C.3 MCPR OPERATING LIMIT 15.C.3-1 15.C.3.1 Abnormal Operational Transients 15.C.3-1 15.C.3.1.1 Feedwater Controller Failure - Maximum Demand 15.C.3-2 15.C.3.1.2 Generator Load Rejection With Bypass Failure 15.C.3-3 15.C.3.1.3 Summary and Conclusions 15.C.3-5 15.C.3.2 Rod Withdrawal Error 15.C.3-6 15.C.3.3 Operating MCPR Limit 15.C.3-7 15.C.4 STABILITY ANALYSIS 15.C.4-1 15.C.4.1 Phenomena 15.C.4-1 3
15.C.4.2 Compliance to Stability Criteria 15.C.4-2 15.C.5 LOSS-0F,-COOLANT ACCIDENT ANALYSIS 15.C.5-1 15.C.S.1 Brea)SpectrumAnalysis 15.C.5-2 15.C.S.2 Single-Loop MAPLHGR Determination 15.C.5-2
- 15. C. S.'3 Small Break Peak Cladding Temperature 15.C.5-3 15.C-i
0 0 -
l HCGS FSAR TABLE OF CONTENTS (Continued)
P_ag 15.C.6 CONTAINMENT ANALYSIS 15.C.6-1 15.C.7 MISCELL NE0US IMPACT EVALUATION 15.C.7-1 15.C.7.1 Anticipated Transient Without Scram Impact Analysis 15.C.7-1 15.C.7.2 Fuel Mechanical Performance 15.C.7-1 15.C.7.3 Vessel Internal Vibratio'n 15.C.7-2 15.C.8 REFERENCES' 15.C.8-1 i
i 9
9 i
15.C-ii
(..
8-HCGS FSAR LIST OF TABLES NUMBER TITLE PAGE 15.C.3-1 Input Parameters and Initial Conditions 15.C.3-9 15.C.3-2 Sequence of Events for Figure 15.C.3-1, Feedwater Controller Failure, Maximum Demand 15.C.3-11 15.C.3-3 Sequence of Events for Figure 15.C.3-2, Generator Load Rejection with Bypass Failure 15.C.3-12 15.C.3-4 Summary of Transient Peak Value and CPR Results 15.C.3-13 e
f 9
15.C-iii
- ~
LIST OF FIGURES NUMBER TITLE PAGE 15.C.2-1 Illustration of Single Recirculation L'oop Operation Flows 15.C.2-5 15.C.3-1 Feedwater Controller Failure - Maximum Demand, 75% Power /60% Core Flow 15.C.3-14,15, 16,17 15.C.3.2 Generator Load Rejection with Bypass Failure, 75% Power /60% Core Flow 15.C.3-18,19 20,21 15.C.5-1 Uncovered Time vs. Break Area - DC Power Source Failure 15.C.5-4 l
I 15.C-iv
(
il HCGS FSAR 15.C RECIRCULATION SYSTEMS SINGLE-LOOP OPERATION 15.C.1 INTRODUCTION AND
SUMMARY
Single-loop operation (SLO) at reduced power is highly desirable in the event recirculation pump or other component maintenance renders one loop inoperative. To justify single-loop operation, accidents and abnormal operational transients associated with power operations, as presented in Sections 6.2 and 6.3 and the main text of Chapter 15.0, were reviewed for the single-loop case with only one pump in operation. This appendix pre-sents the results of the safety evaluation for the operation of the Hope Creek Generating Station (HCGS) with single recirculation loop inoperable.
This safety evaluation was performed for both GE-6 and GE-7 fueled cores.
The transient safety analysis was performed on an initial cycle basis consistent with that for the FSAR. The analysis shows that the transient consequences for SLO (t.CPR) are bounded by the full power analysis results given in the FSAR. The conclusion drawn from the transient analysis results presented in this report is applicable to reload cycle operation as well as initial cycle operation for HCGS. The conditions are those of continued cperation in the operating domain currently defined in Figure 4.4-6 of Chapter 4 up to maximum power of approximately 70% of rated.
Increased uncertainties in the core total flow and Traversing In-Core Probe (TIP) readings resulted in a 0.01 incremental increase in the Minimum Critical Power Ratio (MCPR) fuel. cladding integrity safety limit during 3
single-loop operation. No increase in rated MCPR operating limit and no change in the flow dependent MCPR limit is required because all abnormal operational transi,ents analyzed for single-loop operation indicated that there is more tharf enough MCPR margin to compensate for this increase in MCPR safety limit.I The recirculation flow rate dependent rod block and scram setpoint equation given in Chapter 16 (Technical Specifications) are adjJsted for one-pump operation.
15.C.1-1
b i
8 HCGS FSAR Thermal-hydraulic stability was evaluated for its adequacy with respect to General Design Criteria 12 (10CFR50, Appendix A).
It is shown that this stability criterion is satisfied during' SLO. It is further shown that the increase in neutron noise observed during SLO is independent of system stability margin.
To prevent potential control oscillations from occurring in the recir-culation flow control system, the operation mode of the recirculation flow control system must be restricted to operation in the manual control mode for single-loop operation.
The limiting Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) reduction factor for single-loop operation is calculated to be 0.86.
The impact of single-loop operation on the FSAR specifications for containment response including the containment dynamic loads was evaluated.
It was confirmed that the containment response under SLO is within the present design values.
The impact of single-loop operation on the Anticipated Transient Without Scram (ATWS) analysis was evaluated.
It is found that all ATWS acceptance criteria are met during SLO.
The fuel thermal and mechanical duty for transient events occurring during SLO is found to be bounded by the fuel design bases. The Average Power Range Monitor (APRM) fluctuation should not exceed a flux amplitude of
^
215% of rated and the core plate differential pressure fluctuation should i
not exceed 3.2 psi peak to peak to be consistent with the fuel rod and assembly design bases.
=
A recirculatipn pump drive flow limit is imposed for SLO. The highest drive flow that me'ts acceptable vessel internal vibration criteria is the e
drive flow limit for SLO. The pump speed at Hope Creek Generating Station.
should be limited to 90% of rated under single-loop operating conditions.
i 15.C.1-2
3 HCGS FSAR 15.C.2 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT Except for, core total flow and TIP reading, the uncertainties used in the statistical analysis to determine the MCPR fuel cladding integrity safety limit are not dependent on whether coolant flow is provided by one or two recirculation pumps. Uncertainties used in the two-loop operation analysis are documented in the FSAR. A 6% core flow measurement uncertainty has been established for single-loop operation (compared to 2.5% for two-loop operation). As shown below, this value conservatively reflects the one standard deviation (one sigma) accuracy of the core flow measurement system documented in Reference 15.C.8-1.
