ML17199U411

Proposed Tech Specs,Supporting Refueling & Startup for Cycle 11 Operation in June 1988

ML17199U411
Person / Time
Site:	Dresden
Issue date:	03/09/1988
From:	COMMONWEALTH EDISON CO.
To:
References
NUDOCS 8803150322
Download: ML17199U411 (78)

Text

• - ~~*

ENCLOSURE E TECHNICAL SPECIFICATION/LICENSE CHANGES FOR D3Cll 4321K 060~1~c*)*A~~ onl*,~c>6

( I VV

~ ~ ~LL 00 ~ '

I PDR ADOCK.-05000249 P_ _

_ _ _: __ ~_ _ L\\C:_D_ -*

.* DPR-25 Am. 31

2.

1/30/78 E.

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Dresden Nuclear Power Station, Units Nos. l, 2 and

3.

Am.

Am.

1997b

3.

This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations; 10 CFR Part 20, Section 30.34 of 10 CFR Part

• 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A.

Maximum Power Level Commonwealth Edison is authorized to operate the facility at steady state power levels not in excess of 2527 megawatts (thermal), except that Commonwealth Edison shall not operate the facility at power levels in excess of five (5) megawatts (thermal), until satisfactory completion of modifications and final*

testing of the station output transformer, the auto-depressurization interlock, and the feedwater system, as described in Commonwealth Edison's telegrams dated February 2G, 1971, have been verified

.in writing by the Commission.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications, C.

Reports Commonwealth Edison shall make certain reports in accordance with the requirements of the Technical Specifications.

D.

Records E.

Commonwealth Edison shall keep facility operating records in accordance with the requirements of the Technical Specifications.

Restrictions Operation in the coastdown mode is permitted to 40%

power.

I*)"*

I.,

Figure 2.1-3 Figure 4. 1. 1 Figure 4.2.2 Figure 3.4.1 Figure 3. 4.,2 Figure 3.5-1 Figure 3.5-1 Figure 3. 5-lA Figure 3.5-lB Figure 3.5-2 Figure 3.5-2 Figure 3.5-2 Figure 3.6.1 Figure 3.6.2 Figure 4.6.1 Figure 4.6.2 Figure 4.8-1 Figure 4.8-2 Figure 6. 1-*1 1997b DRESDEN III DPR-25 Amendment No. ~, 119, fl List of Figures APRM Bias Scram Relationship to Normal Operating Conditions Graphi~al Aid in the Selection of an Adequate Interval Between Tests Test Interval vs. System Unavailability Standby Liquid Control Solution Requirements Sodium Pentaborate Solution Temperature Requirements MAPLHGR Limit vs Bundle Average Exposure ANF 8x8 Fuel (Sheet 1 of 2)

MAPLHGR Limit vs Sund le Average Exposure ANF 9x9 Fuel (Sheet 2 of 2)

Steady State Linear Heat Generation Rate vs Nodal Exposure Transient Linear Heat Generation Rate vs.

Nodal Exposure MCPR Limit for Reduced Total Core Flow (Sheet 1 of 3) 8x8 MCPR Operating Limit for Automatic Flow Control (Sheet 2 of 3) 9x9 MCPR Operating Limit for Automatic Flow Control (Sheet 3 of 3)

Minimum Temperature Requirements per Appendix G of 10 CFR 50 Thermal Power vs Core Flow Limits for Thermal Hydraulic Stability Surveillance In Single Loop Operation Minimum Reactor Pressuri~ation Temperat~re Chloride Stress Corrosion Test Results at 500°F owner Control led/Unrestricted Area Boundary Detail of Central Complex Corporate and Station Organization viii B 112:1-17 B 3/4.1-18 8 3/4.2-38 3/4.4-4 3/4.4-5 3/4.5-18 3/4.5-19 3/4.5-20 3/4.5-21 3/4.5-25 3/4.5-26 3/4.5:-27 3/4.6-23 3/4.6-24 B 3/4.6-29 B 3/4.6-31 B 3/4.8-38 B 3/4.8-39 6-3

DRESDEN III DPR-25 Amendment No. pt;,, ~

1.0 DEFINITIONS (Cont'd.)

1997b plant can be operated safely and abnormal situations can be safely controlled.

J.

Limiting Safety.System Setting (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded.

