ML19210A374
| ML19210A374 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/22/1978 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19210A375 | List: |
| References | |
| NUDOCS 7910290603 | |
| Download: ML19210A374 (18) | |
Text
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S, UNITED STATES
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NUCLEAR REGULATORY COMMisslON j
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WASHINGTON, D. C. 20666
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METROPOLITAN EDISON COMPANY
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JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY
_D_0CXET NG. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Anendnent No. 45 License No. DPR-50 1.
The Nuclear Regulatory Comission (the Connission) has found that:
A.
The application for amendment by Metropolitan Edison Company, Jersey Central Power & Light Company, and Pennsylvania Electric Company (the licensees), dated June 23, 1978, as supplemented August 7, 1978, complies with the standards and requirements of the Atomic Energy Act of 1054, &: amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Connission; 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 conpliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the coninon defense and security or to the healt.'t and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Connission's regulations and all applicable require:nents have been satisfied.
s
$910 200 6o3
l
. 2.
Accordingly, Facility Operating License No. DPR-50 is hereby amended as indicated below and by changes to the Technical Specifications as indicated in the attachment to this license amendment:
A..
Revise paragraph 2.c.(2) to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 45, are hereby incorporated in the license. Metropolitan Edison Company shall operate the facility in accord-ance with the Technical Specifications.
B.
Delete paragraph 2.c.(3) which was added by Amendment No. 40, dated May 19, 1978.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULAT MMISSION
\\
s Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: September 22, 1978
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ATTACHMENT TO LICENSE AMENDMENT N0.45 FACILITY OPERATING LICENSE NO. DPR-50 00CrF7 NO. 50-289 Revise the ' Appendix A Technical Specifications is follows:
Remove Paces Insert Paces vi and vii vi and vii Figure 2.1-2 Figure 2.1-2 2-4 2-4 2-7 thru 2-9 2-7 thru 2-9 Figures 2.3-1 and Figures 2.3-1 and 2.3-2
- 2. 3-2 3-16 3-16 3-34a 3-34a Figure 3.5-2H Figures 3.5-2B, 3.5-2D, 3.5-2F and 3.5-2H The changed areas on the revised pages are shown by marginal lines.
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i LIST OF FIGURES Figuy Title 2.1-1 Core Pre v aton Safety Limit 2.1-2 Core Ptytection Safety Limits 2.1-3 Core Protection Safety Basis 2.3-1 Protection System Maximum Allowable Set Points 2.3-2 Protection System Maximum Allowable Set Points 3.1-1 Reactor Coolant System Heacup Limitations 3.1-2 Reactor Coolant System Cooldown Limitations 3.1-3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter H O 2
3.5-1 Incore Instrumentation Specification Axial Imbalance Indication 3.5-2 Incore Instrumentation Specification Raditi Flux Tilt Indication 3.5-2A Rod Position Limits for 4 Pump Operatien Appix-cable During the Period from 0 to 125 i EFPD; Cycle 4 3.5-23 Rod Position Limits for 4 Pu=p Operation from 125 1 5 EFPD to 265 1 15 EFPD; TMI-1, Cycle 4 3.5-2C Rod Position Limits for 2 and 3 Pump Operation Applicable During the Period from 0 to 125 1 5 EFPD; Cycle 4 3.5-2D Rod Position Limits for 2 and 3 Pu=p Operation from 125 1 5 EFPD to 265 1 15 EFPD; TMI-1, Cycle 4 3.5-2E Power Imbalance Envelope Applicable to Operation from 0 to 125 i 5 EFPD; Cycle 4
, pl.