The random noise component of the TIP reading uncertainty was revised for single recirculation loop operation to reflect the operating plant test results given in Subsection 15.C.2.2.
This revision resulted in a single-loop operation process computer effective TIP uncertainty of 6.8% of initial cores and 9.1% for reload cores. Comparable two-loop process computer uncertainty values are o.3% for initial cores and 8.7% for reload cores. The net effect of these two revised uncertainties is a 0.01 incremental increase in the required MCPR fuel cladding integrity safety limit.
15.C.2.1 Core Flow Uncertainty 15.C.2.1.1 Core Flow Measurement During Single-Loop Operation The jet pump core flow measurement system is calibrated to measure core 3
ficw when both sets of jet pumps are in forward flow; total core flow is the sum of the indicated loop flows. For single-loop operation, however, some inactive jet pump,s will be backflowing (at active pump speeds above approxi-mately 40%). TheYefore, the measured flow in the backflowing jet pumps must be subtracted from the measured flow in the active loop to obtain the total core flow.
In addition, the jet pump coefficient is different for reverse
~
flow than for forward flow, and the measurement of reverse flow must be modified to account for this difference.
15.C.2-1 t
J
0 r-
~
In single-loop operation, the total core flow is derived by the follow-ing formula:
Total Core \\
" f Active Loop i
[InactiveLoopI
-C Flow Indicated Flow
\\
l-
\\
}
\\IndicatedFlowl The coefficient C (=0.95) is defined as the ratio of " Inactive Loop True Flow" to " Inactive Loop Indicated Flow". " Loop Indicated Flow" is the flow measured by the jet pump " single-tap" loop flow sumers and indicators, which are set to read forward flow correctly.
O The 0.95 factor was the result of a. conservative analysis to appropri-ately modify the single-tap flow coefficient for reverse flow.
If a more exact, less conservative, core flow is required, special in-reactor cali-bration tests can be made. Such calibration tests would involve:
calibrating core support plate AP versus core flow during one-pump and two-pump operation along with 100% flow control line and calculating the correct value of C based on the core support plate AP and the loop flow indicator readings.
15.C.2.1.2 Core Flow Uncertainty Analysis The uncertainty analysis procedure used to establish the core flow uncertainty for one-pump operation is essentially the same as for two-pump operation with some exceptions. The core flow uncertainty analysis is described in Reference 15.C.8-1.- The analysis of one-pump core flow uncer-tainty is sumarized below.
- The analytical ex ected value of the "C" coefficient for HCGS is 0.84.
E 1
15.C.2-2 I
L
[
s
.l HCGS FSAR For single-loop operation, the total core flow can be expressed as follows (refer to Figure 15.C.2-1):
WC*NA~NI where:
.~
WC = total core flow, W4 = active loop flow, and Wg = inactive loop (true) flow.
By applying the " propagation of errors" method to the above equation, the variance of the total flow uncertainty can be approximated by:
2 2
r 1
g,\\,
o8 o
W N
+
1-a W rand (1-aj
+
W C,
C sys A
grand where:
uncertainty of total core flow; o
=
g g
uncertainty systematic to both loops; o
=
g random uncertainty of active loop only; o
=
g random uncertainty of inactive loop only; o
=
uncertainty of "C" coefficient; and
=
C ratio of inactive loop ficw (W ) to active loop flow (W )*
a
=
g A
From an uncertainty analysis, the conservative, bounding values of
'W,,,,
'W
, 'W and oC 1re 1.6%, 2.6%, 3.5%, and 2.8%,
4 g
respectively. Based on the above uncertainties and a bounding value of 0.36 for "a", tha' variance of the total flow uncertainty is approximately:
J
- This flow split ratio varies from about 0.13 to 0.36. The 0.36 value is a conservative bounding value. The analytical expected value of the flow split ratio for HCGS is s 0.33.
15.C.2-3
I r
5 HCGS FSAR Tb 16)s 3.5)2+(2.8)2) c = (1.6)
- (2.6)2 +
+
=(5.0%)2 When the effect of 4.1% core bypass flow split uncertainty at 12%
(bounding case) bypass flow fraction is added to the total core flow uncer-tainty, the active ~ coolant flow uncertainty is:
= (5.0g r 1{b}2 )
(4.1%)2 (5.1%)2
=
ctive
+
coolant which is less than the 5% flow uncertainty assumed in the statistical analysis.
In summary, core flow during one-pump operation is measured in a conservative way and its uncertainty has'been conservatively evaluated.
15.C.2.2 TIP Reading Uncertainty To ascertain the TIP noise uncertainty for single recirculation loop operation, a test was performed at an operating BWR. The test was performed at a power level 59.3% of rated with a single recirculation pump in opera-tion (core flow 46.3% of rated). A rotationally symetric control rod pattern existed during the test.
Five consecutive traverses were made with each of five TIP machines, giving a total of 25 traverses. Analysis of this data resulted in a nodal TIP noise of 2.85%. Use of this TIP noise value as a component of the process computer total uncertaint,y results in a one-sigma process computer total effect TIP uncertainty value for single-loop operation of 6.8% for initial cores and 9.1% for reload cores.
f
- 5.C.2-4
-r-,.-
sans s
I l
J I
/
We wg A
We
- Total Core Flow W4 = Active Loop Flow WI
- Inactive Loop Flow PSE&G ILLUSTRATION OF SINGLE RECIRCULATION LOOP FIGURE OPERATION FLOWS 15.C.2-1 15.C.2-5
[
F 5
15.C.3 MCPR OPERATING LIMIT 15.C.3.1 Abnormal Operational Transient's Operating with one recirculation loop results in a maximum power output which is about 30f below that which is attainable for two-pump operation.
Therefore, the consequences of abnormal operational transients from one-loop operation will'be considerably less severe than those analyzed for two-loop operation. For pressurization, flow increase, flow decrease, and cold water injection transients, the results presented in Chapter 15 bound both the thermal and overpressure consequences of one-loop operation.
The consequences of flow decrease transients are bounded by the full power analysis. A single pump trip from one-loop operation is less severe than a two-pump trip from full pcwer because of the reduced initial power level.