The region between the safety limit and these settings represents margin with normal operation lying below these settings.

The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.

K.

Steady State Linear Heat Generation Rate (SLHGR) - The steady I state linear heat generation rate limit protects against exceeding the fuel end-of-life steady state design criteria developed by Advanced Nuclear Fuels.

L.

Logic System Function Test - A logic system functional test means a test of all relays and co.ntacts of a logic circuit from.

sensor to activated device to insure all components are operable per design intent. Where possible, action will go to completion, i.e., pumps will be started and valves opened.

M.

Minimum Critical Power Ratio (MCPR) - The minimum in-core critical power ratio corresponding to the most limiting fuel assembly in the core.

N."

Mode - The reactor mode is that which is established by the mode-selector-switch.

0.

Operable - A system, subsystem, train, component, or device shall be operable when it is capable of performing its specified function{s).

Imp}icit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of

. performing.their related support function(s).

P. -Operating.:.:-operating -means that a system, subsystem, train; component or device is performing its intended functions in its required manner.

Q.

Operating Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage.

1.0-2

DRESDEN III DPR-25 Amendment No. 'f?, '7 1.0 DEFINITIONS (Cont'd.)

1997b AA.

Shutdown - The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alternations are being performed.

When the mode switch is placed in the shutdown position a reactor scram is initiated, power to the control rod drives is removed, and the reactor protection system trip systems are de-energized.

1.

Hot Shutdown means conditions as above with reactor coolant temperature greater than 212"F.

2.

Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 212"F.

BB.

Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.

CC.

Surveillance Inte~val - Each surveillance requirement shall be performed within the specified surveillance interval with:

a.

A maximum allowable extension not to exceed 25% of the surveillance interval.

b.

A total maximum combined interval time for any 3 consecutive intervals not to exceed*3.25 times the specified surveillance interval.

DD.

Fraction of Rated Power (FRP) - The fraction of rated power is the ratio of core thermal power to rated thermal power of 2527 Mwth.

EE.

Transition Boiling - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

FF.

Fuel Design Limiting Ratio for Exxon Fuel (FDLRX) - The fuel design limiting ratio for Exxon fuel is the limit used to*

assure that the fuel operates within the end-of-life steady

.. -- -state design criteria.

• FDLRX-assures acceptable end-of..:. life -

conditions by, among other items, limiting the release of fission gas to the cladding plenum.

GG.

Dose Equivalent I-131 - That concentration of I-131 (microcurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites".

1.0-5

DRESDEN III DPR-25 Amendment No. pl, JrJ 1.0 DEFINITIONS (Cont'd.)

1997b HH.

Process Control Program (PCP) - Contains the sampling, analysis, and formulation determination by which solidification of radioactive wastes from liquid systems is assured.

II. Offsite Dose Calculation Manual (ODCM) - Contains the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, and in the calculation of gaseous and liquid effluent monitor alarm/trip setpoints.

JJ.

Channel Functional Test (Radiation Monitor) - Shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify operability including alarm and/or trip functions.

KK.

Source Check - The qualitative assessment of instrument response when the sensor is exposed to a radioactive source.

LL.

Member( s) of the Public - Shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors, or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational, or other purposes not

~ssociated with the plant.

MM.

Rated Recirculation Pump Speed - is the recirculation pump speed that corresponds to rated core flow (98 x 106 lb/hr) when operating at rated thermal power (dual loop operation).

NN.

Dual Loop Operation - reactor power operation with both recirculation pumps running.

00.

Single Loop Operation (SLO) - reactor power operation with one recirculation pump running.

PP.

Transient Linear Heat Generation Rate (TLHGR) - The transient linear heat generation rate limit protects against fuel centerline melting and 1% plastic cladding strain during transient conditions throughout the life of the fuel.

ct:).

Fuel Design Limiting Ratio for Centerline Melt (FDLRC) - The fuel design limiting ratio for centerline melt is the limit used to assure that the fuel will neither experience centerline melt nor exceed 1% plastic cladding strain for transient overpower events beginning at any power and terminating at 120%

of ra_ted thermal power.

RR.

Linear Heat Generation Rate (LHGR) - The linear heat generation rate is the operating fuel pin power level.

