ML20085G159
| ML20085G159 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/18/1991 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20085G158 | List: |
| References | |
| NUDOCS 9110240166 | |
| Download: ML20085G159 (7) | |
Text
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Z o
TABLE 4.3.1.
x 3E M$
REACT 6R PROTECTION SYSTEM INSTRUMENTAT
.RVEILLANCE REQUIREMENTS 2-g OG; CHANNEL OPERATIONAL
\\
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH D
@ $g'FUNCTIONAL UNIT CHECK -
TEST CALIBRATION (,)
SURVEILLANCE REQUIRED c
o R$
o.f 1.
a-a.
Neutron Flux - High S/U,5,(b)
S/U, W R
2
' 85 5
W R
3, 4, S 325
'I-b.
Inoperative HA W
NA 2,3,4,S II) 2.
Average Power Range Monitor:
S/U S.(b)
S/U, W SA 2
a.
Neutron Flux - High, '
S W
SA 3, S Setdown b.
Flow Biased Simulated W(d)(e), SA, R(I) 1 Thermal Power - High 5
Q y
W(d), SA 1
c.
Neutron Flux - High 5
Q d.
Inoperative NA Q
NA 1, 2, 3, S 3.
Reactor Vessel Steam Dome fd)
Pressure - High 5
Q R(9) 1, 2 4.
I91 1, 2 Low, level 3 5
Q R
I k
5.
R(g) 1 High, level 8 5
Q B
k 6.
Main Steam Line Isolation Valve - Closure NA Q
R 1
2 9
7.
Main Steam Line Radiation -
[
High 5
Q R
1,2(d)
I9)
I) 1, 2 8.
Drywell Pressure - High 5
Q R
TABLE 4.3.1.1-1 (Continued)
Z g
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS r
}
5 CHANNEL OPERATIONAL o
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH S.
{
FUNCTIONAL UNIT CMCK TEST CALIBRATION SURVEILLANCE REQUIRED y
/.
9.
Scram Discharge Volume Water D
5 Level - High
[
a.
Transmitter / Trip Unit S
Q R(9) 1, 2, SII)
II) b.
Float Switch NA Q
R 1, 2, S 10.
Turbine Stop Valve - Closure S
Q R(9) 1 11.
Turbine Control Valve Fast Closure Valve Trip System Lil Pressure - Low S
Q R(9) 1 12.
Reactor Mode Switch Shutdown Position NA R
NA 1,2,3,4,5 13.
Manual Scram NA W
NA 1,2,3,4,5 Y
m (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) The IRH and SRM channels shall !;e determined to overlap for at least 1/2 decade during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be deter-mined to overlap for at least 1/2 Jecade during each controlled shutdown, if not perfcrmed within the previous 7 days.
(c) [ DELETED]
(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED Deleted THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.
F (e)
This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a g
calibrated flow signal.
(f)
The LPRMs shall be calibrated at least once per 1000 MWD /T using the TIP system.
g (g)
Calibrate trip unit at least once per 92 days.
(h)VVerify secsvred drive ficw to bc less thca er equal to estcblished dri;c flew at the existing fic.; con-g trol valvc positiafw-(i) This calibration shall consist of verifying the 611 second simulated thermal power time constant.
O (j) Not applicable when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.
(k) Not applicable when.DRYWELL INTEGR PY is not required.
(1) Applicable with any control rod withdrawn.
Not applicable to control rods removed per Specifica-Lion 3.9.10.1 or 3.9.10.2.
' to GNRO-91/00176 LOSS OF FEEDWATER llEATING (LFWil) ANALYSES l
l 09110162/SNLICFLR - 13
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Attachnent 3 to GNRO-91/00176
-Page 1 of 4 LOSS OP FEEDWATER llEATING (LFWil) ANA!,YSES -
The LFWil analysis is central to the issues surrounding the requested-change to the simulated; thermal-power (STP) trip surveillance requirement because historically it is' the only safety analysis for which STP trip credit has been taken.
The analysis methodology for this event changed significantly during Cycle 1.
As originally licensed for Cycle 1, the LFWil analysis credited the STP trip to minimize the calculated severity of the transient._ An improved analytic methodology, first employed during the Maximum Extended
-Operating Domain (MEOD) analyses during Cycle 1 and then repeated for each reload, assumes no STP. trip (or any trip for that matter).