The worst flow increase transient results from a recirculation flow controller failure, and the worst cold water injection transient results from the loss of feedwater heating. For the former event, the K curve is f
derived assuming both recirculation loop controllers fail. This condition produces the maximum possible power increase and hence maximum aCPR for transients initiated from less than rated power and flow. During operation with only one recirculation loop, the flow and power increase associated with this failure with only one loop will be less than that associated with curve derived with the two-pump assumption is both loops; therefore, the Kf consarvative for single-loop operation.
The latter event, loss of feedwater heating, is generally the most seve e cold water' event with respect to increase in core power. This power increase is caused by positive reactivity insertion from increased core inlet subcooling and it is relatively insensitive to initial power level. A generic statistical loss of feedwater heater analysis using different i
e 0
9 e
15.C.3-1 4>
HCGS FSAR initial power levels and other core design parameters concluded one-pump operation with lower initial power level is conservatively bounded by the full power two-pump analysis. Inadvertent restart of the idle recirculation pump has been analyzed in Chapter 15.4.4 and is still applicable for single-loop operation.
In the following sections, results of the two most limiting pres-surization transients analyzed for single-loop cperation are presented.
They are respectively:
a.
Feedwater Controller Failure-Maximum Demand, (FWCF) b.
Generator Load Rejection %ith Bypass Failure, (LRBPF)
/
The plant initial conditions are given in Table 15.C.3-1, 15.C.3.1.1 Feedwater Controller Failure - Maximum Demand This event is postulated on the basis of a single failure of a master feedwater control device, specifically one which can directly cause an increase in coolant inventory by increasing the total feedwater flow. The most severe applicable event is a feedwater controller failure during maximum flow demand. The feedwater controller is assumed to fail ta as upper limit at the beginning of the event.
A feedwater controller failure during maximum flow demand at 75% power and 60% flow during single rer.irculation loop operation produces the se-quence of events listed in Table 15.C.3-2.
Figure 15.C.3-1 shows the changes in important variables during this transient.
The computer model described in Reference 15.C.8-2 was used to simulate this event.
The analysis as been performed with the plant conditions tabulated in Table 15.C.3-1, with the initial vessel water level at Level 4 (instead of nonnal water level) for conservatism. By lowering the initial water level, more cold feedwater will be injected before Level 8 is reached resulting in higher heat fluxes.
15.C.3-2
(
a
'r 1
(
0 The safety analysis condition is at 75% rated thermal power and 60%
rated core flow, which represents single recirculation loop operation at 100% pump speed on the 105% rod line. End of cycle (all rod out) scram characteristics are assumed. The safety-relief valve action is conserva-tively assumed to occur with higher than nominal setpoints. The transient is simulated by. programming an upper limit failure in the feedwater system such that 159% of rated feedwater flow occurs at the reactor dome pressure of 973 psig, and 135% of rated feedwater flow would occur at the design pressure of 1060 psig.
The simulated feedwater controller transient is shown in Figure 15.C.3-1.
The high ' water level turbine trip and feedwater pump trip are initiated at approximately 6.1 seconds. Scram occurs simultaneously from stop valve closure, and limits the neutron flux peak and fuel thermal transient. The turbine bypass system opens to limit peak pressures in the steam supply system. Events caused by low water level trips, including initiation of HPCS and RCIC core cooling system functions are not included in the simulation. Should these events occur, they will follow sometime after the primary effects have occurred, and are expected to be less severe than those already experienced by the systein.
Table 15.C.3-4 gives a summary of the transient analysis results. The calculated MCPR is 1.17, which is well above the safety limit MCPR of 1.07 so no fuel failure due to boiling transition is predicted. The peak vessel pressure predicted is 1121 psig and is well below the ASME limit of 1375 psig.
15.C.3.1.2 Generator Load Rejection with Bypass Failure i
Fast closure'of the turbine control valves (TCV) is initiated whenever electrical grid disturbances occur which result in significant loss of electricalloadohthegenerator. The turbine control valves are required to close as rapidly as possible to prevent excessive overspeed of the turbine'-generator'(T-G) rotor. Closure of the main turbine control valves,
will increase system pressure.
15.C.3-3
T I
.( q HCGS FSAR A loss of generator electrical load with bypass failure at 75% power and 60% flow during single recirculation loop operation produces the se-quence of events listed in Table 15.C.3-3.
Figure 15.C.3-2 shows the changes in important variables during this transient.
Turbine control valve (TCV) fast closure initiates a scram trip signal for power levels greater than 30% NB rated.
In addition, recirculation pump trip is initiated. Both of these trip signals satisfy single failure criterion and credit is taken for these protection features.
The pressure relief system which operates the relief valves indepen-dently when systen pressure exceeds relief valve instrumentation setpoints is assumed to function normally during the time period analyzed.
All plant control systems maintain normal operation unless specifically designated to the contrary.
The computer model described in Reference 15.C.8-2 was used to simulate this event.
The analysis has been performed with the plant conditions tabulated in Table 15.C.3-1, except that the turbine bypas: function is assumed to fail.
The safety analysis condition is at 75% rated thermal power and 60%
rated core flow, which represents single recirculation loop operation at 100% pump speed on the 105% rod line.
The turbine electro-hydraulic control system (EHC) power / load unbalance device detects load rejection before a measurable speed change takes place.
The closure characteristics of the turbine control valves are assumed such that the valv'es operate in the full arc (FA) mode and have a full strokeclosuretimb,fromfullyopentofullyclosed,of0.15second.
9 15.C.3-4
f I
r i
HCGS FSAR Auxiliary power would normally be independent of any turbine-generator overspeed effects and continuously be supplied at rated frequency as auto-matic fast transfer to auxiliary power supplies occurs.
The simulated generator load rejection with bypass failure is shown in Figure 15.C.3-2, '
uDC Events caused by low water level trips, including initiation of HPCI
^
and RCIC core cooling system functions are not included in this simulation.
If these events occur, they will follow sometime after the primary concerns of fuel margin and overpressure effects have passed, and will result in effects less severe than those already experienced by the reactor system.
and will provide long-tenn reactor inventory control.
Table 15.C.3-4 summarizes the transient analysis results. The peak neutron flux reaches about 120% of rated and average surface heat flux peaks at about 104*. of its initial value. The peak vessel pressure predicted is 1162 psig and is well below the ASME limit of 1375 psig. The calculated MCPR is 1.16 which is considerably above the safety limit MCPR of 1.07.
15.C.3.1.3 Sumary and Conclusions The transient peak value results and the Critical Power Ratio (CPR) results are summarized in Table 15.C.3-4.
