ML20024G377
| ML20024G377 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/23/1976 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20024G375 | List: |
| References | |
| NUDOCS 9102110395 | |
| Download: ML20024G377 (11) | |
Text
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I "U M U-20 0 3.0 LD1ITING CONDITIONS FOR OPERATIOtI
>- 4.0 SI?9Vril.tANCE REQUIRD1ENTS 00 _
OW N 3.11 REACTOR RJEL ASSEMBLIES 4.11 REACTOR FUEL ASSE2iBLIES 09 Ue Applicability 00 Applicability M8*
2 0-T*ne Limiting Canditinns for Operation %e Surveillance Requirements apply to u associated with the fuel rods apply to the parameters which monitor the fuel those parameters which monitor the fuel rod ,pera ting conditions, rod operating conditions.
OblectIve O_tjectEve The objective of the Limiting Condit ions %e uhjective of the Surveillance Requirements for Operation is to assure the perfor- is to speci fy the type end frequency of surveil-mance of the fuel rods. lance to be applied to the fuel rods.
Specifications Spec i fi ca t ion s A. Average Plana r Linea r lleat Genera- A. Ave rage Planar Linear IIcat Genera-tion Rate (APIECPQ t ion Ha t e ( APUICR)
During steady state :uwer operation, l We APLIICR for each type of fuel as a the APUlGR for each type of fuel as a f u >ctioa of average planar exposure shall function of average planar exposure be d< termined daily during reactor operation shall not exceed the limiting value , at 2. n7. rated thermal power.
shown in Figures 3.11.1. If at any time during operation it is determined that the limiting value for APUIGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. Surveillance and correspond-ing action shall continue until reactor operation is within the prescribed limits. If the APUlGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 189B 3.11/4.11
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIRDf ENTS B. Linear Heat Generation Rate (UIGR) B. Linear Heat Generation Rate (UIGR)
During steady state power operation, The UIGR as a function of core height the IRCR as a function of core height shall be checked daily during reactor shall not exceed the limiting value operation at 2257. of rated thermal shown in Figure 3.11.2 If at any power.
time during operation it is determined ,
that the limiting value for UIGR is I being exceeded, action shall be ini-tiated within 15 minutes to restore operation to within the prescribed limits. Surveillance and corres-ponding action shall continue until reactor operation is within the prescribed limits. If the UIGR is not returned to within the prescribed limits within two (2) hours, the 4
reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
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189C 3.11/4.11
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE RQUIRDtENTS C. Minimum Critical Power Ratio (MCPR) C. Minimum Critical Power Ratio (MCPR)
During steady state power operation, MCPR shall be determined dailf dur-the Operating MCPR Limit shall be ing reactor power operation at 2257.
l 31.38 for 8x8 fuel and 31.29 for rated thermal pow c and following -
7x7 fuel at rated power and flow. any change in power level or distri-If at any time during operation it bution which has the potential of is determined that the limiting bringing the core to its MCPR limit.
value 6 m is being exceeded, action shall be initiated within 15 l mirr.stes to restore operation to with-in the prescribed limits. Surveil-i lar.ce and corresponding action shall
, contim: r.".51 reactor operation is within the preceribed limits. If the steady state MCPR is not returned to within the pre scribed limits with-
) in two (2) hours, the reactor shall be brought to the Cold Shutdown con-dition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. For core flows other than rated the Operating MCPR Linit shall be the above value times Kg where Kg 1.s as shown in Figure 3.11.3.
,! 189D -
3.11/4.11 i
Bases 3.11 A. Average Planar Linear Heat Generation Rate (APLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the Ibmit specified in the 10CFE50, Appendix K.
The peak cladding temperature following a postulated loss-of-coolant accident is prLmarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly. Since expected local variations in power distribution within a fuel assembly affect the calculated peak ciadding temperature by less than + 20 F relative to the peak temperature for a typical fuel design, the ILnit on the average linear heat generation rate is sufficient l
to assure that calculated temperatures are within the 10CFR50 Appendix K ILait. The ihniting value for APLHGR is given by this specification.
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B. Local LHCR This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densf.fication is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of Reference 1 and in References 2 and 3, and assumes a linearly increasing variation and axial gaps between core bottom and top and assures with a 05% confidence, that no more than one fuel rod exceeds the design 10near heat generation rate due to power spiking.
