ML20090L702
| ML20090L702 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 06/01/1973 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20090L685 | List: |
| References | |
| NUDOCS 9102120547 | |
| Download: ML20090L702 (11) | |
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2.2 The reactor coolant system integrity is an important barrier in the prevention of uncontrolled l
release of fission products. It is essential that the integrity of this system be protected by
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establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.
The pressure safety limit of 1335 psig as measured in the vessel steam space is eaulvalent to 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig value was derived fron the design pressures of the reactor pressure vessel, coolant piping, and recirculation p=p casing.
1 The tespective design pressures are 1250 psig at 575 F, 1148 psig at 562 F, and Ih00 psig at 575 F.
I The pressure safety limit was chosen as the Icwer of the pressure transients permitted by the ap-l plicable design codes: A3fs Boiler and Pressura vessel Code Section III-A for the pressure vessel, j
ASME Boiler and Pressure Vessel Code Section III-C for the recircula tion prp casing. an1 the USA 3 Piping Code Section B31.1 for the reactor coolant system pipMg. The A3?E Code permits pressure transients up to 10 percent over the vessel design prassure (lNI, x 1250 = 1375 psig) and the 13A3 Code pemits pressure transients up to 20 percant over the piping design pressure (120% x 1145 =
1378 psig).
The design basis for the reactor pressure vessel makes evident the substantial margin of protection
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against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a l
general membrane stress no Ereater than 26,700 psi at an internal pressure of 1250 psig and te=per-l ature of 575 F; this is more than a factor of 15 below the yield strength of L2,300 psi at this temperature. At the pressure limit of 1375 psig, the general membrane stress increases to 29, Loo psi, still safely balow the yield strength.
i The reactor coolant systen piping provides a comparable margin of prctection at the established pressure safety limit.
CLose A G O ** A LL l
The normal operating pressure of the reactor coolant system is approximately 1025 psig. The -
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~~T' represents tha most severe primary system l
pressure increas+ resulting from an abnormal oparationst transient. The peak pressure in this l
transient is 11f3 psig.
In addition, the safety valves are si::ed assuming no
' ' " M = M 5 f v a u=wg r 2.2 BASES 24 1
EXHIBIT B (Cont) j l
Bases Continued:
I I
2.2 scram in the above transient. The only scram assu:ned is from an indirect means (high flux) and the j
p
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pressure at the bottom of the cessel is limited to 12.t3 psi 6 in this case. Peactor pressure io l
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continuously monitored iri t he control room during operation on a 1500 psig full scale pressure j
q recorder.
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2.4 The settings on the reactor high pressure scra=, reactor coolant system safety / relief valves, turbine control valve fast closure scram, ar..
turbine stop valve closure scram have been established to assure never reaching the reactor coolant system pressure safety limit as well as assuring the sys-tem pressure does not excead the ran6e of the fuel cladding integrity safety limit.
The AWR neutron flux scram and the turbine bypass system also provide protection for these safety limits.
In addition to preventing power operation above 1075 psig, the pressure scram backs up the AFPR neutron flux scram for steam line isolation type transients.
The reactor coolant system safety valves offer yet another protective feature for the reactor coolant system pressure safety limit.
In empliance with Geetion III of the ASME Boiler and Pressure Vessel Code,1965 edition, the safety valves must be set to open at a pressure no hi her than 105 percent 6
of design preoure, and they must limit the reacter pressure to no more than 110 percent of design The safety valves are sized according to the egeJorge pressure.
ion of "f c'
M n M58V closure while operating at lo70 M4t, followed by fiQ. no * - - -- -
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, but M scram from an indirect (high flux) means. With the safety valves set as specyied herein n e maximum vgssel SeNetion L.L.3 Ed:grgyg (agh,eytm,og'g pgsy vessel) would be about 12f3 psig.
Evaluations presenteI =
""ra indi-cate that a total of five valves (2 safety valves and 3 dual purpose safety /relier valves) set at f
the specified pressures maintain the peak pressure during the transient within the code allowable and safety limit pressure.
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The operator will set the reactor coolant high pressure scram trip setting at 1075 psig or lower.
l However, tha actual setpoint can be as much as 10 psi above t he 1075 psig initcated set point due g,
to the deviations discussed in the basis of Specification 2 3 on Fage 22.
