ML112970853

Proposed Tech Spec Changes Allowing Plant to Remain at Substantial Power Level W/One Recirculation Pump in Operation & Equalizer Valve Closed

ML112970853
Person / Time
Site:	Monticello
Issue date:	07/02/1982
From:	Northern States Power Co
To:
Shared Package
ML112970851	List: ML112970850 ML112970852 ML112970853 ML20063E332
References
NUDOCS 8207130384
Download: ML112970853 (19)

Text

EXHIBIT B Revision '1License Amendment Request Dated - Sept 7, 1976 Exhibit B, attached, consists of the following revised pages of the Appendix A Technical Specifications which incorporate the proposed changes.

Pages 6

7 8

14 15 17 20 56 57 114 114a (new page) 119 211 213 214 215 216 217 113094820702 820 12DR ADOK 05000263 PP DR

I*

2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 1*

2.1 FUEL CLADDING INTEGRITY 2.3 FUEL CLADDING INTEGRITY Applicability: Applicability:

Applies to the interrelated variables Applies to trip settings of the instruments and associated with fuel thermal behavior. devices which are provided to prevent the reactor system safety limits from being exceeded.

Objectives: Objectives:

To establish limits below which the integrity of the fuel cladding is preserved.

To define the level of the process variables at which automatic protective action is S initiated to prevent the safety limits from being exceeded.

Specification: Specification:

A. Core Thermal Power Limit (Reactor The Limiting safety system settings shall be as Pressure > 800 Psia and Core Flow is specified below:

> 10% of Rated)

A. Neutron Flux Scram When the reactor pressure is>800 Psia and core flow is >10% of rated, the 1. APRM - The APRM flux scram trip setting existence of a minimum critical power shall be:

ratio (MCPR) less than 1.07 for two recirculation loop operation or less than S 0.65 (W-dw) + 55%

1.08 for single loop operation for 8x8 and where, 8x8R fuel shall constitute violation of the fuel cladding integrity safety limit.

S = Setting of percent of rated thermal power, rated power 01 being 1670 MWT W = recirculation drive flow in percent dw = single loop operation recirculation reverse flow in the idle loop.

dw = 0 For two recirculation loop operation dw = 5.4 For one recirculation loop operation 6

2.1/2.3 REV

2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS B. Core Thermal Power Limit (Reactor except in the event of operation with a Pressure <800 psia or Core Flow maximum fraction of limiting power density

<10% of rated) for any fuel type in the core greater than the fraction of rated power, when the setting shall be modified as follows:

When the reactor pressure is 800 psia or core flow is<10% of rated, the core S ![0.65 (W-dw) + 55%] FRP MFLPD thermal power shall not exceed 25% of rated thermal power. where, FRP = fraction of rated thermal power, I

C. Power Transients rated power being 1670 MWt MFLPD = maximum fraction of limiting To insure that the safety limit established power density for any fuel type in Specification 2.1.A is not exceeded, each in the core.

required scram shall be initiated by its primary source signal as indicated by 2. IRM - Flux Scram setting shall be 20% of rated the plant process computer neutron flux B. APRM Rod Block - The APRM rod block setting shall be:

S<0.65 (W-dw) + 43% I where, S = Setting of percent of rated thermal power, rated power being 1670 MWT W = recirculation drive flow in percent dw = Single loop operation recirculation reverse flow in the idle loop.

dw = 0 For two recirculation loop operation dw = 5.4 For one recirculation loop operation 2.1/2.3 7 REV

2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS except in the event of operation with a D. Reactor Water Level (Shutdown Condition) maximum fraction of limiting power density for any fuel type in the core greater than Whenever the reactor is in the shutdown the fraction of rated power, when the setting condition with irradiated fuel in the shall be modified as follows:

reactor vessel, the water level shall not FRP be less than that corresponding to 12 S . [0.65 (W-dw) + 43%] MFLPD inches above the top of the active fuel where, when it is seated in the core. This FRP = fraction of rated thermal power, level shall be continuously monitored rated power being 1670 MWt whenever the recirculation pumps are not MFLPD = maximum fraction of limiting operating. power density for any fuel type in the core.

C. Reactor Low Water Level Scram setting shall be >

10'6" above the top of the active fuel.

D. Reactor Low Low Water Level ECCS initiation shall be :6'6" 6'10" above the top of the active fuel.

2.1/2.3 8 REV

Bases:

2.3 The abnormal operational transients applicable to operation of the Monticello Unit have been analyzed throughout the spectrum of planned operating conditions up to the thermal power level of 1670 MWt. The analyses were based upon plant operation in accordance with the operating map given in Figure 3-2-3 of the FSAR. The licensed maximum power level 1670 MWt represents the maximum steady-state power which shall not knowingly be exceeded.