This table indicates that for the transient events analyzed here, the MCPRs for all transients are above the single-loop operation safety limit value of 1.07.
It is concluded that the operating limit MCPRs established for two-pump operatior. are also applicable to single-loop operation conditi3ns.
4 For pressuriiation, Table 15.C.3-4 indicates that the peak pressures are well below the.ASME code value of 1375 psig. Hence, it is concluded that the pressure harrier integrity is maintained under single-loop opera-tion.
9 15.C.3-5
I' e
1 HCGS FSAR From the above discussions, it is concluded that the transient conse-quences for one-loop operation are bounded by previously submitted full power analyses.
15.C.3.2 Rod Withdrawal Error The rod withdrawal error at rated power is given in the FSAR. These
~
analyses are performed to demonstrate, even if the operator ignores all instrument indications and the alarm which could occur during the course of the transient, the rod block system will stop rod withdrawal at a minimum critical power ratio'(MCPR) which is higher than the fuel cladding integrity safety limit. Modification of the rod block equation (below) and lower power assures the MCPR safety limit is not violated.
One-pump operation results in backflow through 10 of the 20 jet pumps while the flow is being supplied into the lower plenum from the 10 active jet pumps. Because of the backflow through the inactive jet pumps, the present rod block equation was conservatively modified for use during one-pump operation because the direct active-loop flow measurement may not indicate actual flow above about 40% core flow without correction.
A procedure has been established for correcting the rod block equation to account for the discrepancy between actual flow and indicated flow in the active loop. This preserves the original relationship between rod block and actual effective drive flow when operating with a single-loop.
The two-pump rod block equation is:
RB = mW + RB100 - m(100)
The one-pump, equation becomes:
RB = mW'+ RB100 - m(100) - mW O
15.C.3-6 a
m
b 4
where difference between 'two-loop and single-loop effective aW
=
drive flow at the same core flow. This value is expect-ed to be 8% of rated (to be determined by PSE&G).
RB power at rod block in %;
=
flow reference slope for the rod block monitor (RBM) m
=
drive flow in % of rated.
W
=
top level rod block at 100% flow.
=
100 If the rod block setpoint (RB100) is changed, the equation must be recal-culated using the new value.
The APRM trip settings are flow biased in the same manner as the rod block monitor trip setting. Therefore, the APRM rod block and scram trip settings are subject to the same procedural changes as the rod block monitor trip settings discussed above.
15.C.3.3 Operating MCPR Limit For single-loop operation, the operating MCPR limit remains unchanged from the normal two-loop operation limit. Although the increased uncer-tainties in core flow and TIP readings resulted in a 0.01 incremental increase in MCPR fuel cladding integrity safety limit during single-loop i
operation (Section 15.C.2), the limiting transients have been analyzed to indicate that the're is more than enough MCPR margin during single-loop operation to compensate for this increase in safety limit. For single-loop operation-at loweg flows, the steady-state operating MCPR limit is estab-lished by the K curve. This ersures the 99.9% statistical limit require-f ment is always satisfied for any postulated abnormal operational occurrence 15.C.3-7
i e'
5 HCGS FSAR Since the maximum core flow runout during single loop operation is only about 60% of rated, the current flow dependent K curve which is generated f
based on the flow runout up to rated c6re flow are also adequate to protect the flow runout events during single-loop operation.
W 9
9 4
em I
l' i
15.C.3-8
0 a'
s HCGS FSAR TABLE 15.C.3-1 INPUT PARAMETERS AND' INITIAL CONDITIONS 1.
Thermal Power Level, MWt 2470
~
6 2.
Steam Flow,' lb per hr 10.17 x 10 6
3.
Core Flow, lb per hr 60.00 x 10 4.
Feedwater Flow Rate, lb per sec 2824 5.
Feedwater Temperature, 'F 390 6.
Vessel Dome Pressure, psig 973 7.
Vessel Core Pressure, psig 978 8.
Turbine Bypass Capacity, % NBR 25 9.
Core Coolant Inlet Enthalpy, Btu per lb 512.1
- 10. Turbine Inlet Pressure, psig 944
- 11. Fuel Lattice C(P8x8R)
- 12. CoreAverageGagConductance, Btu /sec-ft "F
0.1744
- 13. Core Bypass Flow, %
11.27
- 14. Required Initial MCPR 1.28
- 15. MCPR Safety' Limit 1.07
- 16. Doppler Coefficient, c/ F
- 17. Void Coefficient, c/% Rated Voids
- 18. Core Average Rated Fraction', %
45.1
- 19. Scram Reactivity, $aK
- 20. Control Rod Qrive Speed Position versus Time Figure 15.0-2 This value is calculated within the computer code (Reference 15.C.8-2) for end of' Cycle 1 conditions based on input from the CRUNCH file.
- K times the Rated Operating Limit MCPR f
15.C.3-9
I r
1 HCGS FSAR TABLE 15.C.3 Cont.
INPUT PARAMETERS AND INITIAL CONDITIONS
- 21. Fuel Exposure End of Cycle 1
~
- 22. Jet Pump Ratio, M 3.56
- 23. Safety / Relief Valve Capacity, % NBR 0 1121 psig 85.8 Manufacturer Target Rock Quantity Installed 14
- 24. Relief Function Delay, seconds 0.4
- 25. Relief Function Response Time Constant, seconds 0.15
- 26. Setpoints for Safety / Relief Valves Safety / Relief Function, psig 1121, 1131, 1141
- 27. Number of Valve Groupings Simulated Safety / Relief Function 3
- 28. Safety / Relief Valve Reclosure Setpoints, psig 1087, 1097, 1107
- 29. High Flux Trip Setpoint, % NBR (121 x 1.043) 126.2
- 30. High Pressure Scram Setpoint, psig 1071
- 31. Vessel Level Trips, Feet Above Separator Skirt Bottom Level 8 - (L8), feet 6.042 Level 4 - (L4), feet 3.625 Level 3 - (L3), feet 1.792 Level 2 - (L2), feet
-3.708
- 32. APRM Simulated Thermal Power Trip Setpoint,
% NBR (117 x 1.043) 122.0
- 33. Recirculation Pump Trip Delay, seconds 0.175 34.