1.0-6

I-1.1 SAFETY LIMIT FUEL CLADDING INTEGRITY Applicability:

The Safety Limits established to preserve the fuel cladding integrity apply to these variables which monitor the fuel thermal behavior.

Objective:

The objective of the Safety Limits is to establish limits below which the integrity of the fuel cladding is pre-served.

Specifications:

A.

Reactor Pressure greater than 800 psig and Core Flow greater than 10% of Rated.

The existence of a minimum critical power ratio (MCPR) less than 1.05 shall eonstitute a violation of the MCPR fuel cladding integrity safety limit.

When in Single loop Operation, the MCPR safety limit shall be increased by 0.01.

DRESDEN III DPR-25 Amendment No. 1'5, p!I 2.1 LIMITING SAFETY SYSTEM SETTING FUEL CLADDING INTEGRITY Applicability:

The Limiting Safety System Set-tings apply to trip settings of the instruments and devices which are provided to prevent the fuel cladding integrity Safety Limits from being exceeded.

Objective:

The objective of the Limiting Safety System Settings is to define the level of the process variables at which automatic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.

Specifications:

A.

Neutron Flux Trip Settings The limiting safety system trip settings shall be as specified below:

1.

APRM Flux Scram Trip Setting (Run Mode)

When the reactor mode switch is in the run position, the APRM flux scram setting shall be:

S less than or equal to

[.58WD + 62) during Dual Loop Operation or S less than or equal to [.58Wo +

58.5) during Single Loop Operation with a maximum setpoint of.

120% for core flow equal to 98 x 106 lb/hr and greater, where:

S - setting in percent of rated thermal power.

1/2.1-1 1997b

1.1 SAFETY LIMIT (Cont'd.)

1997b DRESDEN III DPR-25 Amendment No. ~, ¢ 2.1 LIMITING SAFETY SYSTEM SETTING (Cont'd.)

1/2.1-2 w0 = percent of drive flow required to produce a rated core flow of 98 Mlb/hr.

In the event of operation of any fuel assembly with a fuel design limiting ratio for centerline melt (FDLRC) greater than 1.0, the setting shall be modified as follows:

Where: S is less than or equal to

(.58Wo + 62)/FDLRC during Dual Loop Operation or (.58 w0 + 58.5)/FDLRC during I

Single Loop Operation

-I The value of FDLRC shal 1 be t set equal to 1.0 unless the actual operating value is greater than 1.0, in which case the actual operating value will be used.

This adjustment may also be performed by increasing the APRM gain by FDLRC, which f.

accomplishes the same degree of protection as reducing the trip setting by 1/FDLRC.

f -

2, APRM Flux Scram TriQ Setting (Refuel or Startup and Hot Standby Mode)

1.1 SAFETY LIMIT (Cont'd.)

1997b B.

Core Thermal Power Limit (Reactor Pressure is less than or equal to 800 psig)

When the reactor pressure is less than or equal to 800 psig or core flow is less than 10% of rated, the core thermal power shall not exceed 25 percent of rated thermal power.

DRESDEN III DPR-25 Amendment No. "¢; ¢ 2.1 LIMITING SAFETY SYSTEM SETTING (Cont'd.)

1/2.1-3 When the reactor mode switch is in the refuel or the startup/hot standby position, the APRM scram shall be set at less than or equal to 15% of rated core thermal power.

3.

IRM Flux Scram Trip Setting The IRM flux scram setting shall be set at less than or equal to 120/125 of full scale B.

APRM Rod Block Setting The APRM rod block setting shall be:

S is less than or equal to

[.58WD + 50] during Dual Loop Operation or S is less than or equal to [.58 WD

+ 46.5] during Single Loop Operation.

The definitions used above for the APRM scram trip apply.

In the event of operation of any fuel a5sembly with a fuel design limiting ratio for centerline melt (FDLRC) greater than 1.0, the setting shall be modified as follows:

S is less th~n~.or equa! to

(.58WD + 50)/FDLRC during Dual Loop Operation or S is less than or equal to c.sawD + 46.5)/FDLRC I

during Single Loop Operation The definitions used above for the APRM scram trip apply.

The value of FDLRC shall be set equal to 1.0 unless the actual operating value is greater than 1.0. In which case the actual operating value will be used.

1.1 SAFETY LIMIT (Cont'd.)

c.

D.

Power Transient

1.

