ML20003J079
| ML20003J079 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/23/1981 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20003J077 | List: |
| References | |
| TAC-43369, NUDOCS 8105080414 | |
| Download: ML20003J079 (26) | |
Text
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UNITED STATES y%
NUCLEAR REGULATORY COMMISSION g..
,y WASHINGTON D.C.20H4 SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SECO NUCLEAR GENERATING STATION l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.33 License No. DPR-54 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Sacramento fiunicipal Utility District (the licensee) dated March 13, 1981, as supplemented April 10 and 17,1981, complies with the standards and require-cents of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; 1
l D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; i
and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requiremer.ts have been satisfied.
l 61050804lq
~ Accordingly, Facility Op(erating License No. DPR-54 is hereby amended 2.
fications as indicated in the attachment to this license amendment:
2.C.(2) Technical Specifications The Technical Specifications centained in Appendices A and B, as revised through Amendment No. 33, are hereby incorporated in the license. The licensee shall operate the i cility in accordance with the Technical Specificacions.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY C0t94ISSION i 464 Jo F. Stolz, Chie 3 erating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 23,1981
ATTACHMENT TO LICENSE AMENDMENT NO. 33 FACILITY OPERMING LICENSE NO. DPR-54 DOCKET NO. 50-312 Revise Appendix A as follows:
Remove Pages Insert Pages Figure 2.1-2 Figure 2.1-2 2.1-3 2.1-3
- 2. 3-1
- 2. 3-1
- 2. 3-2 23-2 3.5.2-1 3.5.2-1
- 3. 5. 2-2 3.5.2-2 3.5.2-3 3.5.2-3 3.5.2-4 3.5.2-4 3.5.2-5 3.5.2-5 3.5.2-6 3.5.2-6 3.5.2-7 3.5.2-7 3.5.2-8 3.5.2-8 3.5.2-9 3.5.2-9 3.5.2-10 3.5.2-10 3.5.2-11 3.5.2-11 3.5.2-12 3.5.2-12 2-3 2-3 2-5 2-5 2-6 2-6 2-7 2-7 2-9 2-9 5-4 5-4 5-5 5-5 Changes on the revised pages are shown by marginal lines.
l
Figure 2.1-2 CORE PROTECTION SAFETY LIMITS, REACTOR POWER IMBALANCE (CYCLE 5) 1 THERMAL POWER LEVEL, %
-.120 112 CURVE 1 (35,112) l
(-34,112)
. iiu J
IACCEPTABLE I
-100 l
l4 PUMP
.0 "I = 1.69 IOPERATION l
"2
= -
88.65
. 90 CURVE 2 (46,90)
I
(-47,90) l l ACCEPTABLE
,l I
l 1
l4 s 3 PUMP
. 80 l
I l
l OPERATION l
I I
I I
I(46,66.65)
- 70 i
1
(-47,66.65)
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61.51 CURVE 3
. ou I
l ACCEPTABLE l
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I4, 3, & 2 PUMP l
50 l
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Reactor Power imbalance, %
l CURVF.
REACTOR COOLANT DESIGN FLOW, GPM ji 387,600 l2 288,374 3
187,986
$SMUD SACRAMENTO MUNICIPAL UTit.lTY DISTRICT Amendment No. 7/, h, }d,, 33
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l FIGURE 2.1-3 Core Protective Safety Sases.
'2400 2
3
" 2200- -
2.
E E:
1 2000- -
O 5
2 3 1800- -
t l
1600 560 580 600 620 640 Reactor Outlet Temperature, F Pumps operating Curve Reactor coolant flow, cem Power,%
(tvoe of limit) 1 387600 112 Four (DNBR Limit)
Bree (DNBR Limit) 2 288374 88.65 One in each loop 3
187986 61.51 (queIity iImlt) l AmendmentNo.((,33 h
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Figure 2 3-1 Protet.tive System Maximum Allowable Setpoints, Pressure Vs Temperature 2600 2400- -
P = 2300 psig T - 618 3
Acceptable Operation E
2200- -
2 4
4 j
Unacceptable Operation f
n 6
2000-S
.O.
{
P - 1900 psig g
1800 l
l 540 560 580 600 620 640 Reactor outlet Temperature, F i
RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS AmendmentNo.)hff.33.
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Rod Index Vs Power.l.evel FIGURE 3 5 2-1
(
Rancho Seco Cycle 5 l
0 - 60 EFPD l
Four Pump Operation 110 l
l 105 (197,102)
(288.102) l 100 l
j 95 o.92)
I 90 f
85 (270,8a) f 80 operation l
75 Not Allowed f
70 "_
l 65 1
69 Restricted f
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(122,50) g 50 45 a.
