ML19326C612
| ML19326C612 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/13/1977 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19326C607 | List: |
| References | |
| NUDOCS 8004230712 | |
| Download: ML19326C612 (21) | |
Text
~
DNBR of 1.3 corresponds to a 95 percent probability at a 95 percent confi-dence level that DNB will not occur; this is considered a conservative mar-gin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits.
The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip set points to correspond to the elevated location where the pressure was actually measured.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.3 is predicted. The curve is the most restrictive com-bination of 3 and 4 pump curves, and is based upon the maximum possible thermal power at 106.5% design flow per applicable pump status.
This curve is based on the following nuclecr power peaking factors (2) with potential fuel densification effects; N
N N
Fq = 2.67; Fan = 1.78; F = 1.50 These design limit power peaking factors are the most restrictive calculated full power for the range from all control rods fully withdrawn to maximum at allowabic control rod insertion, and form the core DNBR design basis.
The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowing:
1.
The 1.3 DNBR 1. mit produced by a nuclear power peaking factor of F{ = 2.67 or the combination of the radial peak, axial peak and position of the axial peak that yields no less than 1.3 DNBR.
2.
The combination of radial and axial peak that prevents central fuel melting at the hot spot. The limit is 19.4 it.
Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.
The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.
The curve of Figure 2.1-1 is the most restrictive of all possibic reactor coolant pump maximum thermal power combinations shown in Figure 2.1-3.
The curves of Figure 2.1-3 represent the conditions at which a minimum DNBR of 1.3 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of mini-mum DNBR is equal to 22 percent (1), whichever condition is more restrictive.
8004230 7 4
Using a local. quality limit of 22 percent at the point of minimum DNBR as a basis for' eurve 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.
The DNBR as calculated by the BAW-2 correlation continually increases from
' point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.
The maximum thermal power for three pump operation is 86.0 percent due to a power level trip produced by the flux-flow ratio (74.7 percent flow x 1.065=
79.6 percent power) plus the maximum calibration and instrumentation error.
The maximum thermal power for other reactor coolant pump conditions is pro-duced in a similar manner.
For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pump situation.
Curves 16 2 of Figure 2.1-3 are the most restrictive because any pressure / temperature point above and to the left of this curve will be above and to the left of the other curve.
REFEPENCES (1) Correlation of Critical ifeat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, May, 1976.
(2) FSAR, Section 3. 2.3.1.1.c 9
~
2600 2400 E.
2200 b
ACCEPTABLE m
OPERATION E
2000 o
O UNACCEPTABLE OPERATION 1800 1
~
1600 560 580 600 620 640 660 Reactor Outlet Temperature
- F ARKANSAS POWER & LIGHT CO.
I. NO.
ARKANSAS NUCLEAR ONE-UNIT 1 CORE PROTECTION SAFETY LIMIT 2.1 -1 9a
l THERMAL POWER LEVEL, 5 UNACCEPTABLE OPERATION 120 Kw/ft LIMIT ACCEPTABLE 4 PUMP OPERATION 100 CD ACCEPTABLE 3g4 80 PUMP OPERATION a
60 ACCEPTABLE 2,3,& 4 PUMP OPERATION y
40 20
- MORE LIMITING THAN CALCUL ATE 0 ONBR OR KW/FT LlylTS I
I I
I i
i 60
-40
-20 0
20 40 60 Reactor Poner imoalance CURVE REACTOR COOLANT FLOW (gpm) 1 374,880 2
280,035 l
3 184,441 l
ARKANSAS POWER & LIGHT CO.
FIG. NO.
ORE PROTECTION SAFETY LIMITS ARKANSAS NUCLEAR ONE - UNIT 1 2.1-2 9b
2600 2400 E
2200 f
a I
2000
'h 1800 f
f 1600 560 580 600 620 640 660 Reactor Outlet Temperature. F CURVE GPM POWER PUMPS OPERATING (TYPE OF LIMIT) 1 374,880 (1005)*
1125 FOUR PUMPS (DNBR LIMIT) 2 280,035 (74.75) 86.75 THREE PUidPS (DNBR 1.lMIT) 3 184,441 (49.25) 59.05 ONE PUMP IN EACH LOOP (OUALITY LIMIT)
- 106.55 0F DESIGN FLOW ARKANSAS POWER & LIGHT CO.