Inertia Time-Constant of Recirculation Purrp Trip (maximum), seconds 4.5
- 35. Total.SteamlfneVolume,ft3 6619
- 36. Pressure Setpoint of ATWS Recirculation Pump Trip, psig 1101 15.C.3-10
g ar -
3 HCGS FSAR TABLE 15.C.3-2 SEQUENCE OF EVENTS FOR FIGURE 15.C.3-1 FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND Time-sec Event 0
Initiate simulated failure to the upper limit on feedwater flow.
6.1 LB vessel level setpoint trips main turbine and feedwater pumps'. Turbine bypass operation initiated.
6.1 Reactor scram trip actuated from main turbine stop valve position switches.
6.1 Recirculation pump trip (RPT) actuated by stop valve position switches.
6.2 Main turbine stop valves closed and turbine bypass valves start to open.
6.3 Recirculation pump motor circuit breaker opens causing decrease in core flow.
9 D
15.C.3-11
.\\
h
- ~
A HCGS FSAR TABLE 15.C.3-3 SEQUENCE OF EVENTS FOR FIGURE 15.C.3-2 GENERATOR LOAD REJECTION WITH BYPASS FAILURE Time-sec Event
(-)0.015 Turbine-generator detects loss of electrical load.
(approx.)
0 Turbine-generator load rejection sensing devices trip to initiate turbine control valve fast closure.
O Turbine bypass valves fail to operate.
O Fast control valve closure (FCV) initiates scram trip and recirculation pump trip (RPT).
0.07 Turbine control valves closed.
0.175 Recirculation pump motor circuit breaker opens causing decrease in core flow.
2.2 Group 1 relief valves actuated.
2.6 Group 2 relief valves actuated.
2.8 Group 3 relief valves actuated.
i 4.4 Gr'oup 3 relief valves start to close.
9.5 All relief valves are closed.
O 9
15.C.3-12
--.w
0 t-
SUMMARY
OF TRANSIENT PEAK VALUE AND CPR RESULTS FWCF LRBPF Initial Power / Flow (% Rated) 75/60 75/60 Peak Neutron Flux (% Rated) 91.2 119.7 Peak Heat Flux (% Initial) 103.3 103.9 Peak Dome Pressure (psig) 1107 1148 Peak Vessel Bottom Pressure (psig) 1121 1162 Required Initial MCPR 1.28 1.28 Transient MCPR 1.17 1.16 Safety Limit MCPR (for SLO)
.l.07 1.07 Margin to Safety Limit 0.10 0.09
- K times the Rated Operating Limit MCPR.
f
- Includes Option A adder.
4 E
t O
15.C.3-13
'I a'
4 a;
1 LEVEL (INC H-REF-SEP-SKIRT 2 W R SENSE D LEVEL (INCHES) 150.
3 N R SENSE D LEVEL (INCHES) 4 CORE INLE T FLOW (PCT 1 5 DRIVE FLC W 2 (PCT) 100.
p i
P w
ui m
k 4
' //
3 50.
~
~
~
0.'iI'
O.
5.
10.
15.
20.
TIME (SEC).
l FIGURE PSE&G FEEDWATER CONTROLLER FAILURE - MAXIMUM DEMAND 15.C.3-1 75% POWER /60% CORE FLOW O
=
! = jll; 1
PX MU E 3. 'D R
EL U C.T G
N TF I 5O W
F1C l
RT O
2 ER L
~
5CI TE WF NH O
XE LM r
UCE FA.
L C
E 0
t FLF FT 2
EF ES NUR T
OFU A 'L R
I EEV EE D
N NPAFV AM E
123 i5 5
l D
~
1 M
)
J4 C
W l
l t
IO E
X L A F S
M
(
E R
h(I O
E E C R
M U %
L0 I
I6
.T A/
0 FR I
E
\\
1 R W E O L P LO%
R5 T 7 NO 8
5 C
E R
5 E
l T
f AW D
E 5
E F
.W l
4 2
4~
'O
~
0 0
0 0
5 0
5
~ 1 1
G
.A 5$' ta W5oFoS E
t S
P F CL.w
4 t
1 VESSEL PF ES RISE (PSI) 2 STM LINE PRES RISE (PSI) 200*
3 TURBINE F RES RISE (PSI) 4 BYPASS ST ERM FLOW (PCT) 5 RELIEF VF LVE FLOW (PCT) 6 TURB STEF M FLQW (PCT) 100~
3
=
2.
N i,
N m
5 4
4 3
5 \\\\
0-12 %
5 6
5 6
~.
- 100.
' ' ' ' l ' ' ' '
0.
5.
10.
15.
20.
TIME (SEC)
PSE8G FEEDWATER CONTROLLER FAILURE - MAXIMUM DEMAND
_j 75% POWER /60% CORE F10W rout'n.
4 4
1 VOID REAC TIVITY 2 DOPPLER F EACTIVITY 3-3 SCRAM REF CTIVITY yn emw 1
/
0 W~
S I
1 i
1 5
8 n
b
-1.
i 1
)
3 h
~
Y1
-7 o
3 CE LLI
~
l''
lt
-2'O.
5.
10.
15.
20.
TIME (SEC)
FIGURE PSE&G FEEDWATER CONTROLLER FAILURE - MAXIMUM DEMAND 15.C.3-1 75% POWER /60% CORE FLOW CONT'D.
.o
J 4
1 LEVEL (INC H-REF-SEP-SKIRT 2 W R SENSE D LEVEL (INCHES) 150.
3 N R SENSE D LEVEL (INCHES) 4 CORE INLE T FLOW (PCT) 5 ORIVE FLC H 2 (PCT) e 100'
\\
f S
h L
a 50.
s +ti
~
3 q
[
~
N
~ 'l'
0 'O "
2.
Lj,
6, 8.
TIME (SEC) i PSE&G' GENERATOR LOAD REJECTION WITH BYPASS FIGURE FAILURE, 75% POWER /60% CORE FLOW 15.C.3.2 w
ac 5
',=
1 2
PX E 3. 'D R
MU U C.T G
N EL f
I5O F1C TF W
RT O
EA L
TE WF NH O
XE LM 1
1 UCE FR L
C E
8 FLF FT 4
EF ES NUR T
OFU AL R
S WE TK DS UAE ES EEV EE NPA FV 123 45 6
S 1
)
S C
AW PO E
YL S
BF
(
WC 6
M N%
D O0 I
I6 5
.T T/
CR EE 4
JW N
1 A5
~
O7
%x L
b G
2 N
i
- c 1
J i
s 2
_a 1
- h 0
~
0 0
0 5
0 1
1 G
8W u
zy E
S P
J 1 VESSEL PF ES RISE (PSI) 2 STM LINE PRES RISE (PSI) 200-
,A 3 TURBINE F RES RISE (PSI) v 4 RELIEF VF LVE FLOW (PCT) l 100.
