ML19210B119
| ML19210B119 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/22/1977 |
| From: | Goller K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19210B115 | List: |
| References | |
| NUDOCS 7911040055 | |
| Download: ML19210B119 (44) | |
Text
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UNITED $TATES
- 4 NUO EAR REGULATORY COMMisslON
..j \\.IJ ' j WASHINGTON. D. C. 20565 e
c METR 0POLITAN E01 SON COMPANY JERSEY CENTRAL POWER lND_ LIGHT CO'tPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATIO*l. UNIT NO.1 AMENDMFNT TO FACillTY OPERATING LICENSE Amendment No. 29 License No. DPR-50 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment by Metropolitan Edison Company, Jersey Central Power & Light Company, and Pennsylvania Electric Company (the licensees) dated October 29, 1976, as supplemented December 29, 1976, and January 20, 1977; January 26, 1977, as supplemented March 31, 1977; and February 23, 1977, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confamity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) 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; 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; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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2 Accordingly, the license is amended by changes to the Technical 2.
Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-50 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 29. are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
This license amendment is effective as of the date of its 3.
issuance.
FOR THE NUCLEAR REGULATORY COPHISSION lyf.
sea & 0 Karl R. Go11er. Assistant Director for Operating Reactors Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: April 22, 1977 j'
isvJ 4JU
v i
ATTACHMENT TO LICENSE AMENDMENT NO. 29 FACILITY OPERATING LICENSE NO. DPR-50 Docket No. 50-289
</
The changed areas on the revised pages are shown by marginal lines.
Pages 4-27 and 4-28 are unchanged and are included for convenience only.
Remove Pages Insert Pages y
v vi vi vii vii 1-5 1-5 2-3 2-3 Figures 2.1 2.1-3 Figures 2.1 2.1-3 Figure 2.3-2 Figure 2.3-2 3 3-5 3 3-5 Figures 3.1-1 & 3.1-2 Figures 3.1-1 & 3.1-2 3-16 3-16 3-34 & 3-34a 3-34 & 3-34a 3-35 & 3-35a 3-35 & 3-35a 3-36 3-36 & 3-36a Figures 3.5-2A - 3.5-2J Figures 3.5-2A - 3.5-2M 4 4-13 4 4-13a 4-27 4-27 & 4-27a 4-28 4-28 vd
,,)
7' u
Y e.
LIST OF TABLES n
Table Title f.agg, Reactor Protection System Trip Setting Limits 2-9 2 3-1 3 5-1 Instruments operatins Conditions 3-29 k.1-1 Instrument Surveillance Requirements 4-3 k.1-2 Minimum Equipment Test Frequency
- 4-8 k.1-3 Minimum Sampling Frequency k-9 k.2-1 Instrument Surveillance Program k-Ik 4.2-2 Surveillance Capsule Insertion & Withdrawal Schedule at TMI-2 4-27a k.k-1 Selected Tendons and Corresponding Inspection k-35a Periods h.h-2 Tendons Selected for Tendon Physical Condition Test k-36 k.k-3 Ring Girder Surveillance k-36g k.15-1 Radioactive Liquid Waste Sa=pling and Analysis k-59 k.15-2 Radioactive Gaseous Waste Sampling and Analysis k-63 6.12-1 Protection Factors for Respirators 6-23 e
k
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jgg iUC AmendmentNo./. 29
LIST OF FIGURES Figure Title 2.1-1 Core Protection Safety Limit 2.1-2 Core Protection Safety Limits y
2.1-3 Core Protection Safety Basis 2 3-1 Protection System Mar 4== A11ovable Set Points 2 3-2 Protection System Mav4== A11ovable Set Points 3.1-1 Reactor Coolant System Hes, tup Limitations 3.1-2 Reactor Coolant System Cooldova Limitations 3.1-3 Limiting Pressure Vs. Temperature Curve for 100 STD cc/ Liter H2O 3 5-1 Incore Instrumentation Specification Axial Imbalance Indication 3 5-2 Incore Instrumentation Specification Radial Flux Tilt Indication 3 5-2A Rod Position Limitt. for h Ptarp Operation Applicable During the Period fr a 0 to 100 10 EFPD; Cycle 3 3.5-2B Rod Position Limits for h Pump Operation Applicable During the Period from 100 2 10 EFFD; to 2k62 10 EFFD; Cycle 3 3 5-2C Rod Position Limits for k P ap Operation Applicable During the Period after 2k6 1 10 EFPD; Cycle 3 3 5-2D Rod Position Limits for 2 and 3 Pu=p Operation Applicable During the Period frcm O to 100 10 EFFD; Cycle 3 3 5-2E Rod Position Limits for 2 and 3 Pump Operation Applicable During the Period frem 100 1 10 to 2462 10 EFFD; Cycle 3 3 5-2F Rod Position Limits for 2 and 3 Pump Operation Applicable During the Period After 2k6 210 EFPD; Cycle 3 3 5-2G Operational Power Imbalance Envelope Applicable to Operatien frem 0 to 1001 10 IFFD; Cycle'3 n
en
.) V U
vi Amendment No. g g, 2 9
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Title 3 5-2H Operational Power Imbalance Envelope Applicable to Operation from 1001 10 to 2k61 10 2TFD; Cycle 3 3 5-2I Operational Power Imbalance Envelope Applicable to Operation after 2h61 10 EFFD; Cycle 3 3.5-2J LOCA Limited Maximum A11ovable Linear Heat Rate 3 5-2K APSR Position Limits for Operation from 0 to 1001 10 EFFD; Oycle 3 3 5-2L APSR Position Limits for Operation from 1001 10 to 2k61 10EFFD; Cycle 3 3.5-2M APSR Position Limits for Operation after 2k61 10 EFFD; Cycle 3 3 5-3 Incore Instrumentation Specification h.2-1 Equipment and Piping Requiring Inservice Inspection in Accordance with Section XI of the ASME Code k.h-1 Ring Girder Surveillance k.k-2 Ring Girder Surveillance Crack Pattern Chart 4.h-3 Ring Girder Surveillance Crack Pattern Chart h.h k Ring Girder Surveillance Crack Pattern Chart.
h.h-5 Ring Girder Surveillance Crack Pattern Chart 6-1 Organization Chart e
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Isvu duri
[, [, 20 Amendment No.
sti
s c
1.6 POWER DISTRIBUTION 1.6.1 QUADRANT POWER TILT Quadrant power tilt is defined by the following equation and is expressed in percent.