FIG. NO.
CORE PROTECTION SAFETY LIMITS ARKANSAS NUCLEAR ONE UNIT 1 2.1-3 9e s
^
s 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
' decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pu=p operation.
For every flow rate there is a maximta permissible power level, and for every power level there is a minimum permissible low fit a rate. Typical power level and low flow rate combinations for the punp situations of Table 2.3-1 are as follows:
1.
Trip would occur when four reactor coolant pumps are operating if power is 106.5 pcreent and reactor flow rate is 100 percent or flow rate is 93.9 percent and power level is 100 percent.
2.
Trip would occur when three reactor coolant pumps are operating if power is 79.5 percent and reactor flow rate is 74.7 percent or flow rate is 70.4 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 52.3 percent and reactor flow rate is 49.2 percent or flow rate is 46.0 percent and the power level is 49.0 percent.
The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.
The penalty in reactor coolant flow through the core was taken for an open
~
core vent valve because of the core vent valve surveillance program during each refueling outage.
For safety analysis calculations the maximum cali-bration and instrumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor thermal ibnits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNBR limits. The reactor power imbalance (power in top half of core minus power in the bottom _ half of core) reduces the power level trip prcduced by the power-to-flow ratio so that the boundaries of Figure-2.3-2 are produced. The power-to-flow ratio reduces the power level trip associated reactor power-to-reactor power imbalance boundaries by 1.065 percent for a 1 percent flow reduction.
B.
Pump monitors In conjunction with the power imbalance / flow trip, the pump moni-tors prevent the minimum core DNBR from decreasing below 1.3 by trip-
. ping the reactor due to the loss of reactor coolant pump (s).
The pump monitors also restrict the power level for the number of pumps in operation.
C.
During a 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 12
. shown in Figure 2.3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety Ibnit -(2750 psig) for any design transient. (2)
The low pressure (1800 psig) and variable low pressure (11.75 Tout
-5103) trip setpoint shown in Figure 2.3-1 have been established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction.
(2,3)
Due to the calibration and instrumentation errors,the safety analysis used a variable low reactor coolant system pressure trip value of (11.7Ffout-5143).
D.
Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (619F) shown in Figure 2.3-1 has been established te prevent ex-cessive core coolant temperatures in the operating range.
Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620F.
E.
Reactor building pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
F.
Shutdown bypass
, 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 Table 2.3-1.
Two conditions are imposed when the bypass is used:
1.
A nuclear overpower trip set point of <5.0 percent of rated power is automatically imposed during reactor shutdown.
2.
A high reactor coolant system pressure trip set point of 1720 psig is automatically imposed.
The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection system bypassed. This high pressure _ trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated.
The overpower trip set point of <5.0 percent prevents any significant reactor power from being produced when performing the physics tests. Sufficent natural circulation (5) would be availabic to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.
13
t 2500 P = 2355 psig T = 619*F ca 2300 E.
ACCEPTABLE OPERATION N
S a-2100 P = 11.75 Taut' O
5103 psi g m
1900 UNACCEPTABLE OPERATION P = 1800 psig 1700 1500 560 580 600 620 640 660 Reactor Outlet Temperature, F
ARKANSAS POWER & LIGHT CO.
PROTECTIVE SYSTEM MAXIMUM FIG. NO.
ARKANSAS NUCLEAR ONE-UNIT ~1 ALLOWABLE SET PO!NT 2.3 1 14a s
THERMALPOWERi.EVEL,",
- - 120 UNACCEPTABLE OPERATION (106.5) h W
100
+,
//
%\\
ACCEPTABLE 4 y
PUMP OPERATION
'8/p (79.5) 80 ACCEPTABLE 3& 4 PUMP OPERATION 60 (52.3)
A CCEP T A BL E 2.3
& 4 PUMP OPERATION 40 20 h
l I"
I" l
l
-60
-40
-20 0
20 40 60 Power imbalance, ",
ARKANSAS POWER & LIGHT CO.