Transient analysis performed each reload are given in Reference 1. Models and model conservatisms are also described in this reference. As discussed in Reference 2, the core wide transient analysis for one recirculation pump operation is conservatively bounded by two-loop operation analysis and the flow dependent rod block and scram setpoint equations are adjusted for one-pump operation.

2.3 BASES 14 REV

Bases Continued:

Deviations from as-left settings of setpoints are expected due to inherent instrument error, operator setting error, drift of the setpoint, etc. Allowable deviations are assigned to the limiting safety system settings for this reason. The effect of settings being at their allowable deviation extreme is minimal with respect to that of the conservatisms discussed above. Although the operator will set the setpoints within the trip settings specified, the actual values of the various setpoints can vary from the specified trip setting by the allowable deviation.

A violation of this specification is assumed to occur only when a device is knowlingly set outside of the limiting trip setting or when a sufficient number of devices have been affected by any means such that the automatic function is incapable of preventing a safety limit from being exceeded while in a reactor mode in which the specified function must be operable. Sections 3.1 and 3.2 list the reactor modes in which the functions listed above are required.

The bases for individual trip settings are discussed in the following paragraphs.

A. Neutron Flux Scram The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1670 MWt).

Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.

Therefore, during abnormal operation transients, the thermal power of the fuel will be less than 2.3 BASES 15 REV

Bases Continued:

backed up by the rod worth minimizer. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated power per minute, and the IRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The IRM scram remains active until the mode switch is placed in the run position. This switch occurs when reactor pressure is greater than 850 psig.

The analysis to support operation at various power and flow relationships has considered opera tion with either one or two recirculation pumps. During steady-state operation with one recircula tion pump operating the equalizer line shall be closed. Analysis of transients from this operating condition are less severe than the same transients from the two pump operation.

The operator will set the APRM neutron flux trip setting no greater than that stated in Specifica tion 2.3.A.1. However, the actual setpoint can be as much as 3% greater than that stated in Specification 2.3.A.1 for recirculation driving flows less than 50% of design and 2% greater than that shown for recirculation driving flows greater than 50% of design due to the deviations discussed on page 18.

B. APRM Control Rod Block Trips Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate, and thus to protect against the condition of a MCPR less than the Safety Limit (T.S.2.1.A). This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit 2.3 BASES 17 REV

Bases Continued:

that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of the neutron scram protection over the entire range of applicability of the fuel cladding integrity safety limit.

The operator will set this pressure trip at greater than or equal to 825 psig. However, the actual trip setting can be as much as 10 psi lower due to the deviations discussed on page 18.

References

1. "Generic Reload Fuel Application", NEDE 24011-P-A-1, July 1979

2. "Monticello Nuclear Generating Plant Single-Loop Operation" NEDO 24271, June 1980 2.3 BASES 20 REV

Table 3.2.3 Instrumentation That Initiates Rod Block a 1II Reactor Modes in Which Min. No. of Oper Function Must be Operable Total No. of able or Operating or Operating and Allow Instrument Instrument Channels able Byeass Conditions**- Channels per Per Trip System Required Function Trip Settings Set1 1 Refuel I Startup I Run Trip System 4 (Notes 1.6)

- Conditions*

F

1. SRM

a. Upscale <5x105 cps x X(d) 1 (Note 3) A or B or C

b. Detector X (a) X(a) 1 (Note 3) A or B or C not fully inserted

2. IRM

a. Downscale >3/125 X(b) X(b) 2 (Note 4) A or B or C full scale

b. Upscale < 108/125 x x 2 (Note 4) A or B or C full scale

3. APRM

a. Upscale See Technical Specifications x 1 (Note 7) Dor E (flow ref erenced) 2.3.B.

b. Downscale > 3/125 full scale x 1 (Note 7) D or E 3.2/4.2 I I' ~

56 REV

Table 3.2.3 - continued Instrumentation That Initiates Rod Block Reactor Modes in Which Min. No. of Oper Function Must be Operable Total No. of able or Operating or Operating and Allow- Instrument Instrument Channels able Bypass Conditions** Channels per Per Trip System Required Function Trip Settings Refuel Startup Run Trip System (Notes 1,6) Conditions*

4. RBM

a. Upscale See Technical X(c) 1 1 (Note 5) D or E (flow ref- Specifications erenced) 2.3.B

b. Downscale >3/125 full X(c) 1 1 (Note 5) D or E

5. Scram Discharge Volume Water Level- <18 gal X X 1 1 B and D, or A High Notes:

(1) There shall be two operable or operating trip systems for each function. If the minimum number of operable or operating instrument channels cannot be met for one of the two trip systems, this condition may exist up to seven days provided that during this time the operable system is functionally tested immediately and daily thereafter.