100 Power in any core quadrant
-y Average Power of all quadrants
~
The quadrant tilt limits are stated in Specification 3.5 2.k.
1.6.2 REACTOR POWER IMBALANCE the Reactor power imbalance is the power in the top half of the core minus power in the bottom half of the core expressed as a percentage of rated power. Imbalance is monitored continuously by the RPS using input from the power range channels. Imbalance limits are defined in Specification 2.1 and imbalance setpoints are defined in Specification 2.3 1.T CONTAINMENT INTEGRITY Containment integrity exists when the foIloving conditions are satisfied:
The equipment hatch is closed and sealed and both doors of the a.
personnel hatch and emergency hatch are closed and sealed except as in "b" below.
b.
At least one docr on each of the personnel hatch and emergency hatch is closed and sealed during refueling or personnel passage through these hatches.
All non-automatic containment isolation valves and blind flanges are c.
closed as required by the " Containment Integrity Check List" attached to the operating procedure " Containment Integrity and Access Limits."
d.
All automatic containment isolation valves are operable or locked closed.
The containment leakage deter ined at the last testing interval e.
satisfies Specification k.k.1.
<, :n "
Amendment No. 29 1-5 6
The curve of Figure 2.1-1 is the most restrictive of all possible 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 mmvimum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 22 percent, (3) whichever condition is more restrictive.
The maximum thermal power for three pump operation is 86.7 percent due to a power level trip produced by the flux-flow ratio (714.7 percent flov x 1.08 = 80 7 percent pover) plus the==v4m = calibration and instrumentation The maximum thermal power for other reactor coolant pump conditions error.
is produced in a similar manner.
Using a local quality limit of 22 percent at the point of minimum DNBR as a basis for curve 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 DNER.
The DNBR as calculated by the BW-2 correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.
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. Curve 1 is more restrictive than any other reactor coolant pump situation because any pressure /te=perature point above and to the left of this curve vill be above and to the left of the other curves.
REFERDICES (1)
FSAR, Section 3.2.3.1.1 (2)
FSAR, Section 3.2.3 1.1.c (3)
PSAR, Section 3.2 3.1.1.k p
R Q
j UU 2-3 Amendment No.
29
o 2600 2400
_=
ACCEPTABLE E
OPEPATION f
2200
/
A
%::5 2000 UNACCEPTABLE OPERATION 1600 7
s 1600 560 560 600 620' 640 660 Reactor Outist Temperature, F Dy TMI-1, UNIT 1, CYCLE 3 b
i CORE PROTECTION SAFETY LIMIT Figure 2.1 1 j]v3 2b Amend:nene No.
29
Tneraal Power Levsi, l
- 120
( 33,112)
(+"i,il2)
ACCEPTABLE Kv/ft Limit
- - 100 4 PUMP Kv/ft Limit OPERATION
(-56,90)
( 33,86.7) 90 (86.7)
(+35.86.7).
h 80 ACCEPTABLE 3 & 4 PUMP OPERAfl0N
( 56,64.7)
( 33,59.1) 60 (59.1)
(+35,59.1)
ACCEPTABLE 5a
(.4,,4,)
2.3 i 4 Pu e c.u,4.)
OPERATION 40 30 i
20 i
10 I
I I
i i
80 40
-20 0
20 40 80 Reactor Power Isaalance, 5 CURVE REACTOR COOLANT FLOW (lo/nt) 3 6
TIl 1, UNIT 1 CYCLE 3 CORE PROTECTION SAFETY LIMITS Figure 2.1 2 4 r a
?CB i sou Amendment No.g 2 9
2400 N
?
2200 E.
{
p 2
5 5
2000
/
5
[
3 5
i 1800 av 1600 560 580 800 620 S40 660 Reactor Outlet Temperature, 'F REACTOR COOLANT FLOW g
CURVE (LBS/HR)
POWER PUMPS OPERATING (TYPE OF LIMIT) 1 139.8 x 106 (1005)*
1125 Foer Pumps (DNBR Limit) 2 104.5 x 106 (74.75) 86.75 Tnree Pumps (DNBR Limit) 3 68.8 x 106 (49.25) 59.15 One Pump in Eacn Loop (Quality Limit)
- l06.5% of Cycle 1 Design Flow Till 1, UNIT l. CYCLE 3 CORE PROTECTION SAFETY BASES Figure 2.1 3 20 I
U AmendmentNo.[,
t' Power Level, 5
--120
- 110 (108)
(+21,108)
(-20,10s)
$+
- 100 i
g r
ACCEPTABLE 4 PUMP 30 D
OPERATION I'I
(-4s,s0)
(.,30}
70 ACCEPTABLE 3 8 4 PUMP OPERATION so I
(+48,52.7)
(-46,52.7)
~~
50 l
I l
40 ACCEPTABLE 2,3 & 4 PUMP 30
(-4s,25.1)
OPERATION
(+43,25.i) 20 l
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-50
-40
-30
-20
-10 0
10 20 30 40 50 Power lebalance, 5 THI-1, UNIT 1 CYCLE 3 PROTECTION SYSTEM MA11 MUM ALLOWABLE SET P0lRT3 Figure 2.3-2 OIO
\\g-3 V
JV Amendment No.
2S
r.
3.1.2 PRESSURIZATION HEATUP AND COOLDOW LIMITATIONS Applicability Applies to pressurization, heatup and cooldown of the reactor coolant system.
Objective To assure that temperature and pressure changes in the reactor coolant system do not cause cyclic loads in excess of design for reactor coolant system components.
Specification 3 1.2.1 For operations until five effective full power years, the reactor coolant pressure and the system heatup and cooldown rates (vith the exception of the pressurizer) shall be limited in accordance with Figure 3.1-1 and Figure 3.1-2 and are as follows:
Heatun/Cooldovn A11ovable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-1.
Heatup and cooldovn rates shall not exceed those shovn on Figure 3.1-1.
t Inservice Leak and Hydrostatic Testing A11ovable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-2.