PROTECTIVE SYSTEM MAXIMUM FIG. NO.
ARKANSAS NUCLEAR UNE - UNIT I ALLOWABLE SETPOINTS 2.3-2 14b e
Table 2.3-1
(.
. Reactor Protection Systen Trio Setting Licits
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One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Purps Operating in Each Loop Operating (Nocinal Operating (No:ninal (Nominal Operating Operating Power - 1005.)
operating Power - 751,)
Power - 491)
Shutdown sypas s
. Nuclear power, i of.
105.5
- 105.5 105.5 5.0(3) rated, max iiuclear sower based on 1.065 times flow minus 1.065 times flow uinus 1.065 times flow minus
. flow (2) ;and imbalance, reduction due to reduction due to reduction due to a assed
% of rated, r.ax i=bal ance (s) imbalance (s) imbalance (s)
. Nuclear power based c,n NA NA putp conitors, i of 5 5 *.
Bypass rated, rax (4)
. - Ifigh reactor coolant 2355 2355 2355
'1720(3) systen pressura, psig,
' max
. Low reactor coolant sys-1300 1800 1800 tem pressure, psig, min Bypassed Vartable low reactor (11 75 Tout-5103)(1)
(11. 75 Tout-5103) (1)
(ll. 75 Tout-5103) (1) coolant system pressure, Bypassed.
psig, min Reactor coolant t e mp,
619 619 619 F, nax 619 liigh reactor building 4 (18.7 psis) 4(18.7 psia) 4 (13.7 psis) 4(18.7 12) pressure, psig, rax (1) T is in degrees Fahrenheit (F).
(3) Automatically set when other segncnts of the RPS (as specified) are bypassed.
out (2) - Reactor coolant sys te:s flow, S.
(4) The purep monitors also produce a trins on: (a) loss of two reactor coolant G
pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant 2
, pumps during two-punp operation.
,r h 6
56.l'.3 B6W-2 DSM Correlation r
The~D&W-2 DSB correlation, a realistic prediction of the b'urnout phenomenone,3, has been reviewed and approved for use with the Lirk B fuc1 assembly design.
In applying this correlation to the ANO Unit 1 Cycic 2 core, two modifications, which have also been applied to other B6W reload report submittals, have been used:
~
- 1.. Thc limiting design DNBR of 1.30 was used, corresponding to a 95%
probability of a 95% confidence icvel that DNS will not occur.
2.
The pressure range applicabic to the correlation has been extended 1
downward from 2000 to 1750 psia.
These modifications have been approved by the USNRC." The use of this l
correlation in conjunction with increased syste flow for the Cycle 2 analysis indicates that the eargin to D::B is greater for Cycle 2 than had been predicted for first core operation, as shewn in Tabic 6-1.
6.2 DNBR ANALYSIS
~
In addition ~to the items discussed abote, the DNBR analysis for Cycle 2 l
operation considered the caximum design conditions, as built fuel assembly
. geometry, and hot operating: conditions.
This resulted in a minimum DNSR of 1.90 at Lil2%, power for undensific d fuel.
The DNBR calculations for undensified l
fuel are based on a 144 in, active icngth.
The effect'of,fuci densification on minicua DNBR is primarily a result of.the reduction'in active fuc1 length,.which increases the' average heat.
flux.' - For this. cvaluation, the active fuel ~1ength of batch 4 fuel (140.49" 4
l densified) was used, resulting in a reduction'in calculated HDNBR of 1.93%,,
L(froml1.956~ to 1.919 atL desinn overpever).
x
- The potential.cf fect of fuch rod bow on MD::BR is considered by l
(incorporating suitabic margins into DNB limited RpS setpoints.
Tac maximum 1
,6-21
-' Babcock & Wilcox m
s-
~- -
T rod'hrr,i manni.tude wan calculated from the equestion t'b = 11. 2 + 0.069 /nu,
where a is the rod bow magnitudet in mlle and h!! is the burnup in If.4D/MTU.
b The D'iBR penalty was determined from the equation P = 34 a /138 where P is b
the percent liNBR penalty.