, 45 1kN Amendment No.
i Figure Title 3.5-2F Power I= balance Envelope for Operation from 125 1 5 EFPD to 265 1 15 EFPD; TMI-1, Cycle 4 3.5-2G LOCA Limited Maximum Allowable Linear Heat Rate 3.5-2H n'SR Position '.imits for Operation from 0 to 1 5 i 15 EFPD; Cycle 4 3.5-2I Deleted 3.5-2J Deleted
- 3. 5-2K Deleted 3.5-2L Deleted
- 3. 5-2M Deleted 3.5-2N Deleted
- 3. 5-3 Incore Instrumentation Specification 4.2-1 Equipment and Piping Requiring Inservice Inspection in Accordance with Section II of the ASME Code 4.4-1 Ring Girder Surveillance 4.4-2 Ring Girder Surveillance Crack Pattern Chart 4.4-3 Ring Girder Surveillance Crack Pattern Chart 4.4-4 Ring Girder Surveillance Crack Pattern Chart 4.4-5 Ring Girder Surveillance Crack Pattern Chart 6-1 Organization Chart 1469 007 vii Amendment No.
,45
Thermal Power Level, 7
-- 120 ONBR Limit (112)
(4s,812)
(-30,112)
- 110 ACCEPTABLE 4 PUMP OPERATION
,gg Kw/Ft Kw/Ft Limit Limit
(-30,87.1) 90 (s7.1) 2 (as,s7.1) r
(-53,80)
ACCEPTABLE 80 po,so) 3&4 PUMP OPERATION 70
(-30,59.6) 60 (5s,3) 3 (s4,59.6)
(- 53,55. I )
(5s.4.58.4)
ACCEPTABLE
(-42. 4,42. 4) 2,3 & 4 PUMP 50 OPERATION 0s.2.49.2)
~
40 30 20 10 I
f I
e t
t I
I f
f I
f
- 50
-40 20
-10 0
10 20 30 40 50 60 Reactor Power imbalance, ".
Curve Reactor Co, ! ant Flow (Ib/hr) 6 I
139.8 x 10 6
2 104.5 x 10 6
3 68.8 x 10 CORE PROTECTION SAFETY LIMITS TH I-1, Cycle h Figure 2.1-2 no. ll
, 45l Amendment 1A69 h
l 2.2 SA M Y LIMITS - ETACTOR SYSTEt PRESSURE Anolicability
' Applies to the 14-4t on reactor coolant system pressure.
Obieetive To maintain the integrity of the reactor coolant system and to prevent the release of significant a=ounts of fission product activity.
Soecificatien 2.2.1 The reactor coolant syste= pressure shall not exceed 2750 psig when there are fuel assenblies in the reactor vessel Bases N
The reactor coolszt syste= (1) serves as a barrier to prevent radionuclides in the reactor coolant frc= reaching the atmosphere.
In the event of e. fuel cladding failure, the reae:cr coolant system is a barrier against the release of fissics products.
Establishing a syste= pressure li=it helps to assure the intes-ity of the reactor coolast system.
"he d-n transient pressure allovabic in the reactor coolant system is 110% of design pressure. (2) pressure vessel under the ASME Code,Section III, The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under ANSI Section B31.7 is 110% of design pressure. Thus, the safety limit of 2750 psig (110%
of the 2500 psig design pressure) has been established.(2) The maximum settings for the reactor high pressure trip (2390' psig) and the pressurizer l
code safety valves (2500 psig)(3) have been established in accordance with ASMI Boiler and Pressu e Vessel Code, Sectics III, ArtTele 9,1' inter, 1968 to assure that the reactor coolant syste= pressure safety limit is not exceeded.
. The initial hydrestatic test vas conducted at 3125 psig (125% of desie,n pressure) to verify the integrity of the reactor ecolant syste=.
Additional assurance that the reactor coolast systes pressure does not exceed the safety li=it in provided by secting the pressurizer electrc=sti: relief valve at 2255 psig. (h)
Seferences (1) FSAR, Section h
<g (2) FSAR, Section h.3.10.1 o-J elfdd u b O' (3) FSAR, Section h.2.k.