(2) (deleted)

(3) Only one of the four SRM channels may be bypassed.

(4) There must be at least one operable or operating IRM channel monitoring each core quadrant.

(5) One of the two RBMs may be bypassed for maintenance and/or testing for periods not in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30 day period. An RBM channel will be considered inoperable if there are less than half the total number of normal inputs from any LPRM level.

3.2/4.2 57 REV

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS I. Recirculation System I. Recirculation System

1. Except as specified in 3.5.1.2 below, whenever 1. Once per month, when irradiated fuel is in the irradiated fuel is in the reactor, with reactor reactor with reactor coolant temperature greater 0

coolant temperature greater than 212 F and both than 212 0 F and both reactor recirculation reactor recirculation pumps operating, the pumps operating, the recirculation system cross recirculation system cross tie valve interlocks time valve interlocks shall be demonstrated to shall be operable. be operable by verifying that the cross tie valves cannot be opened using the normal control 0 switch.

2. The recirculation system cross tie valve inter 2. When a recirculation system cross tie valve locks may be inoperable if at least one cross interlock is inoperable, the position of at tie valve is maintained fully closed. least one fully closed cross tie valve shall be recorded daily.

3. Reactor operation with one loop recirculation 3. When in one loop operation, the following may continue at up to 50% of rated power if the surveillances will be completed:

following conditions are met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after one pump operation commences. If the a. APRM flux noise will be measured once per conditions cannot be met or two pump operation shift and the recirculation pump speed will cannot be restored by the end of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an be reduced if the flux noise average over orderly reactor shutdown shall be initiated. k hour exceeds 5% peak to peak as measured on the APRM chart recorder.

a. The Minimum Critical Power Ratio (MCPR)

Safety Limit will be increased per T.S. b. The core plate delta P noise will be measured 2.1.A once per shift and the recirculation pump speed will be reduced if the noise exceeds

b. The MCPR Limiting Condition for Operation 1 psi peak to peak.

(LCO) will be changed per T.S. 3.11.C.

c. The Maximum Average Planar Linear Heat Generation (MAPLHGR) will be changed as noted in Table 3.11.1 3.5/4.5 114 REV

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

d. The APRM scram and rod block setpoints and the RBM setpoints shall be reduced as noted in T.S. 2.3.A and T.S. 2.3.B.

e. The suction valve or the main discharge and main discharge bypass valves in the idle loop is closed and electrically isolated until the idle loop is being prepared for return to service.

f. The equalizer line shall be isolated.

3.5/4.5 114a REV

Bases Continued 3.5:

G. Emergency Cooling Availability The purpose of Specification G is to assure that sufficient core cooling equipment is available at all times.

It is during refueling outages that major maintenance is performed and during such time that all core and containment cooling subsystems may be out of service. Specification 3.5.G.3 allows all core and containment cooling subsystems to be inoperable provided no work is being done which has the potential for draining the reactor vessel. Thus events requiring core cooling are precluded.

Specification 3.5.G.4 recognizes that concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chamber for maintenance or for the inspection required by Specification 4.7.A.1. In this situation, a sufficient inventory of water is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal integrity.

H. Deleted I. Recirculation System The capacity of the Emergency Core Coolant System is based on the potential consequences of a double ended recirculation line break. Such a break involves 3.9 sq. ft. when the cross tie valves are closed and 5.3 sq. ft. when the cross tie valves are open. Specification 3.11.A is based on an ECCS evaluation assuming a break area of 3.9 sq. ft.; the limitations of 3.11.A do not apply to the larger break area.

Therefore, at least one cross tie valve must remain closed during power operation to reduce the potential break area.

An analysis of one-pump operation (equalizer valve closed) identifies certain limitations peculiar to that mode of operation. Reference the September 7, 1976 License Amendment Request from NSP to NRR.

Operation with only one pump is not a normal mode; it will generally involve a forced outage of equipment. There may be insufficient time to make adjustments to the RBM and APRM flow referenced rod block and scram prior to commencing one-pump operation. The reduction in power with the reduced core flow will cause the APLHGR to reduce accordingly, naturally moving in the direction of the new limit. Specification 3.5.1.3 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before these new limits are required to be implemented.