The resulting DNBR penalty, based on a burnup equivalent to throc cycles of operation, was 5.97..
6.3 PPISSURE-TEMPERATURE LIMIT EVALUATION Pressure-temperature l'imits for cycle 1 operation were based on one vent valve failing open which would reduce the effective coolant flow for heat transfer by 4.6%.
An NRC Staff evaluation ( I) of operating data (fromB&Wplants(12)relievedB6Wfromhavingtoincludeaventvalve flou penalty in safety analyses.
Therefore, the evaluation of Pressure-Temperature limits for cycle 2 operation did not consider an open vent
. valve. As discussed in 6.2, these limits include the effects of the reduction in active long'th resulting from densification and the maximum three-cycle rod bow DNBR penalty.
6.4 FLUX-TO-FLOU SETPOINT EVALUATION The method of evaluating the flux-to-flow setpoint for cycle 2. operation deviated from the cycic 1 setpoint analyses in that the~ power versus time
-- uns input directly to the transient DNBR calculation.
During a RC pump coas tdown, the time to reactor trip and hence the power transient is dif ferent
' fo' cach setroint. Therefore, the technique used is to select a setpoint and determine the power transient for that setpoint.
This is accomplished by first deternining the f]cw fraction associated with the flux-to-flow setpoint and the indicated power icyc1 (102% of design).
The flow fraction is then related 'to. the appropriate pump coastdown curve. (two pump for ~ ANO-1) to obtain the time at which the trip signal occurs.
The delay time from the-
. initiation of the trip cignal to the start of control rod motion (1.4 sec.)
- is then added'to deternbic the tine;when the power starts to decrease as a 8
+.
6-3_
5abcock a Wilcox s
M.
m
result of the reactor scram.
"Ih l t, power transient in conjunction ulth the pump coastdown (flow trant;ient) are input to the RADAlt(I3} code and a IJNBR is calculatej based on an assumed initial power luvel of 108% full power.
This procedure is completed for a second, and if necessary, a third setpoint.
The smallest DNBR occurring during each transient analysis is then plotted versus the correspondi,ng setpoint.
From this plot the minimum setpoint necessary to maintain a minimun DNER greater than 1.30 throughout the transient is determined.
In addition to the ecthod just described, the ANO-1 flux-to-flow setpoint evaluation also included the increased RC system flew (Section 6.1.2), the fuel densification penalty, and the maximum three-cycle fuel rod bow penalty (Section 6.2).
The results of this analysis indicate that a flux-to-flow setpoint of 1.11 would provide adequate protection for cycle 2 operation.
This value, however, represents the thercal-hydraulic lini't.
The actual setpoint specified is lower to conservatively account for instrunent crror and flow noise.
6-4 Dabcock & \\VilCOX s
TAHl.l: 6-1.
Cyc]c ! and 2 !'axinon De nf on Conef f eltinr.
9 Cycle 1 Cycle 2 Design poier levci,If.It 2568 2568-System pressure, psia' 2200 2200
' Reactor coolant flow, %. design flow 100.0 106.5 Vessel inlet coolant'tecperature, 100% power, F 554.0
~555.6 Vessel outlet coolant temperature, 100% power, F 603.8 602.4 Ref. design axial flux shape 1.5 cosine 1.5 cosine Hot channel factors Enthalpy rise 1.011 1.011
!! cat flux 1.014 1.014 Flow area 0 98 0.98 Active fuel length, in.
Refer to Table 4-4 Avg. heat flux -100% power, Btu /h-ftz(nj' 175205 175205
}!ax. heat flux, 100% power, Btu /h-ft2(b) 467797 467797 OlF correlation U-3 B&W-2
}!inicun DNBR No densification penaltics 1.55 (ll4% power) 1.956 (112% ' power)
- Fuel densification penalty 1.46 (114% power) 1.919 (112% power)
.(a) Based on densified length of batch 3' fuel and hot fuel rod' diameter (b) Based on averano heat-flux with reference peaking.
