ML19254E761
| ML19254E761 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 10/29/1979 |
| From: | PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19254E759 | List: |
| References | |
| NUDOCS 7911020298 | |
| Download: ML19254E761 (3) | |
Text
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LCA 51, Revision 1 Page 1 of 3 LICENSE CHANGE APPLICATION 51 The current maximum allowable Fg limit is reduced from 2.32 to 2.25.
Proposed replacement pages are included as Attachment A.
REASON FOR CRANGE An error in the Westinghouse ECCS evaluation resulted in an underestimate of the heating effect of the zirconium-water recction. Correcting this error leads to a higher calculated peak cladding temperature (PCT).
To meet the 10 CFR 50.46 PCT limit of 2200*F, the maximum allowable total peaking factor (F ) must be reduced from its current limit.
Q The NRC issued an exempticn on December 20, 1978 permitting operation of Trojan until April 1, 1979 without an ECCS evaluation which conforms to the requirements of 10 CFR 50.46(a)(1).
This LCA provides the required evaluation and will permit continued operation of Trojan beyond April 1, 1979.
SAFETY / ENVIRONMENTAL EVALUATION The attached recalculation of the Trojan ECCS performance was completed by Westinghouse using a corrected ECCS Evaluction Model.
The NRC has previously reviewed the methodology used by Westinghouse and has advised PGE that its application to Trojan is acceptable.
The Westinghouse recalculation constitutes the safety evaluation for the proposed Techni-cal Specification change.
It is demonstrated that, provided the maximum allowable Fq is reduced to 2.25, the Trojan ECCS performance meets the requirements of 10 ?FR 50.46 and Appendix K.
There are no environ-mental effects associated with the proposed change.
Reanalysis of Trojan ECCS Performances with Evaluation Model Corrected for Zircaloy-Water Heat of Reaction Error Introduction The currently applicable ECCS evaluation 4: r Trojan was forwarded f rom PGE to the NRC in a letter dated March 31, 1977.
This evaluation used the NRC-approved " October 1975" model.
On April 12, 1976, PGE notified the NRC that this evaluation was deficient in that the zircaloy-water heat of reaction was incorrectly treated. The NRC, on December 20, 1978, issued an exemption to PGE from the requirements of 10 CFR 50.46(a)(1) to permit interim operation until April 1, 1979 pending submittal of a corrected ECCS analysis.
Westinghouse has since developed an improved model (the " February 1978" model) that is the current NRC-approved Westinghouse ECCS Evaluation Model. This model correctly treats the zircaloy-water heat of reaction.
The present analysis does not employ the " February 1978" model; rather, the " October 1975" model was used with the LOCTA code appropriately 1249 122 7 911020 }[jy' }f'
LCA 51 Attachment A Page 1 of 3 POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-Fn(Z)
LIMITING CONDITION FOR OPERATION F (Z) shall be limited by the following relationships:
3.2.2 Q
F (Z) $ [2.25] [K(Z)] for P > 0.5 Q
P F (Z) $ [(4.50)] [K(Z)] for P f 0.5 Q
where P = THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY: MODE 1 ACTION:
With F (Z) exceeding its limit:
Q a.
Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds Q
the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent STARTUP and POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% Fo(Z) exceeds the limit. The Over-power AT Trip Setpoint reduction shall be perfomed with the reactor subtritical.
b.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore Q
mapping to be within its limit.
1249 123 TROJAN-UNIT 1 3/4 2-5 Amendment No. 6
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LCA 51 o 3 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:
(a) maintaining the minimum DNBR in the core >1.30 during nomal operation and in short tem transients, and (b) 1Tmiting the fission gas release, fuel pellet temperature & cladding mechanical properties to within assumed design criteria.
In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.
l The definitions of certain hot channel and peaking factors as used l
in these specifications are as follows:
F (Z)
Heat Flux Hot Channel Factor, is defined as the maximum local g
heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for man-ufacturing tolerances on fuel pellets and rods.
F[H Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.
Fxy(Z)
Radial Peaking Factor, is defined as the ratio of peak
/\\
power density to average power density in the horizontal
/30\\
plane at core elevation Z.
3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)
The limits on AXIAL FLUX DIFFEREtCE assure that the F bound envelope of 2.25 times the noma 11 zed axial peaking f9c(Z) upper tor is not l
exceeded during either normal operation or in the event of xenon redis-tribution following power changes.
Target flux difference is detennined at equilibrium xenon conditions with the part length control rods withdrawn from the core. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels.
The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions.
Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burcap considerations.
TROJAN-UNIT 1 B 3/4 2-1 Amendment No. 30 June 22, 1978 1249 125