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l 30 l
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15 (60,15) l i
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25 50 75 300
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0 25 50 75 100 Bank 6 l
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$SMUD SACRAMENTO MUNICIPAL. UTILITY DISTRICT Amendment No. J6',JAf, 33
l FIGURE 3 5.2-2 Rod index Vs Power Level Rancho Seco Cycle 5 50 - 250 EFPD Four Pump Operation 110 105 (259,102)
(280,102) f 100 92)
!(280, l
95 -.
l 90 f
85 Operation Not Allowed 270,80 I
)
f 75 g
70
.t' 65 E
f 60 55 l
(196,50)
(232,50) l l
50 45 l
a.
40 l
35 j
30 1
I 25 Operation 20 15 (118,15)
Allowed f
(0,6.6)10 5
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0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 f
Rod index l
Bank 7 Bank 5 1
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25 50 75 100 0
25 I
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25 50 75 100 Bank 6 i
&) SACRAMENTO MU SMUD
_ Amendment tio. X ) g 33 i
_. _..... ~ _ -...,... _..... _ _.. _ _.,., _ -.. _. - _ _.. _.. -.. _. _..... _
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FIGURE 3.5.2-3 Rod index Vs Power Level Rancho Seco Cycle 5
Four Pump Operation 110 105 (244,102) j(273,102) 100 95 (267.92' 90 p
.f (240,80) 85 80 Ope ra c ion 4
"
- AII **d d
75 70 65 60 n
55 (174.50) o 50 d
45 40 operation 35
^II **d 30 )
25
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0 1 0,0) i i
i i
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i i
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20 40 60 80 100 120 140 160 180 200 220 240 260 280 300-Rod Index Bank 7 Bank 5 I
I i
1 1
1 I
I I
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25 50 75 100 0
25 50 75 100 t
1 i
f I
o 25 50 75 100 Bank 6 t
$NNU SACRAMENTO MUNICIPAL UTILITY DISTRICT Amendment flo.g)9, 33
FIGURE 3 5.2-4 Rod index Vs Power Level Rancho Seco Cycle 5 0 - 60 EFPD 3 Pump Operation 110 (270,10r) 105
,197,102)
(
100 95 90 Operat;on 85 Not Allowed 80 75 70 Restricted (232,64) 65 60 55 (122,50) e 50 2
45 40 operation 35 Allowed 30 25 20 (60,15) 15 10 (0,6.6) 5 [
0 I
I t
i i
e i
I I
I I
I 8
0 20 40 60 80 ~100 120 140 160 180 200 220 240 260 280 300 Rod Index sank 7 f
Bank 5 I
I
' l I
J l
0 25 50 75 100 0
25 50 75 100 t
I t
t
]
1
?
t 1
i 0
25 50 75 100 Bank 6 l
$SMUD SACRAMENTO MUNICIPAL UTit.lTY DISTRICT Amendment No. X % 33 r
w w g.,.._
FIGURE 3 5.2-5 Rod index Vs Power t.evel
. Rancho Seco Cycle 5 50 - 250 EFPD 3 Pump Operation 110 105 (259,102)
(270,10 !)
100 I
95 l
90 l
85
[
80 J
operation d
75 Not Allowed l
70 (232,67:)
f 65 60 1
as f
55 (196,50) 50 I
o f
d 45 l
40 f
35 l
30 Operation f
25 Allowed l
15 (118.15)
(0,6.6) 0 e
e i
e i
e i
i e
i e
i i
i 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index-l Bank 5 Bank 7 I
f f
i i
f I
f f
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25 50 75 100 0
25 50 75 100 t
i I
t t
0 25 50 75 100 Bank 6 g
uuo SACRAMENTO MUNICIPAL UTILITY DISTRICT Amendment No. Mg 33
FIGURE 3 5 2-6 Rod index Vs Power t.evel Rancho Seco Cycle 5 230 EFPD - EOC (APSR Out) j 3' Pump Operation 110 105 (244,102) 100 95 90.,,
85 80 75 Operation Not Allowed 70 65 60 s
55 (174,50)
=
50 d
45 40 35 30 Operation j
25 AII **d l
20 (96,15)
S 15 10 l
5 (0.4) i e
i i
e i
e i
i 0
0 20 40 60 80 100 120 140 160 180 Zoo 220 740 260 280 300 Rod index Bank 7 I
I I
i i
1 I
t t
l Bank 5 0
25 50 75 100 0
25-50 75 100 f
f 1
1 f
0 25 50 75 100 i
Bank 6 I
$SMUD SACRAMEN^n MUNICIPAL UTILITY DISTRICT Amendment No. M g 33
a FIGURE 3 5.2-7 APSR Withdrawal Vs Power level Rancho Seco Cycle 5 APSR Position Limits (0-60 EFPD).