6-5
' Babcock & Wilcox
~
~~
3 _ :-.
n I h otectJon r.nd Tic re 8-1.
A':0 ;t: nit. 1,_ Cycle 2 - Ce Safety 1.imitu 2000 s
2400 2
2200 o
w y
5 ACCEPTABLE m.
S OPERAil0N a.
w U
~
2000 m
o Eg UNACCEPTABLE OPERATION 1800 F
4-1000 040 000 550-500 000 020 Reactor Dutict Temperature, F'
t g 27 Babcock t.Wilcos t
u:
_J
Ti ;ure ' i'-2.
At:0 thill 1, Cych 2 - Cort: l'rotection zand t
.Sti f e ty I.f iai tn TilERil.At P0f:Lk LEVEL. 5 UllACCEPTABLE OPE!!AT lCil 120 1
Kn/ft lilliI ACCEPTABLE 4 PU!!P OPERATl0ft 100 ACCEPTABLE 3L4 80 PUMP OPERATIO!!
i 60
~
. _.. ACCEPT ABLE
~ ~ '"~~~~'~~~~~~~' -'
P d
~ ~ ~ ~ ~ ' - - - ' ' '
OPERAIION V
40 20 o
'*140iiE fil.!fTl!!s TilAlf CALCULATED DHSR On KVI/fT LilitTS
-I I
I I
I J
-20 0
.20 40 00
-00
-40 Reactor Power lanalance CURVE REACTOR COOL A!!T FL0li (Rpm) j -- - - - -. -
374,880 3*'
I' 184'.441
Babcock s. Wilcox
~i 8, 3
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C.
Fi r.ut e ' 8-3.. /,::0 tia i t.. )., Cycle 2 - Core ' Protect Jh: and-SnicLy l.isalt:;
k 2000 k
'2400 a
3 2200 3
0 i
t.
2 2
O
/!
3 2000 O
as 3
3 l
1000
/
/
p l
1000 560 500 000 E20 040 600 Reactor Outlet Tenperature, 'T l
~' ;
i j
CURYC -.
GPl4 POWER PU!!PS OPERAll!?G (TYi'E Of Ll!.!!T) 1 374,8B0 (1005)$
"1125 FOUR PU?'?S (P!:i.R Llyii) l 200,035. (74. E).
00.75 THREE PUP.PS (Oh5R LIMIT) 2-
-3 184,441 (49.25)
.59.05 OI C PU!!P 11! EACH LOOP (0UALITY LIMii) j.
- 10G.50 0F DESIGh' FL0il 3
' I
. I l 4 '
Dabcock r.Wilcox 3
p 1.
y,
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Figure t 4.
c;9 lin I L i. Cic 1 'I ~ h ' ' 't 8 ' 'i :.p t <. ia 1::ixJ nun A1 t o ali) e.':e ti </ '-
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4
- 2500' 9
P =.2355 psiE-T = GIO*F w
2300 v,
O ACCEPTABLE OPERATION 5
E~
2100 Z
2 P = 11.75 Tout ~
D 8
5103 psir.
/
0
/
M 1900 u*
UNACCEPTABLE OPERATION P'= 1800 psiE 1700
~
.1500' 560 580 000 020 040 660 Reactor Outlet Tcaperature, F
Babcocit t. Wilco):
G-5 em>
t r.,x1 nina /.1 fr,..it,1.: :.ci;.oI.:,
^
Til[H!al P01:EP, LEVEL. S
-- 120'
' UllACCEPT Ai;LE OPERATI0ll.
(10G.5) 4 g
I g.
+
100 D
ACCfPTABL[ 4 y
Pul4P OPERAll0N'
'8f?
(79.5)
-,~r.0_,
P ACCEPTABLE 34 4 P U ? te OPCRATION
- 60 (52.3)
A C C C P T A B L E 2.3
& 4 P Ul!P OPERATION
.:-- 4 0
-- 20
..o c,>
o V
II II ll ll
? '
m m.
4 I,
I" I " _,,
I
=
j
-80.