The neutron flux shall not exceed the scram setting established in Specification 2.1.A for longer than 1.5 seconds as indicated by the proces~ computer.

2.

When the process computer is out of service, this

.safety limit shall be assumed to be exceeded if the neutron flux exceeds the scram setting established by Specification 2.1.A and a control rod scram does not occur.

Reactor Water Level (Shutdown Condition)

Whenever.the reactor is in the shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be less than that corresponding to 12 inches above the top of the active

-fUerwhen-it is--seated in the--

core.

Note:

Top of active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).

DRESDEN III DPR-25 Amendment No. ~, 0, qS 2.1 LIMITING SAFETY SYSTEM SETTING (Cont'd.)

The adjustment may also be performed by increasing the APRM gain by FDLRC, which accomplishes the same degree of protection as reducing the

. trip setting by 1/FDLRC.

I C.

Reactor low water level scram setting shall be greater than or equal to 144" above the top of the active fuel at normal operating conditions.

D.

Note:

Top of active fuel is defined to be 3GO inches above vessel zero (see Bases 3.2).

Reactor low water level ECCS initiation shall be 84" (plus 4",minus 0") above the top of the active*fuel at normal operating conditions.

Note:

Top of active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).

1/2.1-4 1997b

DRESDEN III DPR-25 Amendment No. j/:i, p 1.1 SAFETY LIMIT BASES (Cont'd.)

power ratio (CPR) which is the ratio of the bundle power which would produce the onset of transition boiling divided by the actual bundle power.

The minimum value of this ratio for any bundle in the core is the Minimum Critical Power Ratio (MCPR).

It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables.

(Figure 2.1-3).

The.MCPR Fuel Cladding Integrity Safety Limit assures sufficient conservatism in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at least 99.9% of the fuel rod.s in the core would be expected to avoid boiling transition.

The margin between calculated boiling transition (MCPR=l.00) and the MCPR Fuel Cladding Integrity Safety Limit is based on a detailed statistical procedure which considers the uncertainties in monitoring the core operating state.

One specific uncertainty included in. the safety limit is the uncertainty inherent in the XN-3 critical power correlation.

Refer to XN-NF-524 for the methodology used in determining the MCPR Fuel Cladding Integrity Safety Limit.

The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of

• assurance that the critical power as evaluated by the correlation is within a small percentage of the actual critical power being estimated.

The assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bounclingly high radial power peaking factors and boundingly flat local peaking distributions are used to estimate the number of rods in boiling transition.

• Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number

• of rods in boiling transition. These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that during sustained operation at the MCPR F.uel Cladding Integrity Safety Limit there would be no transition boiling in the core. If boiling transition were to occur, however, there is reason to believe that the integrity of the fuel would not necessarily be compromised.

Significant test data accumulated by the U.S. Nuclear Regulatory Commission and private organizations indicate that the use of

~*... -* -... ---a boiling-fraris-i tiol'l-limitatiori to -protect against claoaing-1997b

• failure is a very conservative approach; much of the data indicates that LWR fuel can survive for an extended period in an environment of transition boiling.

During Single Loop Operation, the MCPR safety limit is

• increased by 0.01 to conservatively account for increased uncertainties in the core flow and TIP measurements.

B 1/2.1-7

DRESDEN III DPR-25 Amendment No. ~, ~, ~

1.1 SAFETY LIMIT BASES (Cont'd.)

1997b If the reactor pressure should ever exceed the limit of applicability of the XN-3 critical power correlation as defined in XN-NF-512, it would be assumed that the MCPR Fuel Cladding Integrity Safety Limit had been violated.

This applicability pressure limit is higher than the pressure safety limit specified in Specification 1.2.

For fuel fabricated by Advanced Nuclear Fuels Corporation (ANF), fuel design criteria have been established to provide protection against fuel centerline melting and 1% plastic cladding strain during transient overpower conditions throughout the life of the fuel.

To demonstrate compliance with these criteria, fuel rod centerline temperatures are

. determined at 120% overpower cone! i tions as a check against calculated centerline melt temperatures.

FDLRC is incorporated to protect the above criteria at all power levels considering.events which wi 11 cause the reactor power to increase to 120% of rated thermal power.

B.

Core Thermal Power Limit (Reactor Pressure less than 800 ps~

At pressures below 800 psia, the core elevation pressure drop (O power, O flow) is greater than 4.56 psi.