S W-T
\\i
=1
=
.p m
L 4
4 3!
o 1
- o O~
4 6
6 i
- 10 0. ~.... I....
0.
2.
4.
6.
8.
TIME (SEC)
FIGURE PSE&G GENEPATOR LOAD REJECTION WITH BYPASS 15.C.3.2 FAILURE, 75% POWER /60% CORE FLO11 CONT'D.
=
x8" E=
s 2
E 3. en R
U C. r G
u
~
F1m I5 Y
T IY Y
YVT T
1 TII I
ITV V
VCI I
IRT T
TECC 4
\\
EFFF R
EE ERRR RELML DPAA 1
IPRT OOC O
VDS T
SSAt 123 4
3 P0 l
)
1 C
1 C
N%
E O0 S
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i HCGS FSAR 15.C.4 STABILITY ANALYSIS e
15.C.4.1 Phenomena The primary contributing factors to the stability performance with one recirculation loof not in service are the power / flow ratio and the recircu-lation loop characteristics. As forced circulation with only one recircu-lation loop in operation, the reactor core stability is influenced by the inactive recirculation loop. As core flow increases in SLO, the inactive jet pump forward flow decreases because the driving head across the inactive jet pumps decreases with increasing core flow. The reduced flow in the inactive loop reduces the resistance that the recirculation loops impose on reactor core flow perturbations thereby adding a pestabilizing effect. At the came time the increased core flow results in a lower power / flow ratio which is a stabilizing effect. These two countering effects result in slightly decreased stability margin (higher decay ratio) initially as core flow is increased (from minimum) in SLO and then an increase in stability margin (lower decay ratio) as core flow is increased further and reverse flow in the inactive loop is estabiished.
As core flow is increased further during SLO and substantial reverse flow is established in the inactive loop an increase in jet pump flow, core flow and neutron noise is observed. A cross flow is established in the annular downcomer region near the jet pump suction entrance caused by the reverse flow of the inactive recirculation loop. This cross flow interacts with the jet pump suction flow of the active recirculation loop and in-creases the jet pump flow noise. This effect increases the total core flow noise which tends to drive the nitutron flux noise.
O To determine'if the increased noise is being caused by reduced stabil-ity margin as SLO; core flow was increased, an evaluation was performed which phenomenologically, accounts for single-loop operation effects on stability, as summarized in R'eference 15.C.8-3.
The model predictions were initially O
15.C.4-1
b HCGS FSAR compared with test data and showed very good agreement for both two-loop and single-loop test conditions. An evaluation was performed to detemine the effect of reverse flow on stability during SLO. With increasing reverse flow, SLO exhibited slightly lower decay ratios than two-loop operation.
However, at core flow conditions with no reverse flow, SLO was slightly less stable. This is, consistent with observed behavior in stability tests at operating BWRs (Reference 15.C.8-4).
In addition to the above analyses, the cross flow established during reverse flow conditions was simulated' analytically and shown to cause an increase in the individual and total jet pump flow noise, which is consis-tent with test data (Reference 15.C.8-3). The results of these analyses and tests indicate that the stability characteristics are not significantly different from two-loop operation. At low core flow, SLO may be slightly less stable than two-loop operation but as core flow is increased and reverse flow is established the stability performance is similar. At higher core flow with substantial reverse flow in the inactive recirculation loop, the effect of cross flow on the flow noise results in an increase in system noise (jet pump, core flow and neutron flux noise).
15.C.4.2 Compliance to Stability Criteria Consistent with the philosophy applied to two-loop operation, the stability compliance during single-loop operation is demonstrated on a generic basis. Stability acceptance criteria have been established to l
demonstrate compliance with the requirements set forth in 10CFR50, Appendix A General Design Criterion (GDC) 12 (Reference 15.C.8-5). The generic i
stability analysis has been performed covering all licensed GE BWR initial core fuel designs including those fuels contained in the General Electric Standard Applicati'on for Reload Fuel (GESTAR Reference 15.C.8-6 through Amendment 10).
The analysis demonstrates that in the event limit cycle ncutron flux oscil ations occur within the bounds of safety system inter-l s
The reload fuel designs contained in GESTAR include fuel designs through the GE-8 design (including barrier fuel).
15.C.4-2 l
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HCGS FSAR vention, specified acceptable fuel design limits are not exceeded. Since the reactor core is assumed to be in an oscillatory mode, the question of stability margin during SLO is not relevant from a safety standpoint (i.e.,
the analysis already assumes no stability margin).
The fuel performance during limit cycle oscillations is characteris-tically dependent on fuel design and certain fixed system features (high neutron flux scram setpoint, channel inlet orifice diameter, etc.). There-l fore the accepta'bility of GE fuel designs independent of plant and cycle parameters has been established. Only those parameters unique to SLO which affect fuel _ performance need to be evaluated. The major consideration of -
-SLO is the increased Minimum Critical Power Ratio (MCPR) safety limit caused by increased uncertainties in system parameters during SLO. However, the increase in MCPR safety limit (0.01) is.well within the margin of the limit
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. cycle analyses (Reference 15.C.8-5) and therefore it is demonstrated that stability compliance criteria are satisfied during single-loop operation.
l Operationally, the effects of higher flow noise and neutron flux noise observed at high SLO core flow are evaluated to determine if acceptable vessel internal vibration levels are met and to determine the effects on fuel and channel fatigue, and are not considered in the compliance to stability criteria.
Service Information Letter-380, Revision 1 (Reference 15.C.8-7) has been developed to inform plant operators how to recognize and suppress l
unanticipated oscillations when encountered during plant operation.
As a result of the above analysis and operator recommendations, the NRC i
staff has approved the generic stability analysis for application to single-loop operation (Reference 15.C.8-8) provided that the recommendations of j
SIL-380 have beert incorporated into the Plant Technical Specifications.
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4 15.C.5
. LOSS-OF-COOLANT ACCIDENT ANALYSIS If two recirculation loops are operating and a pipe break occurs in one of the two recir'culation loops, the pump in the unbroken loop is assumed to insnediately trip and begin to coast down. The decaying core flow due to the pump coastdown results in very effective heat transfer (nucleate boiling) during the initial phase of the blowdown. Typically, nucleate boiling will be sustained during the first 5 to 9 seconds after the accident, for the design basis accident (DBA).