(h) FSAR, Table 4-1
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, 45 Amendment No.
i Reactor coolant system pressure 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 shown in Figure 2.3-1 'or high reactor coolant system pressure has been established to ma stain the system pressure below the safety limit (2750 psig) for.ny design transient.(6) Due to calibration and I
instrument errors, the safety analysis assums' a 45 psi pressure error in the high reactor coolant system pressure trip setting.
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 egual to 1.3 for those design accidents that result in a pressure reduction (3, ),
Due to the calibration and instrumentation errors, the safety analysis used a variable low reactor coolant system pressure trip value of (11.75 Tout -
5143) and a low pressure trip value of 1770 psig.
d.
Coolant outlet te=perature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperature in the operating range. -
The calibrated range of the temperature channels of the RPS is 520 to 620 F.
The trip setpoint of the channel is 619 F.
Under the worst case environment, power supply perturbations, and drif t, the accuracy of the trip string is 11F. This accuracy was arrived at by su= ming the worst case accuracies of each module. This is a conservative method of error analysis since the normal procedure is to use the root mean square method.
Therefore, it is assured that a trip will occur at a value no higher than 620 F even under worst case conditions. The safe analysis used a high temperature trip set point of 620 F.
The calibrated range of the channel is that portfon of the span of indidation which has been qualified with regard to drift, linearity, repeatability, etc.
This does not imply that the equipment is restricted to operation within the calibrated range. Additional testing haa demonstrated that in fact, the temperature channel is fully operational approximately 10% above the calibrated range.
Since it has been established that the channel will trip at a value of RC outlet temperature no higher than 620 F even in the worst case, and since the channel is fully operational approximately 10% above the calibrated range and exhibits no hysteresis or foldover character-1stics, it is concluded that the instrument demign is acceptable.
Reactor building pressure e.
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.
Amendment No.
.. 45 gg @O 2-7
f f.
Shutdown bypass i
j In order to provide for control rod drive tests, zero power physics
)
testing, and startup procedures, there is provision for bypassing certain seg=ents of the reactor protection syste=.
The reactor protection syste= seg=ents which can be bypassed are shovn in Table 2 3-1.
Two conditions are i=pgsed when the bypass is used:
1.
By ad=inistrative contrcl the nuclear overpever trip set point must be reduced to a value 15 0 percent of rated pcver during reactor shutdevn.
2.
A high reactor coolant syste= pressure trip set point of 1720 psig is automatically i= posed.
The purpose of the 1720 psig high pressure trip set point is to prevent nor=al operation yith part of the reactor protection s/ste:
bypassed.
This high pressure trip set point is lover than the nor=al lov pressure trip set point so thet the reactor must be i
tripped before the bypass is initisted.
The overpcVer trip set i
point of 15.0 percent prevents any significant reactor pcver i
fro = being produced when perfor=ing the physics tests. Sufficient natural circulation (5) veuld be available to remove 5 0 percent j
of rsted pcVer if nene of the reactor coolant pumps were operatir.g.
REFERE::CES l
(1)
FSAR, Section lb.1.2.3 (2)
FSAR, Section ik.1.2.2 (3)
FSAR, Section ik.l.2.7 (k)
FSAR, Section ik.l.2.8 (5)
FSAR, Section lb.1.2.6 (6)
Technical Specification Change Request No. 31, January 16, 1976, and Tcchn* cal Specificatic-Change Request No. 84, June 23, 1978.
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Amendment No.45 2-8
5 TAhlLE2.3-1 0
g-REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS a"
Four Reactor Coolant Three Reactor Coolant One Reactor Coolant g
Pumps Op-rating Bunps Operating pump Operating in (Nominal Operating (Nominni Operat.ing Each Loop (Nominal Shutdown K
power - 100%)
power - 75%)
Operating Power
':9%)
Dypass K" 1.
Nuclear power, Max.
105 5 105 5 105 5 5 0 (3)
% of rated power 2.