At low powers and flows this pressure differential is maint~ined in the bypass region of the core.

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and flows will always be greater than 4.56 psi.

Analyses show that with a flow of 28x103 lbs/hr.

bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow B 1/2.1-8

-I

2.1 1997b DRESDEN III DPR-25 Amendment No. ~, 9"'

LIMITING SAFETY SYSTEM SETTING BASES (Cont'd.)

2.

At times it may be necessary to operate wfth one reactor coolant recirculation pump out of service.

During Single Loop Operation, the normal drive flow relationship during Dual Loop Operation is altered.

This is the result of reverse flow through the idle loop jet pumps when the active loop recirculation pump speed is above 20 to 40% of rated.

Some of the active loop flow is then diverted from the core and backflows through the idle loop jet pumps; hence, the core receives less flow than would be predicted based upon the Dual Loop drive flow to core flow relation-ship. If the APRM flow biased trip settings were not altered for Single Loop Operation, the new drive flow to core flow relationship would nonconservatively result in flow biased trips occurring at neutron fluxes higher than normal for a given core flow.

The scram trip setting must be adjusted to ensure that the LHGR transient limit is.not violated for any power distribution.

This is expressed as the fuel design limiting ratio for centerline melt (FDLRC).

The scram setting is adjusted in accordance with the formula in specification 2.1.A.l when FDLRC is greater than 1.0.

The adjustment may also be accomplished by increasing the APRM gain by FDLRC.

This provides the same degree of protection-as reducing the trip setting by 1/FDLRC by raising the initial APRM reading closer to the trip setting such that a scram would be received at the same point in a transient as if the trip setting had been reduced.

APRM Flux Scram Trip Setting (Refuel or Start & Hot Standby Mode)

For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated.

The margin is adequate to accommodate anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not

~much colder than that already *in the system;* temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer.

Of all possible sources of reactivity input, *uniform control rod withdrawal is the most probable cause of significant power rise.

Because B 1/2.1-12

2.1 1997b DRESDEN III DPR-25 Amendment No. ~, Jn LIMITING SAFETY SYSTEM SETTING BASES (Cont'd.)

B.

APRM Rod Block Trip Setting

c.

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 gross rod withdrawal at constant recirculation flow rate to protect against grossly exceeding the MCPR fuel cladding integrity safety limit.

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 increases as the flow decreases for the specified trip setting versus flow relationship; therefore the worse case MCPR which could occur during steady-state operation is at 108%

of rated thermal power during dual loop operation or 104.5 percent during single loop operation because of the APRM rod block trip setting.

As with the APRM flow biased scram, the reduced setpoint during single loop operation accounts for possible reverse flow in the idle loop jet pumps.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward or APRM gain increased if the fuel design limiting ratio for centerline melt (FDLRC) for any fuel assembly exceeds 1.0, thus preserving the APRM rod block safety margin.

Reactor Low Water Level Scram - The reactor low water level scram is set at a p~int which will assure that the water level used in the bases for the safety limit is maintained.

The scram setpoint is based on normal operating temperature and pressure conditions because the level instrumentation is density compensated.

D.

Reactor Low Low Water Level ECCS Initiation Trip Point - The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident and to limit fuel clad temperature to well below the clad melting tempera"ture to

_assure_ that core geometry remains intact and to limit any clad metal-water reaction-to less than -1%. - Tei-accomplish their-intended function, the capacity of each emergency core cooling system component was established based on the reactor low water level scram setpoint.

To lower the setpoint of the low water level scram would require an increase in the capacity requirement for each of the ECCS components.

Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet was not set lower because of ECCS.

capacity requirements.

B 1/2.1-14

3.1 LIMITING CONDITIONS FOR OPERATION 1997b REACTOR PROTECTION SYSTEM Applicability:

Applies to the instrumentation and associated devices which initiates a reactor scram.

Objective:

To assure the operability of the reactor protection system.

Specification:

A.

Reactor Protection System

1.

The setpoints, minimum number of trip systems, and minimum number of instrument channels that must be operable for each position of the reactor mode switch shall be as given in Table 3.1.1.

The syste~ response times from the opening of the sensor contact up to and including the opening of the trip actuator contacts shall not exceed 50 milliseconds.

2.