Heatup and Cooldown rates shall not exceed those shown on Figure 3.1-2.
3.1.2.2 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is belov 100cy, 3.1.2.3 The pressurizer heatup and cooldown rates shall not exceed 17'9F in any one hour.
The spray shall not be used if the temperature dif-ference between the pressurizer and the spray fluid is greater than h300F.
3.1.2.h Prior to exceeding five effective full power years of operatien, Figures 3.1-1 and 3.1-2 shall be updated for the next service period in accordance with 10 CFR 50, Appendix G, Section V.E.
The highest predicted adjusted reference temperature of all the beltline naterials shall be used to determine the adjusted reference te=perature at the end of the, service period. The basis for this prediction shall be submitted for URC staff review in accordance with Specificatien 3.1.2 5 Amendment flo. % % 2 9
) [,i;3 2 h
f 6
3.1.2.5 The updated proposed technical specifications referred to in 3.1.2.h shall be submitted for NRC review at least 90 days prior to the end of the service period. Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accor-dance with 10 CFR 50, Appendix G, Section v.C.
Bases
,j
. All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.(1)
These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The nu=ber of thermal and loading cycles used for design purposes are shown in Table k-8 of the FSAR.
The w h unit heatup and cooldo
} rate of 1000 F in any one hour satisfies stress limits for cyclic Operation.
The 200 psig pressure limit for the secondary side of the steam generator at a temperatipe less than 1000 F satisfies stress levels for temperatures belov the DTI.(33 The heatup and cooldown rate limits in this specification are not intended to limit instantaneous rates of temperature change, but are intended to limit temperature changes such that there exists no one hour interval in which a temperature change greater than the limit takes place.
The unirradiated reference nli ductility temperature (RT NDT) for the surveil-
-lance region materials vere deter =ined in accordance with 10 CFR 50, Appen-dixes G and H.
For other beltline region materials and other reactor coolant pressure boundary materials, the unirradiated impact properties vere estimated using the methods described in BAW-100k6.
As a result of fast neutron irradiation in.the beltline region of the core, there vill be an increase in the RT NDT vith accu =ulated nuclear operations.
The adjusted reference te=peratures have been calculated by adding the pre-dicted radiation-induced RT NDT and the unirradiated RT NDT for each of the reactor coolant beltline materials.
The predicted RT NDT vas calculated using the respective neutron fluence after five effective full power years of operation. To reflect the uncertainty in the copper and phosphorus concentration of the beltline velds, the upper limit of Figure 1 of Regulatory Guide 199, Revision 1, was utilized for added conservatism.
The analysis of the reactor vessel material contained in the first surveillance capsule removed from Three Mile Island Nuclear Station Unit 1 confirmed that the current techniques used for predicting the change in impact properties due to irradiation are conservative.
Analysis of the activation detectors contained in the first surveillance capsule indicates that the average fast flux during Cycle 1 was 1.k5 x 1010 2
D/cm -sec maximum at the pressure vessel vall.
Extrapolation of the Cycle 1 flux based on predicted fuel reload and burn-up conditions indicates that the maximum average fast neutron (E>l Mev) flux during six full power years of 2
10 n/en -see at the reactor vessel vall and 9 33 x operation vill be 1.68 x 10 109 "/tm2-sec at the h T location. The fast neutron exposure during five effective full power years of operation, therefore, is 1 5 x 1018 n/cm2 at the k. T location and 3 7 x 10 7 nen2 at the 3/k T location.
1 O
Amendment No. gg 2 9 3-h o^a 292
-}
v Based on the predicted RT NDT after five effective full power years of operation, the pressure-te=perature limits of Figure 3 1-1 and 3.1-2 have been established in accordance with the requirements of 10 CFR 50, Appendix G.
The methods and~ criteria e= ployed to establish the operating pressure and te=pera-ture limits are as described in BAW-100h6. The protection against nonductile failure is assu=ed by maintaining the coolant pressure below the upper li.mits of these pressure tempersture li=it curves.
The pressure limit lines on Figures 31-1 and 31-2 have been established considering the following:
a.
A 25 psi error in measured pressure.
b.
A 120 F error in measured temperature, c.
System pressure is measured in either loop.
d.
M n4mtm differential pressure between the point of system pressure measurement and reactor vessel inlet for all operating pu=p co=binations.
The spray temperature difference restriction, based on a stress analysis of spray line no::le is i= posed to =aintain the therme.1 stresses at the pressurizer spray line no::le below the design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell.
REFERENCES i
(1) FSAR, Section h.1.2.h (2) ASME Boiler and Pressure Code,Section III, N-h15 (3) FSAR, Section h.3.10 5 (h) BAW-lh39, Analysis of Capsule 'IMI-IE From Metropolitan Edison Co=pany, Three Mile Island Nuclear Station - Unit #1, Reactor Vessel Materials Surveillance Program.
M$1uuub me; Amendment No-29 3-5 1 2 a m, -,
- e
f.
c v
Assumed RT F
Beltline 1/h T 145 Beltline 3/h T 82
'i Closure Head Region 60 Outlet Nozzle 60 2300 Point Temp Press.
E A
75 105 2200 B
125 h10 C
175 h85 D
275 h85 2000 E
320 2275 RC Panp Combinations A11ovable 1800 Above 1950 F All 0
Below 195 F 1-A,1-B,0-A,1-B; 1-A,0-B 1600 g
E 1)
When Decay Heat Removal System (DH)
J
- 1h00, is operating without any RC pumps g
operr. ting, indicated DH return temp.
25 to the Reactor Vessel shall be used.
1200 5
2)
Heat-up and Cooldown rates shall not exceed 100 F in any one hour 1000 m
U u
600 C
D a:
h D
h h00 B
200 A
g 8
0 50 log 150 200 250 200 350 Indicated Reactor Coolant Temperature, F Reactor Coolant System Heat-up/Cooldown Limitat:one (Applicaele to 5 EFPY)
Amendmentho. 29 Figure 3 1-1 1588 294
o.
y 1
Assumed RTg,F Beltline Region 1/h T 1h5 Beltline Region 3/h T 82 Closure Head Region 60 Outlet Nozzle 60 4
Point Temp Press.