110 (6,102)
-(29,102) 100 (6,92)
(30,92)
RESTRICTED REGION 90 80 (0,80) 7,80) 70 5
m 3
60 (100,50) 50 O
PERMISSIBLE OPERATING 40 j
REGION
-g*
8 a.
30 P
20 10 0
I I
I I
I I
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I I
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10 20 30 40 50 60 70 80 90 107 110 APSR withdrawal, %
i l
($SMUD SACRAMENTO MUNICIPAL UWW DISMICT AmendmentNo.g,g,33
l FIGURE 3 5 2-8 APSR Withdrawal Vs Power Level l
Rancho Seco Cycle 5 l
APSR Position Limits (50-250EFPD) 110 (6.102)
(32,102)
RESTRICTED 100 REGION 90 (0,80) 45,80) 80 70
'5m3 60 (100,50)
~
50
-o PERMISSIBLE OPERATING 40
- q REGION l
30
~
20 10 0
O 10 20 30 40 50 60 70 80 90 100 110 i
APSR withdrawal, %
1
/
gSMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT t
FIGURE 3 5.2-9 Core imbalance Vs Power Level Rancho Seco Cycle 5 0 - 60 EFPD Imb'alance Envelope 110
~
(-13,102)
(23,102) 100 RESTRICTED REGION
(-14,92)
(23,92) 1 90 80
-26,80)
(26,80) 3 70 1
s 3
}
60
/(-50,50)
(50,50) 50 e
PERMISSIBLE D
40 OPERATING
[,,
REGION 30 20 10 i
i 1
1 L I I
O 40 20
-10 0
10 20 30 40 50 Core imbalance, %
l
\\
$SMUD SACRAMENTG MUNICIPAL. UTILITY DISTRICT knendment No.,M,% 33 n.,
FIGURE 3 5 2-10 Core imbalance Vs Power Level Rancho Seco Cycle 5 50 - 250 EFPD (APSR In)
Imbalance Envelope 110
(-18.102)
(73,102) 100 RESTRICTED REGION
(-21,92)
(23,92) 80
-32,80)
(26,80) l 70 3
~
60
(-50,50)
(50,50) 50 e
PERMISSIBLE
. 'g 40 OPERATING REGION 30 20 10 0 40
-30 10 0
to 20 30 40 50 Core imbalance, %
i l
l l
l l
4 g' SACRAMENTO MUNICIPAL SMU-Amendment No. M M 33 l - -......
FIGURE 3 5.2-11 Core imbalance Vs Power Level l
Rancho Seco Cycle 5 23.0 EFPD - EOC (APSR Out)
Imbalance Envelope l
~
110
(-27.102)
(14,102)
I 100 RESTRICTED REGION
(.28,92)
(18,92) l 90 l
(-28,80)
(28,80) 80 l
l f
70 I
g 60 PERMISSIBLE
~
o i
OPERATING
._(32,50)
'(-50,50)
% 50 REGION ee I
- 40 3
l l
30 0
THIS FIGURE VALID ONLY FOR OPERATION AFTER APSR WITHDRAWAL 10 8
8 I
I I
i t
i t
I I
0 50 30 10 0
10 20 30 40 50 60-Core imbalance, %
l l
l 1
i
$SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT Amendment flo. )tf, g 33 s
l
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS i
Figure 3.5 2-12 LOCA Limited Maximum Allowable Linear Heat Rate l
21 i
i 20 i
C 19 s
5 18 1
[% N 1
BALANCE OF CYCLE J
16 s
FIRST 50 EFPD-r
/
/
=
E 14 l
l 13 12 0
2 4
6 8
10 12 Arial location of Peak Power f rom Bottom of Core, f t
$SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT Amendment flo.M 33
l RANCHO SECO UNIT 1 TECHNICAL. SPECIFICATIONS Safety Limits and Limiting Safety System Settings l
pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the left of the four-pump curve will be above and to the left of the other curves.
The maximum thermal power for three-pump operation depicted in Figure 2.1-2 is 88.65 percent due to a power level trip produced by the flux-flow ratio 1.06 times 74.4 percent design flow = 78.86 percent power plus the maximum calibration and instrumentation error. The maximum thermal power for other coolant pump conditions is produced in a similar manner. The actual maximum power levels are calculated by the RPS and will be directly proportional to the actual flow during partial pump operation.