Nuct a power based on 1.00 times flow minus 1.08 times flow minus 1.08 'imes flow minus Bypassed flow 2 and imbninnce reduction due to reduction doo to reduc on due to y
max. of rated pov r imbalance (s) innbninnec(s) imba
.ce(s)
U1 3
glearpowerbased rtA NA 91%
Bypassed on pump monitors, muix. % of rated power 1
High reactor coolant 2390 2390 2390 1720 8
y eyotem prenoure, psig, M) mfix.
5 Low reactor coolant 1800 1800 1800 Byp' sed eyeten pressure, psig min.
6.
Variable low reactor (11 75 Tout-5103) (1)
(11.T5 Toit'-5103) (1) 11 75 rout-5103) (1) nypassed coolant system pressure p618, aln.
T.
Reactor coolant teop.
619 619 619 619 F.
Max.
/
8.
High Beactor Dailding Is Is Is II e
P' pressure, peig, meat.
I d
(1)
Tout is in degrees Fahrenheit (F)
(2) ' Henctor coolant system flow, I (3) Administrative 1y controlled reduction set ormly during reactor shutdown g
(4) Autonvitically set when other segments of the RFS (no specified) arc bypassed (5) The pump monitors also produce a trip on:
(n) loss of two reactor coolant pumps in one reactor coolant. loop, and (b) 1000 of one or two reactor coolant pun.ps during t"vo-pump o ation (6)
Trip nettinca limits are set. ting limits on the setpoint oide of the protection syst$m bistable cornparators I
2500 P = -2390 psig 2300 E.
ACCEPTABLE p
OPERATION T = 619 F 2100 e=
%h
@Wj 2
1500 UNACCEPTABLE g
P = 1800 psig OPERATION 1700 1500 540 560 580 600 620 640 Reactor Outlet Temperature. F TMi-1 PROTECTION SYSTEM MAI! MUM ALLOWABLE SET POINTS Figure 2.3 1 s
AadmentNo.h,[,
, 45 g g}3
Thermal Power Level, 5 120
(-17,108) 110 (108) (17,108)
[
100 Mi = 1.28 ACCEPTABLE M2 = -1.0 i
-9
( 5,90)
OPER. TION
(-35,85)
I (80.7) ou ACCEPTABLE 70 3 & 4 PUMP OPERATION (35,62.7) 0
(-35,57.7) l 1
(53.1) 50 l
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ACCEPTABLE 40 2,3 & 4 PU'JP
( -35,3 0.1 )
OPERATI ON (35,35.1) l
-- 30 l
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- 20 o
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e,
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It a.i
=-
10 I
i, I il il E
l;-
E I
t i
I f
t I t
t f
f I
f I
f
-70
-60
-50
-40
-30
-20
-10 0
- 10 20 30 40 50 60 70 Reactor Power imbalance, 5 PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS FOR REACTOR POWER IMSALANCE TMl 1, cycle h
s Figure 2.3-2 hendant No.
N
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3J. 7 NODERATOR TDGERATURE COL'TFICL.3T OF REACTIYrn Applicability Applies to anximu= positive moderator te=perature coefficient of reactivity at full power conditions.
p Objective a
To assure that the moderator temperature coefficient stays within the limits calculated for safe operation of the reactor.
Specification 3.1.7.1 The moderator te=peratuie coefficient shall not be posit,1ve at pcver
- 1evels above 95% of rated pcver.
3.1.7.2 The moderator temperatare coefficient shall be 1 + 0.5x10-4 Ak/k/F at power levels 1 95% of rated power.
Bases A non-positive mcderator et.Micient at power levels above 95% of rated pcver is specified such that the maxim.=t clad te=pe stures vill not exceed the Final Acceptance Criteria based en LOCA analyses. Selev 95% of rated pcver the Final Acceptance criteria vill get be exceeded with a positive moderater te=perature coefficient of +f 5 x 10- 6K/K/F. All other accident analyses as r rerted in the 75AR have been performed for a rar.ge of =oderator temperature coetalcients including +0 5 x 10 k AK/K/F.'