E 7
2h00--
A 75 175 B
115 h05 C
160 h85 2200.-
D 260 h85 E
287 2500 2000.-
RC Pump Combinations A11ovsble 1800 Above 195 F All Below 195 F 1-A,1-B*,0-A,1-B ;
1-A,0-B 1600.."
N 1)
When Decay Heat System (DH) 1h00 is operating without any RC pumps g
operating, indicated DH Return Te=p g
to the Reactor Vessel shall be used.
m N
2)
Heat-up and Cooldown Rates shall nc t 0
exceed 50 F in any one hour.
Ed h
1000 8o m
600 8
C D
o h00 B
o 5
200 --
A 50 100 150 20D 250 30b 0
INDICAT D REACTOR COOLANT TDIPERATURE, F Reactor Coolant System Inservice Leak & Hydrostatic Test Limitations (Applicable to 5 EFPY)
Amendmentho. 29 Figure 3.1-2 15S3 29,.3
y
.2 3.1.7 MODERATOR TEMPERATURE COEFFICIINT OF RZA m y m Applicability Applies to==H==
positive moderator temperature coefficient of reactivity at full power conditions.
,j Objective To assure that the moderator temperature coefficient stsys within the limits calculated for safe operation of the reactor.
Specification 3.1.7 1 The moderator te=perature coefficient shan not be positive at power
- 1evels above 95% of rated power.
3.1.7.2 The moderator temperature coefficient shall be 1 + 0.5x10-4 Ak/k/F at power levels 1 95% of rated power.
Bases A non-positive moderator coefficient at power levels above 95% of rated power is specified such that the==H=m clad temperatures vill not exceed the Final Acceptance Criteria based on LOCA analyses. Belov 95% of rated power the Final Acceptance Criteria vin got be exceeded with a positive moderator temperature coefficient of +0 5 x 10- AK/K/F. All other accident analyses as reported in the FSAR have been performed for a range of moderator temperature coefficients including +0.5 x 10-k AK/K/F' The experimental value of the moderator coefficient vill be corrected to obtain the hot full power moderator coefficient. The correction factor vill be verified during startup testing on earlier B&W reactors.
The Final Acceptance Criteria states that post-LOCA clad temperature vill not
~
exceed 2200 F.
REFERENCES (1)
FSAR, Section ik (2)
FSAR, Section 3 0
(
lJuO 410 I fI O "O /
>16 Amendment No. A( 2 9
i 4
y f.
If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification k.7.1.2.,
operation may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allovable group average position limits of Specification h.7.1.2.
g.
If the inoperable rod in Paragraph "e' above is in groups 5, 6, 7 or 8, the other rods in the group shall be trimmed to the same position. Normal operation of 100 percent of the thermal power allevable for the reactor coolant pu=p ccmbination may then continue provided that the rod that was declared inoperable is maintained within allovable group average position limits in 3.5.2 5 s
3 5.2.3 The worth of single inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the control Rod Position Limits defined in Specification 3.5 2 5 3.5.2.h Quadrant tilt:
a.
Except for physics tests the quadrant tilt shall not exceed
+2.66% as determined using the full incore detector system.
b.
When the full incore detector system is not available and except for physics tests quadrant tilt shall not exceed +1.k7% as determined using the minimum incere detector system.
c.
When neither incore detector system above is available and except for physics tests quadrant tilt shall not exceed +0.81%
as determined using the power range channels displayed on the console for each quadrant (out of core detector system).
d.
Except for physics tests if quadrant tilt exceeds the tilt limit power shall be reduced i= mediately to below the power level cutoff (see Figures 3 5-2A, 3.5-23 and 3 5-2C).
Moreover, the power level cutoff value shall be reduced 2 percent for each 1 percent tilt in excess of the tilt limit. For less than four pump operation, ther=al power shall be reduced 2 percent of '.he thermal pcVer allevable for the reactor coolant pump combination for each 1 percent tilt in excess of the tilt limit.
e.
Within a period of h hours, the quadrant power tilt shall be reduced to less than the tilt limit except for physics tests, l
or the following adjustments in setpoints and limits shall be made:
1.
The protection system reactor pover/i= balance envelope trip setpoints shall be reduced 2 percent in power for
.each 1 percent tilt.
Q$Y h
19 b3 s
3-3h e-3 AmendmentNo./[, 29 IU08 297
Y 2.
The control rod group vithdraval limits (Figures 3.5-2A, 3 5-2B, 3 5-2C, 3.5-2D, 3.5-2E, 3 5-2F, 3 5-2K, 3.5-2L, and 3 5-2M) shall be reduced 2 percent in power for each I percent tilt in excess of the tilt limit.
3.
The operational imbalance limits (Figure 3.5-2G, 3 5-2H and 3 5-2I) shall be reduced 2 percent in power for each 1 I
percent tilt in excess of the tilt limit.
Except for physics or diagnostic testing, if quadrant tilt is f.
in excess of +25 28% determined using the full incere detector system (FIT), or +2h.09% determined using the =ini=um incore detector system (MIT) if the FIT is not available, or +21.39%
determined using tha out of core detector system (OCT) when neither the FIT nor MIT are available, the reactor vill be Placed in the hot shutdown condition. Diagnostic testing during power operation with a quadrant tilt is permitted provided that the thermal power allowable is restricted as stated in 3.5 2.h.d above.
Quadrant tilt shall be monitored on a minimum frequency of once g.
every tvo hours during power operation above 15 percent of rated power.
D"*D D'T WW c
c s
IsuV
(/O 3-3ha Amendment No. 29 1
3.5.2.5 Control k ositionst Operating rod group overlap shall not exceed 25 percent! 5 a.
percent, between two sequentisl groups except for physics tests.
b.
Position limits are specified for regulating and axial power shaping control rods. Except for physics tests or exercising cor. trol rods, the regulating control rod insertion /vithdrawal limits are specified on Figures 3.5-2A, 3 5-23, and 3.5-2C for four pump operation and Figures 3 5-2D, 3.5-2I, and 3 5-2F for three or two pump operation. Also excepting physics tests or exercining control rods, the axial power shaping control rod insertion /vithdrawal limits are specified on Figures 3 5-2K, 3.5-2L, and 3.5-2M.
If any of these control rod position limits are exceeded, corrective measures shall be taken im=ediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours.
c.