189 E 3.11 EASES
-Bases 3.11'(continued) ' ^
C. Minimuss' Critical Power Ratio (MCPR)
~ The' ECCS evaluation presented in Reference 4 asstamed the steady _ st. ate MCFR prior to the postulated loss of coolant accident to be 1.18 for all fuel types. The Operating MCIR Limit l of 1.38 for 8x8 fuel and 1.29. for 7x7 fuel is determined from the analysis of transients discussed in Bases Sections 2.1 and 2.3. By anaintaining an operating MCPR above these lini ,
the Safety Limit of 1.06 (T.S.2.1.A) applicable to all fuel types is maintained in the event of the most limiting abnormal operational transient.
For operation with less than rated core flow the Operating MCPR Limit is adjusted by nultiplying the above limit by K f. Reference 5 discusses how the transient analysis done at re,ted conditions encompasses the reduced flow situation when the proper Kg factor is applied.
Re ferences
- 1. "Puel Densification Effects in General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEDM-10733, August 1973.
- 2. Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 14,1974 (USAEC Regulatory Staff)
- 3. Conununication: V A Moore to I S Mitchell, " Modified GE Model for Fuel Densification,"
Docket 50-321, March 27,19M.
- 4. "Monticello Nuclear Generating Plant Loss-Of-Coolant Accident Analysis Conformance with 10 CFR 50 Appendix K, August 1974," L 0 Mayer (NSF) to J F O' Leary, August 20, 1974.
- 5. " General Electric BWR Ceneric Reload Application for 8 x 8 Fuel," NEDO-20360, Revision 1, November, 1974
'3.11 BASES 189F
l Bases 4.11 The APUIGR, UIGR and FEPR shall be checked daily to determine if fuel burnup, or control rod movement have caused changes in power distribution. Since changes dt.e to burnup are slow, and only a few control rods are removed daily, a daily check of power distribution is adequate. For a limiting value to occur below 257. of rated themal power, an unreasonably large peaking factor would be required, which is not the case for operating control rod sequences. In addition, the M R.is ,
checked whenever changes in the core power level or distribution are made which have the potential of bringing the fuel rods to their thermal-hydraulic lictits.
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LIST OF FIGURES Figure No. Page No. 2.1-1 Deleted 2.3.1 APRM Flow Referenced Scram and Rod Block-Trip Settings 2.3.2 Relationship Between Peak Ileat Flux and Power for Peaking Factor of 3.08 12 4.1.1 'M' Factor - Graphical Aid in the Selection of an Adequate Interval Between Tests 46 4.2.1 System finavailability 74 3.4.1 Sodium Pentaborate Solution Volume - Concentration Requirements 92 3.4.2 Sodium Pentaborate Solution Temperature Requirements 93 3.6.1 Change in Charpy V Transition Temperature versus Neut ron Exposure 122 3.6.2 Minimum Temperature versus Pressure for Pressure Tests 122A 3.6.3 Minimum Temperature versus Pressure for Mechanical ficatup or Cooldown Following Nuclear Shutdown 122B 3.6.4 Minimum Temperature versus Pressure for Core Operation 122C 4.6.1 Deleted 4.6.2 Chloride Stress Corrosion Test Results @ 500*F 123 4.8.1 Off-gas Storage Tank Gross Activity Limits 176A 3.ll.1-A Maximum Average Linear lleat Generation Rate versus Planar Average Exposure l Monticello 8D219 Fuel 18911 3.II.1-B Maximum Average Linear lleat Generation Rate versus Planar Average Exposure Monticello 7D230 Fuel 189I vii
- LIST OF FIGURES L '
Figure No. Page No. 3.ll.1-C Maximum Aurage Linear lleat Generation Rate versus Planar Average Exposure Monticello 8D262 Fuel 189J 3.11.1-D Maximum Average Linear ifeat Generation Rate versus Planar Average Exposure t Monticello 8D250 Fuel 189K 3.11.2 IliGR Versus Core Height 189L i 3.11.3 K g Factor versus Percent of Rated Core Flow IggM i 6.1.1 NSP Corporate Organizational Relationship to On-Site Operating Organization 193 6.1.2 l Functional Organization for On-Site Operating Group 194 i 1 a I viii l}}