1.
Initial Cycle 1 (FWil Analysis The initial Cycle 1 LFWH analysis is discussed in Section 15.1.1 of the. Grand-Gulf FSAR.
With respect to'the STP trip (also called the thermal power monitor trip), Subsection 15.1.1.2.2 of the FSAR notes:
"The thermal power monitor (TPM) is the primary protection system trip in mitigating the consequences of this event.
If'there was no high. thermal power trip scram design available in the Grand Gulf plant design, reactor scram during the loss of feedwater heating transient would occur when the neutron flux exceeds the high APRM flux scram setpoint. Usually,'the high APRM flux scram set point is higher than the high thermal power scram set point by approximately 6 to 8 percent.
Therefore, the-
-loss of feedwater heating transient would be more severe without the high thermal power trip scram design. This would-Icad to a higher operating CPR limit _and reduce the flexibility of plant operation."
In other words, rather than revising the CPR limit which would have unduly-restricted plant operation, Grand _ Gulf chose to credit the STP trip for Cycle 1.
In some sense, the need to credit the STP trip was an artificfal situation necessitated by the limitations imposed by the GE computer codes and analysis methodology at that time.
Subsection 15.1.1.3.1 of the FSAR and its associated references discuan the matheraatical modeling (including a point kinetics core.
model) employed for the initial Cycle 1 analysis of the LFWil event.
As discussed in the next section,- point kinetics models yield extremely _ conservative results for the LFWil event.
2.
Change in Analysis Methodology Subsequent to' the' initial Cycle 1 analyses for Grand Gulf, GE (and the industry) changed its approach to modeling and analyzing LFWil events.
Rather than a transient analysis of the event, a steady state code was employed which examined steady state power level and G9110162/SNLICFLR - 14
~ Attachment'3 to GNRO-91/00176 Page 2 of 4 e
other. core conditions before and af ter the LFWil.
This approach, which'was used on Grand Gulf for the MEOD and all subsequent analyses discussed below, eliminated the need to credit any plant trip (including-the STp trap) in the course of demonstrat'ng. acceptable CPR results.
- Some additional background concerning the change in LFWil analysis methodology may be helpful.
A loss of feedwater heating event results in a decrease in feedwater temperature entering the reactor vessel. This lowers the t emperature of the water-entering the core (i.e., increases subcooling), reduces the average core void fraction and thereby causes-the core average power to increase. The change in subcooling causes a void redistribution and a corresponding power redistribution. While the not change is a total core power increase, not all areas of the core increase at the same rate. This tends to flatten the radini power distribution.
Due to the stored energy and mixing of feedwater.in the downcomer and lower plenum with the recirculation flow, the temperature reduction at the core inlet (and, therefore, the power increase) occur relatively slowly - on the order of seconds to minutes. The relative slowness of this-event allows the core to maintain a quasi-equilibrium in which the water temperature, core power and neutron flux distribution maintain their steady stato relationships.
3
- The LFWH analysis presented in FSAR Section 15.1.1 for Cycle 1-evaluated this event with a transient thermal hydraulics code using a point kinetics model to represent the core foodback mechanisms.
The point kinetics model is a simplistic representation of the actual
. phenomena it cannot represent the flux redistribution that occurs in this_ event-and must assume kinetics paramotors that bound all the statepoints in the event. Consequently, the point kinetics model results in a large overprediction of the LFWil core power increase.
Y Because the LFWil event is relatively slow and-the transient is smooth with no' sudden increasca or decreases of important parameters, the conditions at'the beginning and end of the event bound those throughout'the transient. Tnerefore, performing the LFWil analysis with a three-dinensionni quasi-steady state code is appropriate.
Essentially, a not steady state core power level-(and other steady
- state core conditions) are determined based on the positive reactivity addition of the colder feedwater,-and the effect on CPR is calculated. - Since the radial power redistribution ef fects are accounted _for in tha methodology, the excessive core power overprediction and associated need to credit a trip is eliminated.
3.
Maximum-Extended Operetinh Domain (MEOD) - Cycle 1 During Cycle 1 GE conducted and NRC approved analyses to support Grand Gulf operation In the MEOD, G9110162/SNLICFLR - 15
1 Attachment 3 to GNRO-91/00176-Page 3 of 4 o
A'_ bounding analysis was performed using a standard BWR/6 plant. All.