In a like anner, the operator will set the reactor coolant system safety / relief valve initiation trip setting at 1080 psig or lower. However, the actual set point can be as much as 11 psi above the 1080 psig indicated set point due to the deviations discussed in tha basis of sp>cification 2 3 on Page c~2.
A violation of this specification is assened to occur only when a device is knowingly set outside of the limiting trip setting, or when a sufficient nc.ber of devices have been affected by any means i
2.4 BASES 26 I
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Pases Continued:
vacuum initintes a clcsure cf the turbine stcp valves and turbine bypass valves which 31 condenser ou ee f eliminates the heat e to the concenser. Closure of the turbine stop and typass valves causes a pressure transient, neutran flux rise, and an increase in surface heat flux. To prevent the clad safety limit from Leing exceeded if this occurs, a reactor scram o curs en turbine stop valve closure. The turbine stop valve closure scram function alone is adequate to preggtg; g safety limit frem beine e r aaded in tia avent of a turtpMuovatr, eient without bypass. M Section t ri n a
13 in em env4nsv*~ soeverrreo -
14.'i12.2 soE*sorryna %ine concenser low vacuum scram is a cact-up to Ehe stop valve closure scram and causes a scram before the stop valves are closed and thus the resulting transient is less severe. Geram occurs at 23" Hg vacuum, stop valve closure occurs at 20" Hg vacua =, and bypass closure at 7" Hg vacuum.
High radiation Jevels in tLe min steanline tennel atme that due to the normal nitrogen and oxygen radioactivity is en indication of leaking fuel. A scram is initiated whenever such radiatien level exceeds ten times nortal full pcver tackgrounl. The purpose of this scram is to reduce the source of such radiation to the extent necessar-to prevent excessive release of mdicactive enterials.
Discharge of exctssive amounts of radioactivity to the site environs is prevented by the air ejector off-gas monitors which cause an isolation to the main condenser off-gas line provided the instantan-cous limit specified in Specification 3.8 is exceeded for a 15-cinute period.
6 The main stearline isolaticn valve closure scram is set to scrss when the isolation valves are Klff%
closed from full open. This scram anticipates the pressure and flux transient, which would occur when the valves cloce. By scra ning at this setting the resultant transient is insignificant.
Ref.
Section Ih.51 31 DABn no suvPLeesurna.
ev es m we~
so w errso Fksarne y 13, 1973 l
l A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate i
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l to the Inrticular plant operating status.
Ref. Section 7 7 1 FSAR.
The manual scram function is active in all seles, thus providing for a r:anual reans of rapidly in-serting control rods during all modes of reactor operation.
The IR'i system provides protection against excessive power levels and short reactor periods in the f
3 1 BASES 39 f
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ewueg.
a t
anagC
.ij2 n
st a d n '4 ott o
n t mynt oh-nn ni eRt nt see I
nat o mt rL (oy gi n
i J
t oerosP hA coathhn odi ya e'
e ei n
cmocuS T S a c pi t t i m
c vtb r s*
nnhh o
a li s se et c-hucewhno%
C r
eet aoidd ev a-5 c
hi ne*
1 Tf alttUr t girt S
s S
e E
s N
S s
A E
C B
N 3
4
/
3 3
l
'l j ;
1 4'
l!)l]!4i!l
%. + Y 7 @ Y T ( [ 7 7 f.
I
.. - ~ -
S q
3_. ; y :v,-x
-y e
,n r
i EXHIBIT B (Cont)
Bases Continued 3 3 and h.3:
,s +,.7
- c. n,,,,
- ~
o mx
+mg L_
n
- i a_
-y_.c-s_.-,
Ee scran tines for all control rois vill le detemined at the tire of each refuelirz cutaEe-De weekly control rod exercise test serves as a periolic check against deterioration of the control rol systen and also verifies the ability of the cont:ul rod drive to scran since if a rol can le =cved with drive pressure, it vill scram lecause of higher pressure applied during scra=.
We frequency of exercising the control rols under the conditions of two or ':cre centrol rods cut of service pro-vides even further assurance of the reliability of the rc=ainit.c control rods.
Ee occurrence of scran times within the limits, but significantly longer than the average, should be viewed as an indicatien of a systematic proble= with centrol rod drives especially if the number of drives exhibiting such scran times exceeds six,the allevable nunber of incperable rods.
D.