3.5 BASES 119 REV

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.11 REACTOR FUEL ASSEMBLIES 4.11 REACTOR FUEL ASSEMBLIES Applicability Applicability The Limiting Conditions for Operation associated The Surveillance Requirements apply to with the fuel rods apply to those parameters the parameters which monitor the fuel which monitor the fuel rod operating conditions. rod operating conditions.

Objective Objective The objective of the Limiting Conditions for Opera The objective of the Surveillance Require tion is to assure the performance of the fuel rods. ments is to specify the type and frequency of surveillance to be applied to the fuel rods.

Specifications Specifications A. Average Planar Linear Heat Generation Rate (APLHGR) A. Average Planar Linear Heat Genera tion Rate (APLHGR)

During power operation, the APLHGR for each type of fuel as a function of average planar The APLHGR for each type of fuel as exposure shall not exceed for two recirculation a function of average planar exposure loop operation the limiting value given in Table shall be determined daily during reactor 3.11.1 based on a straight line interpolation be operation at 25% rated thermal power.

tween data points and for one recirculation loop operation the values in Table 3.11.1 reduced by 0.85 for all fuel types. When core flow is less than 90% of rated core flow, the APLHGR shall not exceed 95% of the limiting value given in Table 3.11.1. When core flow is less than 70% of rated core flow, the APLHGR shall not exceed 90% of the limiting value given in Table 3.11.1. If any time during operation it is deter mined that the limit for APLHGR is being exceeded, action shall be initiated within 15 3.11/4.11 211 REV

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS C. Minimum Critical Power Ratio (MCPR) C. Minimum Critical Power Ratio (MCPR)

1. During power operation the Operating MCPR Limit shall be MCPR shall be determined daily during reactor

> 1.43 for 8x8 and 8x8R fuel, 1.47 for P8x8R fuel at power operation at 25% rated thermal power rated power and flow for two recirculation loop operation,' and following any change in power level or provided 14 T8vE* (see section 3.3.C.3). If at any time distribution which has the potential of during operation it is determined that the limiting value bringing the core to its operating MCPR Limit.

for MCPR is being exceeded, action shall be initiated with in 15 minutes to restore operation to within the prescribed limits. Surveillance and corresponding action shall con tinue until reactor operation is within the prescribed limits. If the steady state MCPR is not returned to with in the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown conditions 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 Limit shall be the above applicable MCPR value time K where K is as shown in Figure 3.11.3.

For one recirculation loop operation the MCPR limits at rated flow are 0.01 higher than the comparable two-loop values.

2. If the gross radioactivity release rate of noble gases at the steam jet air ejector monitors exceeds, for a period greater than 15 minutes, the equivalent of 236,000 uCi/sec following a 30-minute decay, the Operating MCPR Limits specified in 3.11.C.1 shall be adjusted to 11.48 for all fuel types, times the appropriate Kf.

Subsequent operation with the adjusted MCPR values shall be per paragraph 3.11.C.1.

For one recirculation loop operation the MCPR limits at rated flow are 0.01 higher than the comparable two-loop values.

• if >-to , the operating MCPR Limit shall be a linear interpolation between the limits in 3.11.C.1 and 1.48 for 8x8 and 8x8R fuel and 1.52 for P8x8R fuel.

3.11/4.11 213 REV

TABLE 3.11.1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE vs. EXPOSURE Exposure MAPLHGR FOR EACH FUEL TYPE (kw/ft) (Note 1) I MWD/STU 8DB262 8DB250 8DB219L 8DRB265L P8DRB265L 8DRB282 P8DRB282 P8DRB284LB 200 11.1 11.2 11.4 11.5 11.6 11.2 11.2 11.4 1,000 11.3 11.3 11.5 11.6 11.6 11.2 11.2 11.4 5,000 11.9 11.9 11.9 11.7 11.8 11 6 11.8 11.8 10,000 12.1 12.1 12.0 11.8 11.9 11.7 11.9 11.9 15,000 12.1 12.1 11.9 11.7 11.9 11.7 11.8 11.9 20,000 12.0 11.9 11.8 11.6 11.8 11.5 11.7 11.7 25,000 11.6 11.5 11.3 11.3 11.3 11.3 11.3 11.4 30,000 10.3 10.6 10.2 10.3 10.5 10.4 10.7 10.6 35,000 9.3 9.3 9.3 9.2 9.5 9.2 9.5 9.5 (36,000) 9.1 9.0 9.1 9.0 9.3 9.0 9.3 9.3 40,000 8.9*

45,000 8.0* i (1) For two recirculation loop operation. For one recirculation loop operation multiply these values by 0.85 I 50,000 7.3*

• For extended burnup program test bundles 214 3.11/4.11 REV

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 limit specified in the 10CFR50, Appendix K.