Except for physics tests, power shall not be increased above the power level cutoff (See Figures 3.5-2A, 3 5-23 and 3 5-2C) unless the xenon reactivity is within 10 percent of the equilibrium value for operation at rated power and asymptotically approaching stability.
d.
Core imbalance shall be monitored on a mini::n::n frequency of once every two hours during power operation above h0 percent of rated power. Except for physics tests, corrective measures (reduction of imbalance by APSR movements and/or reduction in reactor.
pover) shall be taken to maintain operation within the envelope defined by Figures 3.5-2G, 3.5-2H and 3.5-2I.
If the i= balance is not within the envelope defined by Figures 3.5-2G, 3.5-2H and 3.5-2I corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within four hours, reactor power shall be reduced until i= balance limits are met.
e.
Safety red limits are given in 3.1 3.5 3.5.2.6 The control red drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.
3 5.2.7 A power map shall be taken at. intervals not to exceed 30 effective full power days using the incere instrumentation detection system to verify the power distribution is within the limits shown in i
Figure 3.5-2J.
Ii Bases The power-imbalance envelope defined in Figures 3 5-2G, 3 5-2H, and 3.5-2I is based on LOCA analyses which have defined the maxi =um linear heat rate (see Figure 3 5-2J) such that the maxi =um clad temperature vill not exceed the Final Acceptance Criteria (2000F). Operation out:ide of the power i= balance envelepe alone does not con:titute a situation that vould cause the Final Acceptance Criteria to be exceeded should a ICCA occur. The power imbalance envelope I
represents the beun(ary of operation li=ited by the Final Acceptance Criteria only if the control rods are at the withdrtval/ insertion limits as defined by Amendment No. )#,[, 2 9
}gg n
s
~
y" e
s Figures 3.5-2A, 3 5-2B, 3.5-2C, 3.5-2D, 3.5-2E, 3 5-2F, 3 5-2K, 2.5-2L, and 3 5-2M and if quadrant tilt is at the limit. Additional conservatism is introducted by application of:
a.
Nuclear uncertainty factors
,j b.
Thermal calibration uncertainty c.
Puel densification effects d.
Hot rod manufacturing tolerance factors.
e.
Postulated fuel rod bov effects The Rod index versus A11ovable Power curves of Figures 3.5-2A, 3.5-2B, 3 5-2c, 3.5-2D, 3.5-2E, 3 5-2F, 3.5-2K, 3.5-2L and 3 5-2M describe three regions. These three regions are:
1.
Permissible operating Region 2.
Restricted Regions 3.
Prohibited Region (Operation in this region is not allowed)
NOTE:
Inadvertent operation within the Restricted Region for a period of four hours is not considered a violation of a limiting condition for operation. The limiting criteria within the Restricted Region are potential ejected rod vorth and ECCS power peaking and since the probability of these accidents is very lov especially in a h hour time frame, inadvertant operation within the Restricted Region for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is allowed.
1588 00 3-35a AmendmentNo./,20
e The 2525 percent overlap between successive control rod groups is allowed since the worth of a rod is lover at the upper and lover part of the stroke. Control rods are arranged in groups or banks defined as follows: Group Function 1 Safety 9 2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Regulating (Xenon transient override) l 8 APSR (axial power shaping bank) Control rod groups are withdrawn in sequency beginning with group 1. Groups 5, 6 and 7 are overlapped 25 percent. The nor=al position at power is for groups 6 and 7 to be partially inserted. The rod position limits are based on the most limiting of the following three criteria: ECCS power peaking, shutdown cargin, and potential ejected rod vorth. As discussed above, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod vorth, consistent with the rod position limits, provider for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (1). The rod position limits also ensure that inserted rod groups vill not contain single rod worths greater than: 0.65% Ak/k at rated power. These values have been shown to be safe by the safety analysis (2) of the hypothetical rod ejection accident. A maximum single inserted control rod worth of 1.0% Ak/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth 1.0%Ak/k at beginning of life, hot, zero power would result in a lover transient peak thermal power and, therefore, less severe environmental consequences than 0.65% Ak/k ejected red worth at rated power. The plant computer will scan for tilt and-imbalance and vill satisfy the technical specification requirements. If the computer is out of service, then manual calculation for tilt above 15 percent power and imbalance above h0 percent power must be performed at least every two hours until the computer is returned te service. The quadrant power tilt limits set forth in Specification 3 5 2.h aave been established within the thermal analysis design base using an actual core tilt of +3.bl% which is equivalent to a +2.66% tilt =easured with the full incere instrumentation with measurement uncertainties included. During the physics testing program, the high flux trip setpoints are admini-stratively set as follows to assure an additional safety margin is provided: Test Power Trip Setpoint .0 <5% 15 50% LO 50% I O 7* 50 60% lsuu Ul 75 85% ( >75 105 5% Amendment No. y/, 2 9 3-36