A thermal margin credit equivalent ta 1% DNBR to offset the rod bowing penalty has been used as a result of'the flow area (pitch) reduction factor included in the hot-channel thermal-hydraulic analysis. The 1%
DNBR credit was approved (Ref. 4) and is the only credit applied to offset the rod bow penalty.
References (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March, 1970.
(2) Rancho Seco Unit 1, Cycle 2 Reload Report BAW (3)
Rancho Seco Unit 1, Cycle 3 Reload Report BAW-1499, September, 1978, l
I (4)
D. F. Ross and D. G. Eisenhut memorandum for D. B. Vassalls and K. R. Coller, " Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors, dated December 8, 1976.
1 Amendment No. JC X 33 2-3 f
s RANCHO SECO UNIT 1 TECHNICAL SPECIFIC TIONS Safety Limits and Limiting Safety System Settings -
- ~
2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to Instruments monitoring reactor power, reactor power Imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high Reactor Building pressure.
Objective To provije automatic protectirsn action to prevent any combination of process variables from exceeding a sa'ety limit.
Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypass for the instrument channels shall be as stated in table 2.3-1 a.nd figure 2 3-2.
Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.
The trip setting limits for protection system instrumentation are listed in table 2.3-1.
The safety analysis has been based upon these protection system instrumentation trip set points plus calibration and instrumentation errors, Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to.be
~
detected by pressure and temperature measurements.
During normal plant operation with all reactor coolant, pumps operating, reactor trip is initiated when the reactor power level reaches 104.9 percent.
l 2-5 Amendment No. 33 4 g4
-em y.,
.g. -,..,
.n..
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings Adding to this the possible variation in trip set points due of rated power.
to calibration and instrument errors, the maximum actual power at which a trip wouldbeaggyatedcouldbe112 percent,whichwasusedinthesafety analysis.
A.
Overpower trip based on flow and imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been estab-lished to accommodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power.
The analysis in section 14 demonstrates the adequacy of the specified power-to-flow ratio.
The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power' level increases or the reactor coolant flow rate The power level trip set point produced by the power
', decreases.to flow ratio provides overpower DN8 protection for all modes of pump operation. For every flow rate there is a maximum permissible low flew rete. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are as follows:
Trip would occur when four reactor coolant pumps are operating t
l 1.
If power is 106 percent and reactor flow rate is 100 percent,
~
or flow rate is 94.34 percent and power level is 100 percent.
Trip would occur when three reactor coolant pumps are operating 2.
if power is 78.8 percent and reactor flow rate is 74.4 percent or flow rate is 70.75 percent and power level is 75 percent.
3 Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) If the power is 51.4 percent and reactor flow rate is 48.5 percent or flow rate is 46.22 percent and the power level is 49 percent.
.l l
For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.
The power-imbalance boundaries are established in order'to prevent reactor"
~
thermal limits from being exceeded. These thermal limits.are either power peaking kW/ft limits or DNBR limits. The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of The power-to-flow ratio reduces the power level Figure 2.3-2 are produced.
trip and associated reactor-power reactor-power-imbalance boundaries by 1.06
{
percent for a 1 percent flow reduction.
d l
Amendment No.,/. 33 2-6.
l
-__,,,_,_..,....,_,._,,.,._.,_.m
,.m.-
s RANCHO SECTO UNIT 1 TECHIIICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings B.
Pump monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to (a) the loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation. The pump monitors also restrict the power level to 55 percent for one reactor coolant pump operation in each loop.
C.
Reactor coolant system pressure During e startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit shown in figure 2.3-1 for high reactor coolant system pressure (2300 peig) has been established to maintain the systes pressure below the safety limit (2750 paig) for any design transient (1) and minimize the challenges to th,e EHOV and code safeties.
The low pressure (1900 paig) and variable low pressure (12.96 T
- 5834) ong trip set point shown in figure 2.3-1 have been established to maintain the DNE ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction.
(2,3)
Due to the e'alibration and instrumentation errors the safety analysis used a variable low reactor epolant system pressure trip value of (12.96 T
- 5884),
out D.
Coolant outlet temperature The high reactor coolant outlet temperature trip settinC ltaic (618 F) l shown in figure 2.3-1 has been established to prevent excessive cora l
coolant temperatures in the operating range. Due to calibration and
~
instrumentation errors, the safety analysis used a trip set point of 620 F.
E.
Reactor Building pressure The high Reactor Building pressure trip setting limit (4 peig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the Resctor Building or a loss-of-coolant accadent, even in the absence of a low reactor coolant system l
pressure trip.