If during operation, the fuel design limiting ratio for centerline melt (FDLRC) for any fuel assembly exceeds 1.0 when operating above 25% rated thermal power, either:

3/4.1-1 DRESDEN III DPR-25 Amendment No. }1., ~, ~

4.1 SURVEILLANCE REQUIREMENTS REACTOR PROTECTION SYSTEM Applicability:

Applies to the surveillance of the instrumentation and associated devices which initiate reactor scram.

Objective:

To specify the type and frequency of surveillance to be applied to the protection instrumentation.

Specification:

A.

Reactor Protection System

1.

Instrumentation systems shall be functionally tested and calibrated as indicated in Tables 4

• l. l and 4

• l. 2 I respectively.

2.

Daily during reactor power operation above 25% rated thermal power, the core power distribution shall be checked for:

-~

3.1 LIMITING CONDITIONS FOR OPERATION (Cont'd.)

1997b

a.

The APRM scram and rod block settings shall be reduced to the values given by the equations in Specifications 2.1.A.l and 2.1.B.

This may be accomplished by increasing APRM gains as described therein.

b.

The power distribution shall be changed such that the fuel design limiting ratio for centerline melt (FDLRC) for any fuel assembly

• no longer exceeds 1.0

3.

Two RPS electric power monitoring channels for each inservice RPS MG set or alternate source shall be OPERABLE at all times.

3/4. 1-2 DRESDEN III DPR-25 Amendment No. ]A., ¢, yl 4.1 SURVEILLANCE REQUIREMENTS (Cont'd.)

a.

Maximum fuel design limiting ratio for centerline melt (FDLRC).

b.

Deleted.

3.

The RPS power monitoring system instrumentation shall be

• determined OPERABLE:

a.

At least once per 6 months by performing a CHANNEL FUNCTIONAL TEST, and

DRESDEN III DPR-25 Amendment No. Y.,, ¢, ~

4.1 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

1997b A comparison of Tables 4.1.l and 4.1.2 indicates that six instrument channels have not been included in the latter Table.

These are:

Mode Switch in Shutdown, Manual Scram, High Water Level in Scram Discharge Volume Float Switches, Main Steam Line Isolation Valve Closure, Generator Load Rejection, and Turbine Stop Valve Closure.

All of the devices or sensors associated with these scram functions are simple on-off switches and, hence, *calibration is not applicable; i.e., the switch is either on or off.

Further, these switches are mounted solidly to the device and have a very low probability of moving; e.g., the switches in the scram discharge volume tank.

Based on the above, no calibration is required for these six instrument channels.

B.

FDLRC shall be checked once per day to determine if the APRM gains or scram requires adjustment.

This may normally be done by checking the LPRM readings, TIP traces, or process computer calculations.

Only a small number of control rods are moved daily and thus the peaking factors are not expected to change significantly and thus a daily check of the FDLRC is adequate.

B 3/4. 1-20

Minimum No. of Operable Inst.

• Channels Per Trip System (1) 1 1

2 1

1 3

3 3

2 (5) 2 (5) (6) 1 DRESDEN III DPR-25 TABLE 3.2.3 Amendment No. Yt, ¢ INSTRUMENTATION THAT INITIATES ~OD BLOCK Instrument APRM upscale (flow bias) (7)

Dual Loop Operation Single Loop Operation APRM upscale (refuel and Startup/Hot Standby mode)

• APRM downscale (7)

Rod block monitor upscale (flow bias) (7)

Dual Loop Operation Single Loop Operation Rod block monitor downscale (7)

IRM downscale (3)

IRM upscale IRM detector not fully inserted in the core SRM detector not in startup position SRM upscale Scram discharge volume water level - high Trip Level Setting Less than or equal to

(.58 WD plus 50)/FDLRC (See Note 2)

Less than or equal to

(.58 WD plus 46.5)/FDLRC (See Note 2).

Less than or equal to 12/125 full scale Greater than or equal to 3/125 full scale Less than or equal to

(.65 WD plus 45)

(See Note 2)

Less than or equal to

• (.65 WD plus 41)

(See Note 2)

Greater than or equal to 5/125 full scale Greater than or equal to 5/125 full scale Less than or equal to 108/125 full scale N/A (See Note 4)

Less than or equal to 105 counts/sec.

