ML20078R064
| ML20078R064 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/28/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp |
| Shared Package | |
| ML20078R065 | List: |
| References | |
| DPR-50-A-090 NUDOCS 8311140330 | |
| Download: ML20078R064 (7) | |
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k UNITED STATES NUCLEAR REGULATORY COMMISSION wAsMmoros,p.c. sones 5
METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION
'DOCXET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION. UNIT NO. 1 AMENDMENT TO FACILITY GPERATING LICENSE Amendment No. 90 License No. DPR-50 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for anendment by GPU Nuclear Corporation, et al (the licensees), dated June 20,1983, complies 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; 8.
The facility will operate'in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; i
C.
There is reasonable assurance (i) that the activities autnorized 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 Commission's regulations; D.
The issuance of this amendment will not be inimical to the connon 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.
4 4
P
2-2.
Accordingly, the ' license is amended by changes to the Technical 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:
Technical Speciffcations
- The Technical Specifications contained in Appe'. dix A, r
as revised through Amendment No.90, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COP 91ISSION I
slo k F. Stolz, phief
)/.
O rating Reactors Branch #4 vision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 28, 1933 i
e 4
9 i
I ATTACHMENT TO LICENSE AMENDMENT NO. 90 FACILITY OPERATING LICENSE N0. DPR-50 DOCKET NO. 50-2'89 t
Replace the following pa with the enclosed pages.ges of the Appendix "A" Technical Specifications The revised pages are identified by Amendment number and contain vertical lines in~dicating the area of change.
Remove Pages Insert Pages 2-2 2-2 2-9 2-9 3 34 3-34 Fig. 3.5-2G Fig. 3.5-2G E
F 9
L F
1 a' conservative margin to DN3 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 is actually measured.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.3 is predicted for the maximum pesgble thatual power (112 percent) when the reactor coolant flow is 139.8 x 10 lbs/h, which is less than.the actual flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors (2) with potential fuel densification and fuel rod bowing effects; F
= 2.57, F
= 1.71; F
= 1.50 4H The 1.5 axial peaking factor associated with the cosine flux shape provides a lesser margin to a DNBR of 1.3 than the 1.7 axial peaking factor associated with a lower core f p distribution. For this reason the cosine flux shape 4
and the associated r,' = 1.50 is more limiting and thus the mere conservative 7
assumption.
The 1.50 cosine a ial flux shape in conjunction with FAH = 1.71 define' the A
reference design peaking condition in the core for operation at the maximum overpower. ' Once the reference peaking condition and the associated thermal-hydraulic situation has been established for the hot channel, then all other combinations of axial-flux shapes and their accompanying radials must result in a condition which will not violate the previously established design
- criteria on DNBR. The flux shapes examined include a wide range of positive and negative offset for steady staen and transient conditions.
These ' design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn co maximum i
allowable control rod. insertion, and form the core DNBR design basis.
I The curves of Figure 2.1-2 are based on the more restrictive of two ther=al limits and include the effects of potencial fuel densification and fuel rod bowing; a.
The 1.3 DN3R 11mit produced by a nuclear pcVer peaking factor of F 3= 2.57 of the ce=bination of the radial peak, axialpeak,$ndpositienoftheaxialpeakthatyicidsno 1ess than.1.3 DN3R.
b.
The. combination of radial and axial peak that prevents central fuel melting at the hot spot.
The limit is 20.15 kW/ft.
Power peaking is not a directly observable quantity and therefore limits ha've been established on the easis cf the reactor power imbalance preduced i
by the power peaking.
The specified flow races 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.
i
' Amendment No. 17, 39, 90 2-2
HEMiltlH PROTECrillH SYSTEH TRIP SETTINC LIHlTS s
'Funr Reactor. Coolant Tliree Reactor Coolant One Reactor Coolant i,8, l'ineupu Operating eneqis Operating Pusep.Operatisig in ji-
'(Noailual Operating (Nooninal Operating Each Loop (Nominal Slintdown r
Power - 100%)
. Power - 75%)
Operating Power
'492)
Bypass P
g.