-40
-20
~0 20 40 00 Power imoalance.
G. Scock r. Wilcox 8-. 6 b
~
4.2 REACTOR COOLANT SYSTEM SURVEILLANCE Applicability Applies: to the surveillance of the reactor coolant system pressure boundary.
Objective
.To assure the continued integrity of the reactor coolant system pressure boundary.
Specification
.4.2.1 Prior to initial unit operation, an ultrasonic test survey shall be made of reactor coolant system pressure boundary welds as required to establish preoperational integrity and baseline data for future inspections.
4.2.2 Post operational inspections of components shall be made in accordance with the methods and intervals indicated in IS-242 and IS-261 of Section XI of the ASME Boiler and Pressure Vessel Code, 1971, including 1972 Summer Addenda, except as follows:
S-261 Item Component Exception 1.4 Primary Nozzle to 1 RC inlet nozzle to be inspected at Vessel Welds approximately 3 1/3 years of the inspection interval.. All four RC inlet nozzles to be inspected at or near the end of the inspection inter-val. At approximately 6 2/3 years of the inspection interval, both RC outlet nozzles will be inspected.
At approximately 31/3 years of the inspection interval, one core flood nozzle will be inspected and one core flood nozzle will be inspected at or near the end of the inspection interval.
3.3 Safe Ends on Not Applicable Heat Exchanger 4.1 Vessel Safe Not Applicable End Welds 4.2 Valve Pressure Not Applicable Retaining Bolting j
Larger than 2" l
l 4.9 Integrally Welded Not Applicable Supports 6.1 Valve Body Welds Not Applicable
- 6. 3 ~
Valve to Safe Not Applicable End Welds 76
~
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IS-261 Item Component Exception 6.4 Bolting 2 Not Applicable 6.6 Integrally 'Jcided Not Applica$le Valve Supports 4.2 3 The structural integrity of the reactor coolant systen tcur.dary shall be m.aintained at the level required by the original accep-tance standards throughout the life of the station.
Any evidence, as a result of the tests outlined in Table IS-261 of Section XI of the code, that defects have developed or grown, shall be investigated.
4.2.4 To assure the structural integrity of the reactor interr.als thro.sh-out the life of the unit, the two sets of main interr.als tol'.s (connecting the core barrel to the core support shield and to the lower grid cylinler) chall remain in place and under tencior..
This vill be verified by visual inspection to determine that the velded bolt locking caps remain in place. All locking caps vill be inspect-ed af ter hot funct?onal testing and wheneve.- the internals are re=oved from the vessel during a refueling or =2intenance shatdown.
The core barrel to core support shield caps vill be inspected each refueling shutdown.
P 4.2 5 Sufficient records of each idspection shall be kept to allow comparison and evaluation of future inspections.
4.2.6 complete surface and volumetric examination of the reactor coolant pump flywheels vill be conducted coincident with refueling or maintenance shutdowns such that within a 10 year period af ter start-up all four reactor coolant pump' flywheels vill be examined.
h,2,7 Reactor vessel specimens shall be removed and examined, to deter-mine changes in material properties, at specimen exposure (E>lMey) equivalent to 3, 9.5,16 and 22.5 Ef fective Full Power Years- (ETPY) of operation.
This withdraval schedule may be modified to coincide with those refueling outages or plant shutdowns, when the reactor head is removed, most closely noproaching the withdrawal schedule.
Results of these examinations en all be used to update Technical Specification 3.1.2.
Specimens not subj ected to des tructive testing af ter the first 0.93 EFPY of Cycle 1 may be removed and stored during the remainder of Cycle 1, but shall be re-installed prior to Cycle 2.
Bases The surveillance progran has been developed to comply with Section XI of the ASKE Boiler and Pressure Vessel Code Inservice Inspection of Nuclear Reactor Coolant-Sys tems, 1971, including 1972 Summer Addenda cdition.
I-The number 'of reactor vessel specimens and the frequencies for removing and testing these specimens are provided to nasure compliance with the requirements of Appendix H to 10 CFR Part 50.
i 77
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