The peak cladding temperature following a postulated loss-of-coolant accident is primarily 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 cladding temperature by less than + 200 relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10CFR50 Appendix K limit. The limiting value for APLHGR is given by this specification.

Reference 6 demonstrates that for lower initial core flow rates the potential exists for earlier DNB during postulated LOCA's. Therefore a more restrictive limit for APLHGR is required during reduced flow conditions.

Those abnormal operational transients, analyzed in FSAR Section 14.5, which result in an automatic reactor scram are not considered a violation of the LCO. Exceeding APLHGR limits in such cases need not be reported.

Reduction factors for one recirculation loop operation were derived in Reference 8.

B. LHGR This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification 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 95% confidence, that no more than one fuel rod exceeds the design linear heat generation rate due to power spiking.

Those abnormal operational transients, analyzed in FSAR Section 14.5, which result in an automatic reactor scram are not considered a violation of the LCO. Exceeding LHGR limits in such cases need not be reported.

215 3.11 BASES REV

Bases Continued C. Minimum Critical Power Ratio (MCPR)

The ECCS evaluation presented in Reference 4 and Reference 6 assumed the steady state MCPR prior to the postulated loss-of-coolant accident to be 1.24 for all fuel types for normal and reduced flow. The Operating MCPR Limit for two recirculation loop operation is determined from the analysis of transients discussed in Bases Sections 2.1 and 2.3. By maintaining an operating MCPR above these limits, the Safety Limit (T.S. 2.1.A) is maintained in the event of the most limiting abnormal operational transient.

For one recirculation loop operation the MCPR limits at rated flows are 0.01 higher than the comparable two-loop values.

Use of GE's new ODYN code Option B will require average scram time to be a factor in determining the MCPR (Reference 7). In order to increase the operating envelope for MCPR below MCPRA (ODYN code Option A), the cycle average scram time ('Zv&) must be determined (see Bases 3.3.C). If 1Lgyr is below the adjusted analysis scram time, the MCPR Limit can be used. If2 ,&'T1 a linear interpolation must be used to determine the appropriate MCPR. For example:

MCPR = MCPR + ' (MCPR -MCPR B 0.9- , A CPRB MCPRA and MCPRB have been determined from the most limiting accident analyses.

For operation with less than rated core flow the Operating MCPR Limit is adjusted by multiplying the above limit by K . Reference 5 discusses how the transient analysis done at rated conditions encompasses the reduced flow situation when the proper K factor is applied.

Noble gas activity levels above that stated in 3.11.C.2 are indicative of fuel failure. Since the failure mode cannot be positively identified, a more conservative Operating MCPR Limit must be applied to account for a possible fuel loading error.

Those abnormal operational transients, analyzed in FSAR Section 14.5, which result in an automatic reactor scram are not considered a violation of the LCO. Exceeding MCPR limits in such cases need not be reported.

3.11 BASES 216 REV

Bases Continued References

1. "Fuel Densification Effects in General Electric Boiling Water Reactor Fuel," Supplements 6,7, and 8, NEDM-10735, August, 1973.

2. Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 14, 1974 (USAEC Regulatory Staff).

3. Communication: VA Moore to IS Mitchell, "Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974.

4. "Loss-of-Coolant Accident Analysis Report for the Monticello Nuclear Generating Plant," NEDO 24050 -1, December, 1980, L 0 Mayer (NSP) to Director of Nuclear Reactor Regulation (USNRC),

February 6, 1981.

5. "General Electric BWR Generic Reload Application for 8x8 Fuel," NEDO-20360, Revision 1, November 1974.

6. "Revision of Low Core Flow Effects on LOCA Analysis for Operating BWR's," R L Gridley (GE) to D G Eisenhut (USNRC), September 28, 1977.

7. "Response to NRC Request for Information on ODYN Computer Mode," R H Buchholz (GE) to P S Check (USNRC), September 5, 1980.

8. "Monticello N.G.P. Single-loop Operation NEDO 24271, June 1980" Bases 4.11 The APLHGR, LHGR and MCPR shall be checked daily to determine if fuel burnup, or control rod movement have caused changes in power distribution. Since changes due 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 25% of rated thermal power, an unreasonably large peaking factor would be required, which is not the case for operating control rod sequences. In addition, the MCPR 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 limits.

4.11 BASES 217 REV