If only one recirculation loop is operating, and the break occurs in the operating loop, continued core flow is provided only by natural circu-lation because the vessel is blowing down to the reactor containment through both sections of the broken loop. The core flow decreases more rapidly than in the two-loop operating case, and the departure from nucleate boiling for the high power node might occur 1 or 2 seconds after the postulated acci-dent, resulting in more severe cladding heatup for the one-loop operating Case.
In addition to changing theblowdown heat transfer characteristics, losing recirculation pump coastdown flow can also affect the system invento-ry and reflooding phenomena. Of particular interest are the changes in the high-power node uncovery and reflooding times, the system pressure and the time of rated core spray for different break sizes. One-loop operation results in small changes in the high-power node uncovery times and times of rated spray. the effect of the reflooding times for various break sizes is l
also generally small.
Ananalysisdfsinglerecirculationloopoperationusingthemodelsand assumptions docume6ted in Reference 15.C.8-9 was performed for HCGS. Using i
this method, SAFE /REFLOOD computer code runs were made for a full spectrum l
of large break sizes for only the recirculation suction line breaks (most limiting for HCGS). Because the reflood minus uncovery time for the single-l l
15.C.5-1
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loop analysis is similar to the two-loop analysis, the maximum planar linear heat generation rate (MAPLHGR) curves were modified.
15.C.5.1 ' Break Spectrum Analysis SAFE /REFLOOD' calculations were performed using assumptions given in Section II.A.7.3.1 of Reference 15.C.8-9.
Hot node uncovered time (time between uncovery and reflood) for single-loop operation is compared to that for two-loop operation in Figure 15.C.5-1.
The total uncovered time for two-loop operation is 128 seconds for the 100% DBA suction break. This is the most limiting break for two-loop i
operation. For single-loop operation the total uncovered time is 132 seconds for the 100% DBA suction break. This is the most limiting break for single-loop operation.
15.C.S.2 Single-Loop MAPLHGR Determination The small differences in uncovered time and reflood time for the limiting break size would result in a small change in the calculated peak cladding temperature. Therefore, as noted in Reference 15.C.8-9, the one and two-loop SAFE /REFLOOD results can be considered similar and the generic alternate procedure described in Section II.A.7.4 of this reference was used to calculate the MAPLHGR reduction factors for single-loop operation. The most limiting single-loop operation MAPLHGR reduction factor (i.e., yielding the lowest MAPLHGR) is 0.86.
One-loop operation MAPLHGR values are derived by multiplying the current two-loop MAPLHGR values by the reduction factor 0.86.
As discussed in Reference *15.C.8-9, single recirculation loop MAPLHGR values are conservative when calculated in this manner. This MAPLHGR multiplier is applicable to GE-6 and GE-7 fuels in the initial core. For reload situations, the MAPLHGR must be assessed for each cycle to determine ifitisstillapklicablebecausethesingle-loopMAPLHGRmultiplierwas based on the calculated peak cladding temperature from the two-loop analysis for the' initial core fuel.
15.C.5-2 i
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HCGS FSAR 15.C.5.3 Small Break Peak Cladding Temperature Section II.A.7.4.4.2 of Reference 15.C.8-9 discusses the lov sensitivi-ty of the-calculated peak cladding temperature (PCT) to the assumptions used in the one-pump operation analysis and the duration of nucleate boiling. As this slight increa'se (s 50*F) in PCT is overwhelmingly offset by the de-creased MAPLHGR (equivalent to 300*F to 500*F PCT) for one-pump operation, the calculated PCT values for small breaks will be well below the 1694 I small break PCT value previously reported for HCGS, and significantly below the 2200*F 10CFR50.46 cladding temperature limit.
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. CONTAINMENT ANALYSIS The range of. power / flow conditions which are included in the SLO cperating domain for Hope Creek were investigated to determine if there would be any impact on the FSAR specifications for containment response, including the containment dynamic loads. The SLO operating conditions were confirmed to be within the range of operating conditions which have previously been considered in defining the containment pressure and temperature response and containment dynamic loads for two-loop operation.
Therefore, tt.e containment response for Hope Creek with single-loop operation has been confirmed to be within the present design values.
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HCGS FSAR 15.C.7 MISCELLANE0US IHpACT EVAL'JATION 15.C.7.1 Anticipated Transient Without Scram (ATWS) Impact Evaluation The principal difference between single-loop operation (SLO) and normal two-loop operation (TLO) affecting Anticipated Transient Without Scram (ATS) performance is that of initial reactor conditions. Since the SLO initial power flow condition is less than the rated conaition used for TLO ATWS analysis, the transient response is less severe and therefore bounded by the TLO analyses.
It is concluded that if an ATWS event were initiated at HCGS from the SLO conditions, the results would be less severe than if it were initiated from rated conditions.
15.C.7.2 Fuel Mechanical performance Evaluations were performed to determine the acceptability of HCGS single-loop operation on both GE-6 and GE-7 fuel rod and assembly thermal /-
4 mechanical performance. Component pressure differential and fuel rod overpower values were determined for anticipated operational occurrences initiated from SLO conditions. These values were found to be bounded by j
those applied in the fuel rod and assembly design bases.
It is observed that due to the substantial reverse flow established during SLO both the Average Power Range Monitor (APRM) noise and core plate 4
differential pressure noise are slightly increased. An analysis has been carried out to determine that the APRM fluctuation should not exceed a flux amplitude.of 215% 'of rated and the core plate differential pressure fluc-tuation should notl exceed 3.2 psi peak to peak to be consistent with the fuel rod and assembly design bases.
5 15.C.7-1
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<O HCGS FSAR 15.C.7.3 Vessel Internal Vibration Vibration tests for SLO were performed during the startup of two BWR 4-251 plants. An extensive vibration test was conducted at a prototype BWR 4-251 plant, Browns Ferry 1, the results of which are used as a standard for comparison. A confirmatory vibration test was performed at the Peach Bottom 2 & 3 plants.
The Browns Ferry 1 test data demonstrates that all instrumented vessel interrals components vibrations are within the allowable criteria. The highest measured vibration in terms of percent criteria for single-loop operation was 70%. This was measured at a jet pump riser brace during cold flow conditions at 100% of rated pump speed.
The Peach Bottom vibration test data shows that vessel internals vibration levels are within the allowable criteria for all test conditions.