i s REFEFENCES s (1) FSAR, SEction 3.2.2.1.2 (2) FSAR, Section 1k.2.2.2
- /
0 l 6 ROV L 3-36a i i I 1
aJ 100 OPERAil0N IN THIS REGION 15 NOT ALLOWE0 90 225.s. POWER LEVEL 17s.2.so \\so CUTOFF = 90% 10 iso.2.so REGION 23s.a. so SHUTOOWN RESTRICTED 70 MARGIN 14s. 2. 7o REGION LIMIT 2 e.s.ro soc.io j 60 ers.o.so PERNISSIBLE OPERATING
- , 40 REGION 30 29 -
- o. is 10 t
I I I I e i i t 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod index. 5 Withdrawn 0 25 50 75 100 i f f f Group 7 0 25 50 75 100 1 1 I t i Group 6 0 25 50 75 100 1 I l l t Group 5 200 POSifl0N LIMITS FOR 4 PUMP OPERATION FROM 0 TO 100 1 10 N Tul-1. CYCLE 3 Figure 3.5 2A Amendment No.N 2.9 A i
/.' 17s.o.lo2 225.s.lo2 P 100 OPERATION IN TNi$ REGION POWER LEVEL 15 NOT ALLOWED 90 ,17s.2.to 25.s.so iso.2.so RESTRICTED 80 REGION iss.2.70 23s.s.70 300.7o 70 g = $ 60 ShUTOOTH MARGIN s LIMIT a 50 ss.so PERMISSIBLE OPERATING REGION g 40 30 20 5.15 70.15 20 o.o i 1 1 I I I 1 I I I i 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod index 5 Withdrawn 0 25 50 75 100 1 I t f f Grcup 7 100 0 2,5 5,0 ip Grcup 6 f R00 POSITICN LIMITS FOR 4 PUMP i l i Group 5 OPERATION FROM 1001 10 EFP0 TO 246110 EFPO TMI-l. CYCLE 3 Amendment No.[g 2 9 r
- s u,o un
,,..s..., 100 OPERATION IN THIS REGION l$ NOT ALLOWED TO F = 0% ni.e.n 90 80 RESTRICTED 237.s.so REGION 70 g SHUTOOWN $ 60 MARGIN g LlulT c. % 50 I10.50 PERNISSIBLE OPERATING g 40 REGION E 30 20 ss..s 10 o.o I i i l 1 I I I I I l 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, 5 titndrawal 0 25 50 75 100 I f I f Group 7 0 2,5 50 75 10,0 Group'6 ,0 2,5 5,0 p ip pp)) )blM[1pf Group 5 R00 POSITION LlulTS FOR 4 PUhP OPERATION AFTER 246 1 10 EFPD TEl-1 CYCLE 3 Figure 3.5 2C l Amendment No. p 29 ,c., m lJUO . l) 3
s q 112.102 11s.102 150.102 250.102 RESTRICTED r dl0N 6 OPERATION IN THl3 100 FOR 3 PUMP REGION IS NOT ALLOWED OPERAil0N 2so.as soo.as ~ 5 b80 SHUT 00XN ILARGIN 129.7s 5 Likli 70 g@ z M 60 PERMISSIBLE OPERATING g REGION I50 46.50 5 ~ 540 2 30 Restricted Region for 2 & 3 pu:::p operation "g 20 0.15 10 00 I i i i i i i i i t 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod index, 5 titnarawal 0 25 50 75 100 t I t I f Group 7 0 25 50 75 100 t t t t f Group 6 "U R00 POSITION LlulTS FOR 2 AND 3 PUWF OPERATION FROM 0 TO 100 i 10 EFPO TNI-I CYCLE 3 Figure 3.5 20 ^o c n.' Amendment No.p 2 9 lr<J v0 su
its. lot tso. lot 100 RESTRICTED REGION FOR 3 PUNP OPERATION IN THl3 REGION OPERATich 90 IS NOT ALL0fE0 g 80 95 5 70 E u p 60 PERMISSIBLE OPERATING g SHUT 00fN MARGIN REGION u LIMIT = 50 ss.so s .2 ." 40 a" E
- 30 -
= O 20 s. f 53.l5 5 10 --
- 0. 0 0
I I f I f f I f i i 1 0 25 50 75 100 125 150 175 200 225 250 275 300 Roa inaen. 5 Ilthdrawal O 25 50 75 100 i f f f I Group 7 O 25 50 75 100 t 1 I f f Group 6 ]D
- ]D'9'l[&
Jaoj]D o ju 2..y 0 25 50 75 100 L u 12 m Group 5 R00 POSITION LlulTS FOR 2 AND 3 PUMP OPERAil0N FRON 1001 10 EFPD TO 2461 10 EFPD TMl-l. CYCLE 3 Figure 3 5-2E Amendment No. K 2 9 - . Jvd Vl
194 8of 214.Iof 100 OPERATION IN THIS REGION 15 NOT ALLOWED RESTRICTED REGION FOR 90 3 PUMP OPERATION g s 80 E j 70 SHUTDOWN MARGIN j PERMISSIBLE OPERATING Lluli .u REGION s,. 0 50 180.50 a. .o
- 40 30
.~ f 20 g ss.is a. 10
- 0. 0 I
I f f I I f f I t 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod index, 5 Witnaraun 0 25 50 75 100 1 f t i f Group 7 0 25 50 75 100 t 1 I t t Grcup 6 0 25 50 75 100 l 1 i t t I "U R00 POSITION LlWITS FOR 2 AND 3 PUMP OPERATION AFTER 246 1 10 EFP0 Tul-1, CYCLE 3 Figure 3 5 2F Arnendment No.g 2 9 su} ?n rcQ JUU
Power. 5 of 2535 Nst ~ RESTRICTED REGION -17.54.102 12.04.102 -100 -18.00,90 . gg 12.06.90 - 80 70 PERNISSIBLE OPERATlHG-60 REGION - 50 - 40 - 30 20 --10 t i f f f I f f f 40 30 -20 -10 0 10 20 30 40 50 Axial P0wer Imaalance, 5 D h' 3' E ]D * ]ju *a tfu_ A Jh a j ow POWER INBALANCE ENVFLOPE FOR OPERATION FRDW 0 TO 100 : la Eir0 TNI-1 CYCLE 3 Figure 3 5 2G Amendment No.)W 2 9 3 ;- n 7nn I>UJ JV/
9' 1: sa Po,ier, 5 pf 2535 Mit RESTRICTED REGION, ,,,,3,,,,, 1 - 100 -22.50,90 90 9.49.9o 80 70 PERMISSIBLE OPERATING 60 REGION 50 40 30 20 10 d 8 i i i i 1 e 50 40 30 20 -10 0 10 20 30 40 50 Axial Power Imbalance, 5 PONER IMBALANCE ENVELOPE FCR OPERATION FROM 1001 la TO 248 i 10 EFPD ~ TNl-1 CYCLE 3 Figure 3.5-2H AmendmentNo.)MI 29 l '; o O >i
9= Power, ", of 2535 MWt RESTRICTED REGION -31.21,not 100 -2s.8,90 90 is.2s.so - 80 PERNISSIBLE OPERATING 70 REGION 60 50 - 40 30 - 20 - 10 I f f I I I I I I l 50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power imoalance, "' D@ D iJ tj POWER INBALANCE ENVELOPE FOR OPERATION AFTER 2461 10 EFPD TNI-1 CYCLE 3 Figure 3.5-21 rro ', 'l } j Amendmer.t No. M E
21 20 j 19 18 /' \\ / 17 \\ 3 \\ E 16 E / ^ 15 G _E j 14 13 12 0 2 4 6 8 10 12 Axial location of Peak Power From Bottom of Core, it LOCA LIMITED NAXINUM ALLO
- ABL!