A non-positive moderator coefficient at power levcis above 95% of rated power is also required to prevent overpressurization of the reactor coolant system in the event of a feedvater line break (see Specification 2.3.1, Basis C, Reactor Coolant System Pressure).
The experi:mtal value of the moderater coefficient will be corrected to obtain the hot f C pcver moderator coefficient. The correction factor vill be verified during startup testing on earlier EW reactors.
The Final Acceptance Criteria states that post-LOCA clad temperature vill not exceed 2200 F.
RETERL"' ICES Qgh]h[n\\
p" erj (1)
FSAR, Section ik I
(2)
FSAR, Secticu 3 s
AmendmentNo.;/
, 45
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3_16
2.
The control rod group vithdrawal limits (Figures 3.$-2A, 3 5-23, 3 5-2c, 3 5-2D, and 3.5-2H, shall be reduced 2 1ercent in power for each 1 percent tilt in excess of thc tilt limit.
3.
The operational inhalance limits (Figures 3.!-2E, and 3 5-2F) shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limi%
f.
Except for physics or diagnost'ic testing, if quadrant tilt is in excess of +26.T5% deter =ined using the full incore detector system (7IT), or +15 215 determined using.the minimum incere detector system (MIT) if the FIT is not available, or +22 965 l
determined using the out of core detector system (OCT) when neither the FIT nor MIT are available, the reactor vill be placed in the het shutdown condition. Diagnesite testing during power operation with a quadrant tilt is permitted provided that the ther=al power allevable is restricted as stated in 3 5 2.h.d above.
g.
Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15 percent of rated pover.
9 a
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1469 016
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Amendment No. '
' ' 45 3-3ha
g POWER 234.102 274.1.102 a
100 NOT ALLOWED LEVEL N
I:Uf0ff 925
=
214.1.02 88 ESTRICTED
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240.2.00 70 SHUT 00EN NARGIN LINii 00,10 60
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PERNIS$10LE 50 113,50 188.53 OPERAllNG REGION 40 -
30 -
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122,15 140.15 10 -
0.2.3 0
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G 25 50 75 100 125 150 175 200 225 250 215 300 Rod Inden, 5 tithdrasn O
25 50 15
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0 25 50 75 10,0 i
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98 0
25 50 15 100 ROD P03lT10N LIMITS FOR Il PUNP OPERATION w
8 i
8 Group 5 FROM 125 1 5 EFPD TO 265 1 15 EFPD O
TNi-l, Cycle I 4
Figure 3.5-25 M
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stat-oigy areH Jo % 'Ja.ed 45 A: lend: tent No.
s,
Power, % of 2535 MWt
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RESTRICTED AEGION
-22.25,102 10.25,102
_ _ 100
-23.01,92 11.26,92 o
f
-27.72,80' 80 " 11.26,80 70 60 PERMISSIBLE OPERATING 50 REGION 40 30 20 10 t
iI i
i i
i -40
-30
-20
-10 0
10 20 30 40 50 Axial Power imbalance, %
POWER IMBALANCE ENVELOPE FOR OPERATION FROM 125 1 5 TO 265 1 15 EFPD TH I - 1, cycle h Figure 3.5-2F
~-
Amendment No.
45 ass 09
57.9,102 100
'8 RESTRICTED 90 80 < 0.80 70 g
100,70
[
G n
60 o
50 PERMISSISLE E
40 OPERA 71NG REGION 30 20 10 t
t t
t I
f f
f
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10 20 30 40 50 60 70 80 90 100 APSR. #. Witherawn APSR POSITION LIMITS FOR OPERATION FROH 0 TO 255 f 15 EFPD 791-1
' Q gare_ 3.5-2d Amendment No.
, 45 g2Q