Control Toi Accunulaters Ee basis for this specificaticn was not descriled in the GR ani, therefore, is presented in its entirety. Eequiring no more than one incperable accurulater in av nine-rod square array is based on a series of XY FIX4 4 quarter core calculations of a cold, clean core. 'Ihe vorst case in a nine-rol vithdrawal sequence resulted in a kerr /_1.0 -- other repeating rod sequences with = ore rois withdrawn resulted in keff) l.0.
At reactor pressures in excess of 800 psig, even these control rods with incperable accu ulators vill le able to meet required scrar insertien times due to the acticn of reactor pressure. In additicn, they may le ncnsally inserted using the control-rod <J*ive hydraulic system. Procedural centrol vill assure that control rois vith incperable accu ulators vill be spaced in one-in-nine array rather than r.rouped together.
O E.
Eeactivity Anacalles Durin6 each fuel cycle excess operating reactivity varies as fuel depletes and as eg burnable poiscn in supplementary control is burn ~1.
Z e engnitude of this excess reactivity is indiceted by the integrated vorth of control rods inserted into the cere, referred to as the centrol rol inventory in the core. As euel burnup pregresres, enanalcus lehavicr in the excess reactivity may be detected by car:parison of actual rod inventory at any base equilibrium ccre state to predicted rod inventerv at that state. Fod inventory predictions can be normalized to actual initial steady state rol patterns to minimize calculation,1 uncertainties. Experience with other cperatin6 IME's indicates that the control rod inventory shculd be predictable to the equivalent of one per cent in rasetivity.
3 3A-3 BASES E6 4
7-m_
l l
l) 21 1
t 4
f4y oc r
e et a e
ri m.
r t
i op ot nfl i
t)
=ncniOe' c=yCt t
a nI a r
n ei a
r epaC a
e si t
el s ig skhon
.e~
Ed p
ret e
do h
yat oi=r h uet l
tO s e c u. teu eToha un O
b re s
h ht i oa eL ubtf t ss t
t t
h r(
s aoiye
.ioi sr i
eu sr' newt n o
t e
el i
rf D
rr I pt wy p'
uo Cic;rf hb es o
dt Ppa 1t e c '-
warey t s e
I f eii c
l osl cl I
ea l o tlt -
i eical sl.
l e eltt oeaf ra ea eae b r h a ner overu rcl a
t = h eg af =
y re r ;i et cahcn b
t en ena eh o
t, ceru' m
+
rt d =
h or pt t eairual oae ti e o
h nf os t nvl e t p u
pt ef cser e
oeb oao et u
eh s.
b kt yst rt -
e bs t cn t
s e
aso ii p
c sl ap r
r es rf
- u and eh r
g l vl eie t e tiidt co e
e a
t cb k
api i f h
vr ae i n agavet t
eoy wpo er arornyr-lt a st r
e ru: t ;<
o cm p
i r rnl psoir t
ra ued ue aoe st c rst re k t u eem ehn qt isI aua a a mt sC ap eia t
a o
rv eceP o a
t wes
=
,f t evLt ec an hy e
yorrdttN nt st h
ea sra on n
r weet s gre en t h not pst ne i f b -
b pt ane?
m uy ih yem i
u cs i i esn I
l re sf s eamt c
3 k
il' pet ai d eeldi hn fi rt eef n nrae ph s eet v.
u' y
I g
ev srdf t nes s
s
) i i aey uu-
- s sailt ms l rh sdl s-e ept ac 3
uvn r
f C
od rra o
I t g r
yl n e
- 4 ue et C
s n o g a c i d
.r neot s d
F yi et nr i
ift a o
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i pwev R
t wy h
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net l
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__7
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us r
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earA a
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i helo i
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peCih vr yt orfiE e= r t !
s s e
o i
stC m rt h y
r sci t ry-
. t i
R e,
pat Cs u
e C
ul c
m tl w
I s
veP
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lhI I af o t.
S erhyv eeo I
EI e
at Cl
. i.
tft se t e C
r v
fTce eE P
eoI h e l.
e s
l sfk
,l d
h fl lt eA.
,e ydfr y
c s
S seoen sl e e
t e c
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ah
,h anoufh Cayaie I mc u
i i
t t
l ne f ot.
Ilt wt m Cr n
i n
a eevs m
i y C oi ai Poy S
t i e n
r eyt eee t
Bscrit Hnt E
e n
gc o
y a i e s t,.