F.
Shutdown bypase In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in 2-7 Amendment No. )4( g 33 l
l l
w-n r
TAat.2 2.3-1 REACTot P90TECTICII ST57T.M TRtr SETTING LIMITS One Beacter Coolent Pump Four Beactor Coolmat Funpa Three taacter Coolant rumpe Operatses in Each 1mer Shutdove Operating (Nesteet Operettog (Neetaal (hoelmal Operstleg typase operating Fewer - 1003)
Operettes Fewer - 75I)
Power - 492) 1 1.
N ciear power et ra w, man.
104.9 104.9 104.9 5.0(3)
(2) l 2.
se.cseer power bee em flew
- 1. 06 t ie= = f lew = t eu s 1.06 :smee it w stou.
1.06 tsees flow sta.e sypassea a I febalance,1 of tated, nea.
reduccles due to resecties due to redocties due to tobalance(s) aokalance(s) imbalance (s) 3.
Nuclear power baseJ em pump meeleurs, t of rand,.a.
4 Nigh reactor costant 1820 'I system pressure, peig. eaa.
2300
'2300 2300 S.
I.eie reacter coetent ersten 1900 1900 1900 sypassed pressure, pelg, ets.
6.
Variable low reacter teelmet 12.96 7 - 5434 13.94 7 - 5434 12.94 7 - 5414 spassed out owl est I
erstere pressure, pols, ste.
7.
aeoctor coeneet top. r., sea.
618 618 618 618 z
i i
B.
Nigh teacter Bulldteg 4
4 4
4 pressere, pels, maa.
(1) T,g te la degrees lettenhalt (F).
(2) Seatter coolant systee fl.as. 1.
()) AJetnistratively esstralled reducties set sely Jertog reacter sleutdows.
D (4) Automatically set when staar eageests of the RF5 (se specified) are bypassed.
(S) The pump monsters alas produce a trip oss (a) Isas et two reacter costaat pianpa la ese reacter coelaat leep, e
and (b) lose of see er too reacter coolset pumps during two-ptap operaties.
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M Amen iment No.,J(, E, K,33 1"""
RANCHO SECO UNIT 1 TECHNICAt. SPECIFICATIONS Design Features 5.3 REACTOR Specification 5.3.1 Reactor Core 5.3.1.1 The reactor core contains approximately 93.1 metric tons of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Eachfuelassemblycontains208fuelrods.(o{(2)
The reactor core is made up zircaloy-4 tubing to form fuel rods.
l 177. fuel assemblies.
The reactor core shall approximate a right circular cylinder with an 5.3 1.2 equivalent diameter of 128.9 inches and an active height of 144 inches.(2) 5.3.1.3 The average enrichment of the in I core for Rancho Seco is a nominal 2.57 weight percent of U Three fuel enrichments a;e used in the initial core.
iod assemblies (CRA) ar.d 8 axial 5.3.1.4.There are 61 full-length contro?
power shaping rod assemblies (APSRA) distributed in the reactor core as shown in FSAR figure 3.2-45 The full-length CRA contain a 134 Inch length of silver-indium-cadmiumalloy clad with stainless steel.
The APSRA contain a 36 inchgngth of silver-indium-cadmium alloy clad with stainless steel.
- 5. 3.1. 5 The initial core will have 68 burnable poison assemblies with similar dimensions as the full-length control rods. The cladding will be z'.rcaloy-4 filled with aluminum oxide-boron carbide pellets and placed in the core as shown in FSAR figure 3.2-2.
5.3.1.6 Reload fuel as'semblies and rods shall conform to deslor, nad eval-uation described in the FSAR. A reload core may also have burnable poison assemblies with' dimensions similar to the full-length control rods with materials as specified in 5.3.1.5.
5.3.2 Reactor Coolant System 5.3.2.1 The reactor coolant system shall begsigned and constructed in accordance with code requirements 5.3 2.2 The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and pressure, shall be designed for a pressure of 2,500 psig and a shall be designed for a temperature of 670 '.Igurizer surge line temperat;ure of 650 F.
The pressurizer and ore 0
5 3 2.3 The reactor coolant system volume shall be less than 12,200 cubic feet.
Amendment No. g 33 5-4
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Design Features REFERENCES (1) FSAR table 3.2-1 (2) FSAR table 3.2 (3) FSAR paragraph 3.2.4.2 (4) FSAR paragraph 4.1.3 (5) FSAR paragraph 4.1.2 i
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t Amendment No.,#,33 5-5 k
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