LINEAR HEAT RATE Figure 3 5 2 Amendment No. 20 /1a 1r"o sus ,6 G I
4 'i 18.2.102 34.7.102 100 15.8.90 37.9.90 RESTRICTED 90 REGION 37.s.s0 80 6.s.s0 s 70 6.s.70 45.2.70 0.70 5* 60 ,E 45.2.60 100.60 E
- 50 PERMISSIBLE f 40 OPERATING
=2 REGION 30 20 10 i i i i i i i 0 10 20 30 40 50 60 70. 80 90 100 APSR 5 Witndrawn om or g-cJu,2.N"o J oo APSR POSITION LIMITS FOR OPERATION FROM 0 TO 100 1 10 EFP0 TMI 1 CYCLE 3 Figure 3.5 2K Amendmentko. 29 }{}3 ,i 3
20.s.102 38.8.Io2 100 90
- v.s.go
,,,,,o RESTRICTED 80 s.5.so us.s.so REGION $ 70 4.5.70 45.2.70 O.70 m E 60 e5.2.60 100.s0 50 I E 40 PERMISSIBLE 30 OPERATING REGION 20 10 I i i i i i i i 0 10 20 30 40 50 60 70 60 90 100 APSR % Witndrawn el s APSR POSITION LIMITS FOR OPERATION FROM 100 1 10 TO 246 1 10 EFPD TMI-1 CYCLE 3 Figure 3.5-2L knendment No. 29 7 l ) ;a o 1i' s u
e s.5.102 es o,102 100 2.5.so RESTRICTED 90 REGION 80 2.s.8o 0.8o 70 E 60 y is. o, so soo.so ~ ~ 50 PERNISSIBLE e OPERATING 40 REGION E a. 30 t ~ 20 10 f f I f l' I f f f 0 10 20 30 40 50 60 70 80 90 100 APSR 5 Witnarawn APSR POSITION LIMITS FOR OPERATION AFTER 246 1 10 EFPD TNI-l CYCLE 3 Figure 3.5 2M q r1" lr O >13 Amendment No. 29 Ju
t k.2 REACTOR COOIANT SYSTDI INSERVICE I?!3pECTION Applicability This technical specification applies to the inservice inspection of the reactor coolant system pressure boundary and portions of other safety oriented system pressure boundaries as shown on Figure 4.2-1. Objective The objective of this inservice inspection program is to provide assurance of the continuing integrity of the reac*.ar coolant system while at the sa=e time minimizing radiation exposure to personnel in the performance of inservice inspections. Specification k.2.1 The inservice inspection program to be followed is outlined in Table k.2-1. Except as provided for in this table and as discussed herein, the inservice inspection program is in accordance with the AS E Code, Section XI, Rules for Inservice Inspection of Nuclear Reactor Coolant Systems, dated January 1, 1970, as modified by the Winter 1970 Addenda. Prior to initial plant operation a preoperaticnal inspection of the plant vill be performed of at least the areas listed in the ASME Code; provided accessibility and the necessary inspection techniques are available for each of these areas. The only exception to this vill be areas where the necessary base line data is already available and has been obtained by the same techni tues as vill be used during inservice inspection. 4.2.2 The reactor vessel =aterial surveillance capsules removed from TMI-1 during 1976 shall te inserted, irradiated in and withdrawn from the Three Mile Island Unit No. 2 reactor veesel in accordance with the schedule shown in Table 4.2-2. The licensee shall be responsible for the examination of these specimens and for submission of reportc of test results in accordance with 10 CFR 50, Appendix H. h.2.3 The accessible portions of one reactor coolant pu=p motor flyvheel assembly vill be ultrasonically inspected within 3-1/3 years, two within 6-2/3 years, and all four b-/ the end of the 10 year inspec-tion interval. However, the U.T. procedure is developnental and vill be used only to the extent that it is shown to be meaningful. The extent of coverage vill be limited to those areas of the fly -heel which are accessible without =otor disasse=bly, i.e., can be reached through the access ports. Also, if radiation levels at the lover access ports are prohibitive, only the upper access ports will be used Amendment f!o. 29 k_ll m-J OO 310 e
The inspection schedule may be modified to coincide with those 'k.2.4 refueling or maintenance outages =ost closely approaching the inspection schedule. Sufficient records of each inspection shall be kept to allow comparison h.2 5 and evaluation of future inspections. The inse:-vice inspection shall be reviewed at the end of five years h.2.6 to consider incorporation of new inspection techniques and equipment ~ which have been proven practical, and a possible extension of the The conclusions of this program to additional examination areas. review shall be submitted to the NRC for evaluation. The licensee shall submit a report or application for license a=endment h.2 7 to the NRC within 90 days after the occurrence of either of the following: Failure of Three Mile Island Unit No. 2 to achieve commercial 1. operation at 100% power by October 1,1978, or Beginning one year after attainment of ec=mercial operation at 2. 100% power, any time that 'fbree Mile Island Unit No. 2 fails to maintain a cumulative reactor utilization factor of at least 65%. The report shall provide justification for continued operation of THI-1 vith the reactor vessel surveillance program conducted at Three Kile Island Unit No. 2, or the application for license amend =ent shall propose an alternative program for conduct of the TMI-1 reactor vessel surveillance program. For the purpose of this technical specifd cation, the definition of cce=ercial operation is that given in Regulatory Guide 1.16, Revision b. The definition of cumulative reactor utilization factor is: Cumulative reactor utilization fntor = (Cu=ulative megawatt hours (thermal) since attainment of co==creial operation at 100% power x (100)) divided by 0.icensed power (Irat) x (Cumulative hours since attainment of ec==ercial operation at 100% power)). In ' addition to the reports required by Specification h.2.7, a report h.2.8 shall be submitted to the URC prior to September 1, 1982, which summarizes the first five years of operating experience with the TMI-1 If, at the time integrated surveillance program performed at S I-2. it is desired to continue the surveillance of submission of this report, program at L!I-2, such continuatien shall le justified on the basis of the attained operating experience. Amendment Nb. 29 k-12 1 r ': 4 IJu >l Ei us - -