C i aet i
S o
rn t
eeermres f
I e
phiy eel A
C
- ai u
T R h pi h h y
. u E
Ti coI a Ttb h
C hsat nn hhi B
rs A
t slTt s s
5 es 3
a E
F B
6 i
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.t ibm w.o,
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.~~
e D3!IBIT B ( Cont )
30 LIMITIIiG COIrDITICIG FCE CPERATICI!
L.O ST_Ta'EIILL';CE EFRUIREEI;T3 rod or cafety/ relief valves shall every two rafueling cutages. The nominal tion of h2 be operable. The solenoid activated pcpping point of the safety valves shall relief functio 1 of the safety /relier val-be set as follevs:
ves shall te operable as required ty Spec-g ification 3 5.E.
I;urter of Valves Cet Point (psic) 2 s 1210 2
? 1220 2.
If specification 3 6.E.1 is not tet, ini-2.
a.
A minit= sf two safety /r-lief valves shall tlate an orderly chutdsvn and have coolant te tench checked or replaced with a tench pressure end temperature reduced to 110 checked valve each, refueling cut 360-All psig or less and 3L5 F or less within 2h four valves shall te checkel cr mplaced hours.
every two refueling outaces. The popping point of the safety / relief valves shall le set as follevs:
Iiurler of Valves Get Point (psig) k S 1060 h
b.
At least one of the safety /relier valv--
chall be disassembled and inspected each refueling cutage.
The integrity of t!.e safety / relief valve c.
tellows shall le continuously tenitored.
d.
'"he operatility of the tellows =Onitoring 3 6/h.6 119 e
.y.
g
.s, e 4 w n 4'
& ys s
J i.
A-1 IT B (Cbut)
,,,y p' - Q _',' n,'. g
~~-
t Leakaae
~
e The fomer 15 gpa Ifmf t for Icaks from unidentified sources was established assuming such leakage was caning
)
s' from the primary system. Tests have been conducted which demonstrate that a relationship exists between the size l
of a crack and the probability that the crack will propagate. Fran t5e crack size a leakage rate can be determined..
For a crack size which gives a leakage of 5 gps, the probabilit.y of rapid propsgation is less than 10-5 Rus, an unidentified leak of 5 gps when asstraad to be fras the primary system had less than one chance in 100,000 of propa-gating, which provides adequate margin. A leakage of 5 gp: is detectable and measureable. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period allowed for determination of leakage is also based on the low probability of the crack propagating.
The capacity of the drywell sump pumps is 100 gps and the capacity of the dryuell equipment drain tank pumps is also 100 gps. Removal of 25 gpm fras either of these sumps can be acca plished with considerable margin.
g The perfomance of the reactor coolant leakage detection system, including an evaluation of the speed and sensi-tivity of detection, will be evaluated during the first 18 months of plant operation, and the conclusions of this evaluation will be reported to the AEC. Modifications, if required, will be perfomed during the first refueling outage after AEC review. In addition, other techniques for detecting leaks and the applicability of these techniques to the Monticello Plant will be the subject of continued study.
E.
Safety and Relier Valves Experience in safety valve operation shows that a testing of 50% of the safety valves per reibeling outage is l
adequate to detect failures or deterioration. A tolerance value is specified in Section III of the ASME Boiler and l
Fressure Vessel Code as +1% of design pressure. An analysis has been perfomed which shows that with all safety i
valves set 1% higher than the set pressure, the reactor coolant pressure safety limit of 1375 psig is not exceeded.
I Safety / relief valves are used to minimize activation of the safety valves. The operator will set the pressure j
settings at or below the settings listed. However, the actual setpoints can vary as listed in the basis of Specification 2.4.
g 5 _; -=+~ e H.1 x L ib_ p uccu
- tti 6s at cr t-
- lc= rett "cr li-t {
h _- r,
m -ami.e -v,1. tm
--- -- ; -- l i - te d ir % i;;is cf Opccific-tic: 2.h.
The required safety valve steam flow cap city is gt,egi, ped by analyzing the pressure rise accmpanying the main steam flow stoppage resulting from a
- __ @hapu%e. :
' b :
i-2:1 with the reactor at 1670 MWt. The analysis assumes =^ rt: = b g :: cy-t m Scw no t w e
_np scram, but a reactor screen frco indiract means (high ir^ n flux). The relief and safety valve capacity is asstEed to total 50% (35% relief and 15% sarety) of the full power steam generator rate. This capacity corresponds to assuming that three of the four relief / sfaty valves (35 k%)
and two of the four safety valves (18.h%) operated. For additonal margin three safety and
.. _ safety /relier l
valves are required to be operable.