The highest measured vibration in terms of percent criteria for single-loop operation was 96%. This was measured at a jet pump elbow location during 68% power condition at 92% of rated pump speed. This vibration amplitude is the highest, in terms of percent criteria, experienced in vessel internals for the BWR 4-251 plants studied.
The conclusion is that under all operating conditions, the vibration level is acceptable. However, due to the high vibration levels recorded, it is recomended that Hope Creek not perform single-loop operation with pump speed exceeding 90% of rated pump speed. The same recomendation has been accepted by the Browns Ferry and Peach Bottom plants.
3 This analysis is conservative because the criteria is developed by assuming that the -plant operates on a steady state single loop operations throughout the pla,nt life.
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HCGS FSAR 15.C.8 REFERENCES 15.C.8-1 " General Electric BWR Thermal Analysis Basis (GETAB); Data, Correlation, and Design Application", NED0-10958-A, January 1977.
15.C.8-2 " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors", NED0-24154, October 1978.
15.C.8-3 Letter, H.C. Pfefferlen (GE) to C.O. Thomas (NRC), " Submittal of Response to Stability Action Item from NRC Concerning Single-loop Operation," September 1983.
15.C.8-4 S.F. Chen and R.O. Niemi, " Vermont Yankee Cycle 8 Stability and Recirculation Pump Trip Test Report", General Electric Company.
August 1982 (NEDE-25445, Proprietary Information).
15.C.8-5 G.A. Watford, " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria", General Electric Company, October 1984 (NEDE-22277-P-1, Proprietary Information).
15.C.8-6 " General Electric Standard Application for Reload Fuel", General Electric Company, April 1983 (NEDE-240ll-P-A-6).
15.C.8-7 "BWR Core Thermal Hydraulic Stability", General Electric Company, February 10, 1984 (Service Information Letter-380, Revision 1).
15.C.8-8 Letter, C.O. Thomas (NRC) to H.C. Pfefferlen (GE), " Acceptance for Referertcing of Licensing Topical Report NEDE-24011, Rev. 6, Amendmeht 8. Thermal Hydraulic Stability Amendment to GESTAR II,"
April 24, 1985.
0 15.C.8-1
0
. erd HCGS FSAR 15.C.8 REFERENCES (Cont'd) 4 15.C.8-9 " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K Amendment No. 2 -
One Recirculation Loop Out-of-Service", NEDO-20566-2 Revision 1 July 1978.
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15.C.8-2
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ATTACHMENT III 9
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.c., s NO SIGNIFICANT HAZARDS CONSIDERATIONS The proposed Technical Specification changes can be considered not likely to involve significant hazards considerations per the example provided in 48FR14870, paragraph (vi):
"A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan:
for example, a
change resulting from the application of a small refinement of a
previously used calculational model or design method."
This conclusion is based on the following:
I.
The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Specifically:
A.
A review of the limiting Anticipated Operational Occurrence (AOOs) was performed by GE specifically for Hope Creek Generating Station (HCGS) to demon-strate adequate margin to the Maximum Critical Power Ratio (MCPR)
Safety Limit.
A review of the values used in the statistical analysis of the determination of the fuel cladding safety limit was performed.
Increased uncertainties for the total core flow and TIP readings resulted in a
0.01 increase in the MCPR safety limit.
Although the MCPR increased by 0.01, the analysis of the A00s demonstrated there is enough margin not to increase the MCPR operating limit or the flow dependent MCPR limit.
B.
A review of the LOCA event was performed by GE.
The analysis of the limiting recirculating pump discharge pipe break, while in SLO, results in i
a longer
(+4 sec) peak node uncovered time.
To maintain tha same peak clad temperature as in two loop operation, the analysis shows the Maximum Average. Planar Lina r Heat Generation Rate (MAPLHGR) needs to be reduced by a factor of 0.86.
The containment response for a DBA recirculation line break in SLO is bounded by the rated power two-loop operation analysis presented in the FSAR.
O
.o. o l
NO SIGNIFICANT HAZARDS CONSIDERATIONS The proposed Technical Specification changes can be considered not likely to involve significant hazards considerations per the example provided in 48FR14870, paragraph (vi):
"A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan:
for
- example, a
change resulting from the application of a small refinement of a
previously used calculational model or design method."
This conclusion is based on the following:
I.
The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Specifically:
A.
A review of the limiting Anticipated Operational Occurrence ( AOOs) was performed by GE specifically for Hope Creek Generating Station (HCGS) to demon-strate adequate margin to the Maximum Critical Power Ratio (MCPR)
Safety Limit.
A review of the values used in the statistical analysis of the determination of the-fuel cladding safety limit was performed.
Increased uncertainties for the total core flow and TIP readings resulted in a
0.01 increase in the MCPR safety limit.
Although the MCPR increased by 0.01, the analysis of the AOOs demonstrated there is enough margin not to increase the MCPR operating limit or the flow dependent MCPR limit.
B.
A review of the LOCA event was performed by GE.
The analysis -of the limiting recirculating pump discharge pipe break while in
- SLO, results in a
longer
(+ M sec), peak node uncovered time.
To maintain the same peak clad temperature as in two loop operation, the analysis shows the Maximum.
Average Planar Linear Heat Generation Rate (MAPLHGR) needs to be reduced by a factor of 0.86.
The containment response for a DBA recir-culation line break in SLO is bounded by the rated power two-loop operation analysis presented in the FSAR.
.[
C.
Thermal-hydraulic.
stability was evaluated for its adequacy with respect to General Design Criteria 12 (10CFR50, Appendix A).
It is shown that SLO satisfies this stability criterion.
In addition, o
HCGS. Technical Specifications have implemented surveillance requirements for detecting and suppressing power oscillations.
D.
The fuel thermal and mechanical duty for transients occurring during SLO was determined to be bounded by the fuel design bases.
Based on vessel internal vibration, the operating loop pump is limited to 90%
of rated speed.
GE also performed a
vibration analysis and a review of test data taken during SLO on jet pumps with and without restrainer
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set screw gaps.
The results demonstrate that with postulated jet pump gaps, the recirculation pumps can operate up to 80% of rated speed in SLO.
4 II.
The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Although this change allows extended operation in a configuration that was previously allowed only for a limited period, analysis has shown (as described in I above) that operation with one recirculation loop out of service is within existir.y i
analyses based on the proposed revised Technical Specification requirements.
III. The proposed changes does not involve a significant reduction is a margin of safety.
The basis for this statement is outlined in I above for each element of the safety analyses which is affected.
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