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-abnormal operating transients, including the LFWil event were examined.- In analyzing the LFWil event, GE employed the revised steady - state approach discussed above.
At. the request of the NRC staf f, a plant specific LFWil event analysis was. performed-using the GE three dimensional BWR Core Simalator (Reference 1) documented in the NRC approved GESTAR Amendment (Reference 2)..This analysis was submitted to the NRC by Reference 3 and approved by the NRC by letter dated August 15, 1986-(Referenct 4).-
- The analysis was performed for the GE P8X8R fueled Grand Gulf Cy:le 1
- core and was shown to be bounded-by the generic analysis. Neither the generic BWR/6 LFWil analysis nor the Grand Gulf-specific LFWil analysis assumed credit for the STP, or any other, trip.
The results of these analyses were incorporated into Appendix 15D of
.the Grand Gulf FSAR (Revision 2) in December, 1987.
-4.
Cycle 2 Reload and L'fVII Event Analysis Grand Gulf contracted with Advanced Nuclear' Fuels (ANP) to provide reload--fuel for Cycle 2 and to conduct the reload analyses necessary
-to obtain NRC approval for the fuel cycle.
The Cycle 2 reload analyses for-LFWil utilized' the ANF 3-D XTGBWR steady state cote simulator. code which is similar to the GE code used in the MEOD analysis of the LFWil event. The ANF' generic analysis of the LFWH event is documented in Reference 5.
!~
At _ the reqhest of the Staff, Grand _ Gulf ' submitted by Reference 6 a plant specific analysis of the LFWil event utilizing the XTGBWR code.
B This~ analysis provided the steady state conditions of the core before and af ter the LFWil event and the resulting delta CPR. No credit was L
.t taken for.the STP trip.
The -Staff's review 'and approval of the Cycle 2 analysis is documented in Reference 7.'
5.
Subsequent Reloads p
L
' The LFWii. event has been consistently analyzed for each Grand Gulf k
reload:using.a steady state approach and NRC approved. methodology.
'In no instance was the STP trip (or any trip) credited for achieving l
acceptable CPR results.-
~
In particu1ar, the IFWil analysis for the current cycle of operation (Cycle 5) emoloyed a relatively new ANF computer code t
l E
-G9110162/SNLICFLR - 16
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'*1 to GNRO-91/00176
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1 Page 4 of 4
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-(CASM0/MICROBURN), albeit the same steady state approach described above which did not involve credit for a plant trip.
..1n approving the Cycle'5 reload, the Nhu indicates in Reference 8 thatt
"... a-complete new analysis was run for th'e LFWH.
The LFWH was analyzed yith the newly approved MICR0 BURN-B/ANF following the approach previously approved for Cycle 4. using an expanded GG1 dat'a base."
Conclusions The LFWil analyses provided by Grand Gulf for MEOD and each cycle reload allow reactor power to reach a new, higher steady state level and demonstrate acceptable CPR results. This approach is equivalent to considering'the complete failure of the STP (and any other) trip.
-References 1.
J. A. Wooley, "Three-Dimensiona1 BWR Core Simulator," NEDO 20953-A, January _1977.
2.
NEDE-24011-P-A-US, Revision 7, " General Electric Standard Application for Reactor Fuel" (GESTAR) August 1985 (GE-Proprietary).
3.
Entergy' Letter, AECM-86/0174 dated June 9,1986; " Addendum to MEOD Submittal'."
4.
NRC Letter, Lester L. Kintner.to Oliver D. Kingsley dated August 15, 1986.
L5.. XN-NF-900(P), "A Generic Analysis of the Loss of Foodwater lleating Transient for Boiling Water Reactors", February 1986.
- 6..Entergy 2.o_tter, AECH-86/0273 dated September 5,1986; " Cycle 2 Reload
. Submittal Additional Information (LOFWii, LOCA, Fuel Liftoff)."
- 17.. NRC ' Letter, Les ter L. K intner to - Oliver D. ' Kingsley, Jr. - dated October 24,.1986.
8.1 NRC Letter, Lester L. Kintner to William T. Cottle dated November-15, 1990.
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