I l
134 3.6/4.6 BASES 1
a
.9
I r
AEC PE3??.r2IC:t FC3 FART 50 DOCCT !/.'
'IAL (EFJ0;WiY 70?.'1) 00!iTROL : Of 16B9 TIIJ' i~RW.:
DAT ' 0F LOC:
D/sh h:,C 'D L'iM isi;0 RP2 Os.-;En I
Northern States Power Cor_panu Minneapolis, Minn.
55401 t, c. "2ver 6-1-73 6-7-73 X
WI ORIO CC UiMER CEliT AEC FD3 v
EEiT LOCAL PLa x
Mr. f ie ry 3 Dir.ned
~CusSS:
Q/O,or lirio ItiFvT
- 0 cia ascia pag.g;T ;,o
X 40 50-263 DI;sCaliTIC.it 2.:iCL0s;rAEs:
Ltr trano the folloving:
Request for change to Tech Specs, notarized 6-1-73, for the Monticello thiclear cenerating Plant.
i 50 Not Remove py,.;.;i ;xg3;
- enticello
( 40 cya rec'd)
F03 he.'Io.U ;...0.n VM 0:
~
6_g.
w BUTLEE(L)
OC11Ei 0ER(L) g,,,,, ZII2 LU N( L )
YOU;;GBLCOD(E) j W/ Copica V/ Copies W/ 9 Copies W/ Copico CLiD:(L)
CTCIO(L)
ROUSE (F1) 1;LGAli(E)
{
W/ Copico W/ Copies W/ Corien W/ Copics p
COLLER(L)
ViC3iLLO(L)
DICT.ER(E) j V/ Copico W/ Ccpics W/ Copics W/ Copics
' KNII'L(L)
SCIH2 '.EL(L) 13TIGilN:I(E)
W/ Copice U/ Ocpies U/ Copics W/ Copico I! fEL"iL DICTRILtJI'IO71 REG FIII )
TECTI REVIEW DE r:C'1 F&M WADE E
.nu.v m H.fi ORI!'E3 Sla12Y DE0WN E
p OGC, B00:4 P-50s.i CCPn03ER GA!'A LL 1:U3SHiU'ER G. VILLI A".3 E
WiUlrTZII G/DT.'J'F MACCARY KAGT:'ER SHEPPARD E
l
CASE 10TI G"T RiLLiED LIC /.S3T.
6 OIAMBUSSO PAULICTI SPA 1:GLER SEhv1CE L
A/TIMD b BOYD CEiO WILGO:1 L
LI%ITP.i i e V. MOCPE-L(EWR)
STELID E!UIRO GCULIQURNE L SALTZMI
, Q DEYOUI G-L(P;5t)
I!OUOTCH iCLIAR SMITI!
L SK07I10LT-L NOVAK DICEER GEiRIN L
PIAN 3 kr]?. COIJLI::S ROOS KIIGHTC::
DIGG3 L
ITDDRALD IPPOLITO YOUI:3LL%D TLU S L
- DUBE p REO OF71-TFDi'GC0 ITGAN LEE L
y FILL te RECIO:t( 3)
Ih::o rnoJ LEADER wi1 GRIN' L
It!PO
{ MORRIS LiIn\\S SEAFER F &. M C. MIIIS j['.TEELE D
BEU'.SCYA lbhLLS 2 VOLIMER V ATmi CART T.
L':'*! T'! a DT ^~T -J;T ON AL PDR__
uinneanolis. Minn.
'DE(IC:VII'Y)
(1)(E)(9)-NiTIO::AL IAB'3 1-PDR-SAN /LA/tY
-1GIC(LTnc :)
1-R. CirTOLL C, CT-2li:
. N L3-YCIG/CAYRE 1-R. CATLIII,E-256-GT' 7 1-GELLD LEL'EUOHE ERCCXWsVEN NAT. LAB
[CYS ACis m+vk SE!rf TO LIC ASST.
0 WARD /i! GT.
1-Cc::suLT.u: tis 1-Acxrt(WALTER KOILTER,
!!EW%D:/BLUME/AGABInN E C-427, OT)
R. DIocS on 6-8-73 1-GERIs.D ULnInsen....On:L
- 1. np...lmJmR...T-3030 I
.