1.
Nuclear power, Hax.
105.5 105.5 105.5 5.G(3)
I of rated power
-wN
.s. 2.
Nuclear power leased on
'I.08 times flow 1.08 tinies flow '
1.08 timeu flow minns Bypassed Ilow '(2) and ins 12alasice minns reduction due s:Inus reduction due reduction due to m.ix. of rated power to imitenlance to Imbalance limbalance w
- ",3.
Niiclear power liascel
,-NA '
NA 55%
nypassed I
(5). on pung eminitora, g
Hax. I of rated power s,
$ ' 4.
Illgli reactor coolant sys-2300 2300 2300
[
1720(4)'
tem pressure, psig max.
7 Y
e, ao 5.
l.ow reactor coolant sys-1900 1900
'1900 Bypassed' tem pressure, pulg niin.
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6.
varlaisle low reactor (11.75 Tout-5103)(1)
(11.75 Tout-5103)(1)
(11.75 Tont-5103)(1) sypassed roo l.iial P&y s t elei lsi efs-
- uni c pu lg, sein.
1.
Heact or coolant temp.
Isl 9 619 619 619 Y., nax.
r.
II. Illgli He. ort or Heil lalleth 4
4 4
4
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l' r esa'su r e, gua l g.
max.
(!) Tout in in denseer+ l'alia culac i t (F)
H (2) Heactor cool:ent systems flow, %
( 3) Ailmiinlut rat ively ront rol led reiluct iosi set. only during reactor aliutdown.
t (4) Antuinat leally net wison ot tii:r udgmenta of tlee NPS (as specified) are leypassed.
}i
( 5) 'llie puisp iman i t o rs a l s o pi ncluen a trip on:
(a) loss of two reactor coolant pinops in one reactor coolasit loop. anel (le) lons of one or two reactor coolant pisaps diaring two-pump operatiosi.
((3) Trip nettings limita are set Ling limits oei tlie setpoint side of Llie protection systein Isistable connectors.
f f
< ~.
,l
' P' f,
- 7,
7 c.
l l
f; if a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2., operation may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 4.7.1.2.
-t e
3 If the inoperable rod in Paragraph "e" above is in groups 5, 6, 7 or 8 the other rods in the group may be trimmed to the same 4
position.
Normal operation of 100 percent of the thermal power allowable for the reactor ecolant pump combination nay then v
continue provided that the rod that was declared inoperable is s
maintained within allowable group. average position limits in 3.5.2.5.
[ 3.5. 2. 3 The worth of ' single inserted control rods during criticality is s.
limited by the restrictions of Specification 3.1.3.5 and the control q,
Rod Position Limits defined.in Specification 3.5.2.5.
3.5.2.4 Quadrant tilt:
s y
a.
Except for physics tests the quadrant tilt shall not exceed
+3.52% as determined using the full incere detector system.
b.
When the full incore detector system is not available and except for physics tests quadrant tilt shall not exceed +1.96% as determined using the power range channels displayed on the console each quadrant (out of core detection system).
y A
- c.. Wen neither detector system above is available and, except for physics tests, quadrant tilt shall not exceed +1.90 as determined using the minimum incere detector system.
d.- Except for physics tests if quadrant tilt exceeds the tilt limit power shall be reduced ismediately to below the power level cutoff (see Figures 3.5-2A, and 3.5-23). Moreover'; tne power level cutoff value shall be ' reduced 2 percent for each 'l per:ent tilt in excess of the tilt limit. For less than four ptap operation, thermal
. power shall be reduced 2 percent of the thermal power allowable for the reactor coolant pump combination for each 1 percent tilt in excess of the tilt limit.
\\
e.
k*ithin a. period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be g
- reduced to less than the tilt li=ic except for physics tests, or a
the following adjust =ents in setpoints and li its shall be =ade:
1.
The protection system reactor power / imbalance envelope trip x,
setpoints shall be reduced 2 percent in power for each 1 percent N cile.'
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