, Inspection Bases a. The nuclear plant was designed prior to the issuance of Section XI of the ASNE Code, Rules for Inservice Inspection of Nuclear Reactor Coolant Systems dated January 1, 1970. However, sufficient accessibility was included in the design to perform most inspections discussed in the code. The proposed inspection program follows the code except that inspections are focused on araas which engineering analysis has indicated are subject to the more critical stress, radiation, or transient conditions. The areas selected for inspection on this basis are listed in Table 4.2-1. These areas are exposed to the more severe conditions (which are still well within code limits) in the reactor coolant system. Therefore, they are expected to indicate potential problems before significant flaws develop in the selected areas or in other areas. It is considered that the focused approach specified herein will result in a meaningful inspection program in that it will providc assurance of continuink plant integrity. In those areas where inspection methods are developmental, such as for remote inspection of the reactor vessel welds, reactor vessel nozzle inside radii and welds, and ultrasonic inspection of pressurizer support bracket welds, the inspection methods will be developed and tested to the extent practicable during preoperational inspections. (Development of inspection techniques will not be attempted on radioactive equipmert unless necessary to explore a specific problem.) A preoperational inspec-tion is planned of areas listed in the ASME Code whic tre within the inservice inspection boundaries and which are accessible for inspection. However, as discussed above, in areas where inspection methods are develop-mental, the inspections will only be performed to the extent practicable. Once an inspection method is selected for a particular inspection (e.g., U.T. for most volumetric inspections), it is intended that all subsequent inservice inspections be performed using the identical method and on the same component parts wherever practicable. In addition to the above inspection, if any of the components within the inservice inspection boundary are disass.e= bled for maintenance, the accessible parts will be given a normal visual examination as part of the routine plant maintenance operations. b. Because of damage to the surveillance capsule holder tubes originally installed in TMI-1, irradiation of the TMI-l capsules will be conducted in THE-2 pursuant to 10 CFR 50, Appendix H, Section ll.C.4. Because of the similarity of IMI-l and TFE-2, irradiation in TMI-2 will be substan-tially equivalent to irradiation in TMI-1, and appropriate adjustments and margins can be imposed in applying the irradiation data to account for such differences as do exist. The withdrawal schedule has been formulated to optimize the availability of irradiation data from the capsules of both Units 1 and 2. 4-13 i sO lJGa Amendment No. }/, 2 9
g Because the irradiation program is dependent upon the successful opera-tion and a reasonable utilization of TMI-2, reporting requirements are included to permit reevaluation of the progran if 24I-2 does not achieve full power operation in a reasonable period of time or suffers extended outages after the first year of operation. c. The reactor coolant pump motor flywheel ultrasonic test procedure is being developed to detect flaws of a small enough size to provide assurance of continued integrity based upon a conservative fracture mechanic's evaluation. REFERENCE (1).FSAR, Section h.h (2) BAV-10100A, February 1975 h-13a Amendnent No. J, g9 8 1} 1r - D 7 I lsVU "^ w e
' Notes To Table 1. Inspections of the reactor vessel shell velds and other velds in the reactor vessel annulus, vill be atte.pted using remote U.T. inspection equipment from the vessel CD and vill only be inspected to the extent practicable. Further, reactor vessel head penetrations and reactor vessel head and flange velds =ay be inspected using remote, automated U.T. equi;=ent. It is not certain that in all cases the re=cte U.T. equipment can practicably be made to provide =eaningful results, and so=e adjustment to the proposed inspections may be necessary after ec=pletion of preservice development. 2. The = cst probable defects in studs or bolts are cracks starting frc= the reets of threads or in the fillet radil at the ends of the shanks. These =ay be difficult to detect using U.T. Likevise, in nuts, the = cst probable defects may not be revealed by U.T. Therefore, surface inspections ("Ma6naglo") of studs, bolts and nuts may be perfor=ed in lieu of U.T. if joints are dis-asse= bled for other reascns. Ligaments and threads in base material vill be visually inspected when joints are disasse= bled for other reasons. 3. In the event that it is de=enstrated that ultrasonic inspection frc= the outside of a vessel can provide a =eaningful inspection of interier cladding, then U.T. may be substituted for surface or visual inspection of cladding. 0 3 p., 9 ~nn isuJ cU l Section 15, Tech spees. h-27 A=. 41 (t.16-73)
+g r 5' T.\\BLE h.2-2 SURVEILLA!!CE CAPSULE I'TSERTION & WITHDRAWAL SCHEDULE AT TC-2 Schedule _ Capsule Designation Insertion Withdraval DG-1A ' DC-2 Start-up End of 3rd Cycle THI-1B End of 1st Cycle End of 6th Cycle THI-1C End of 3rd Cycle End of lith cycle THI-1D End of 6th Cycle End of 15th Cycle DC-1E Removed end of ist Cycle of D C-1 THI-1F End of loth Cyele End of 24th Cycle \\ W 4-27a pnn j
- ya Amendment No. 29
- ~
~ * - - -
a f. n 4,3 TESTING FOLLOVING OPENING OF SYSTDf glicability .j Applies to test requirements for Reactor Coolant Systen integrity. Objective To assure Reactor Coolant System integrity prior to return to criticality following nomal opening, modification, or repair. Specification k.3 1 When Reactor Coolant System repairs or modifications have been made, these repairs or =odifications shall be inspected and tested to meet all applicable code requirements prior to the reactor being made critical. h.3.2 Following any opening of the Reactor Coolant System, it shall be leak tested at not less than 2285 psig prior to the reactor being made critical. h.3.3 The limitations of Specification 31.2 shall apply. [ Bases Repairs for modifications made to the Reactor Coolant System are inspectable ' and testable under applicable codes, such as 3 31.7, and ASME Boiler and 8 Pressure Vessel Code, Section IX, IS kOO. For nomal opening, the integrity of the Reactor Coolant Syste=, in tems of strength, is unchanged. If the system does not leak at 2285 psig (operating pressure +100 psi; +jio psi is nor=al system pressure fluctuation), it vill be leak ti6ht during nor=al operation.(1) si.r z.r.: :.CE (1) FSAR, Section k fOQ .s V U I-k-28 .}}