Less than or equal to 25 gallons Notes:

(See Next Page) 3/4.2-12 1997b i\\

DRESDEN III DPR-25 Amendment No. ~, p, ~

TABLE 3.2.3 (Notes)

1.

For the Startup/Hot Standby and Run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function, except the SRM rod blocks, IRM upscale, IRM downscale and IRM detector not fully inserted in the core need not be operable in the "Run" position and APRM downscale, APRM upscale (flow bias), and RBM downscale need not be operable in the Startup/Hot Standby mode.

A RBM upscale need not be operable at less than 30%

rated thermal power.

One channel may be bypassed above 30% rated thennal power provided that a limiting control rod pattern does not exist.

For systems with more than one channel per trip system, if the first column cannot be met for both trip systems, the systems shall be tripped.

For the scram discharge volume water level high rod block, there is one instrument channel per bank.

2.

3.

4.

5.

6.

7.

1997b Wo percent of drive flow required to produce a rated core flow of 98 Mlb/hr.

FDLRC = fuel design limiting ratio for centerline melt.

IRM downscale may be bypassed when it is on its lowest range.

This function may be bypassed when the count rate is greater than or equal to 100 cps.

One of the four SRM inputs may be bypassed.

This SRM function may be bypassed in the higher IRM ranges when the IRM upscale Rod Block is operable.

Not required while performing low power physics test at atmospheric pressure during or'after refueling at power levels not to exceed 5 MW(t).

3/4.2-13

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

I-1997b I. Average Planar LHGR During steady state power operation, the Average Planar Linear Heat Generation Rate (APLHGR) of all the rods in any fuel assembly, as a function of average bundle exposure at any a~ial location, shall not exceed the maximum average planar LHGR shown in Figure 3.5-1 (consisting of two curves).

For operation during Single Loop Operation, the values of Figure 3.5-1 shall be decreased by a multiplicative factor of 0.91. If, concurrently, one Automatic Pressure Relief Subsystem relief valve is out-of-service, the values of Figure 3.5-1 shall be decreased by a multiplicative factor of 0.89 for 8x8 fuel

• an<f 0. 76 for 9x9 fuel. If at any time during operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation.

to within the prescribed limits. If the APLHGR is not returned to within the pre-scribed limits within two (2) hours, the reactor shall be brought to the Cold Shut-down condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.* Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

3/4.5-15 DRESDEN III DPR-25 Amendment No. ~. ~

4.5 SURVEILLANCE REQUIREMENT (Cont'd.)

I.

Average Planar Linear Heat Generation Rate (APLHGR)

-I The APLHGR for each type of fuel as a function of average bundle exposure shall be determined daily during reactor operation at greater than or equal to 25% rated thermal power.

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

J.

Local Steady State LHGR During steady state power operation above 25% rated thermal power, the linear heat generation rate (LHGR) of any rod in any fuel assembly at any axial location shall not exceed its maximum steady state LHGR (SLHGR) value shown in Figure 3.5-lA (consists of two curves).

That is, the Fuel Design Limiting Ratio for Exxon Fuel (FDLRX) shall not be greater than 1.0 where FDLRX = LHGR SLHGR Figure 3.5-lA depicts the SLHGR values for 8x8 and 9x9 fuel as a function of nodal I -

exposure.

1997b If at any time during operation above 25% rated thermal power, it is determined* by normal surveillance that FDLRX for any fuel assembly exceeds 1.0, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the FDLRX 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 />.

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

3/4.5-16 DRESDEN III DPR-25 Amendment No. y5, ~

4.5 SURVEILLANCE REQUIREMENT (Cont'd.)

J.

Linear Heat Generation Rate (LHGR)

The Fuel Design Limiting

(

Ratio for Exxon Fuel (FDLRX) shall be checked daily

• during reactor operation at greater than or equal to 25% rated thermal power.

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

K.

Local Transient LHGR At any time during power operation, above 25% rated thermal power the fuel design limiting ratio for centerline melt (FDLRC) shall not be greater than 1.0, where FDLRC = (LHGR) (1.2)

(TLHGR) (FRP)

Figure 3.5-lB depicts the TLHGR values for 8x8 and 9x9.

fuel as a function of nodal exposure.

If during operation, the FDLRC exceeds 1.0 when operating above 25% rated thermal power, either:

a.

b.

The APRM scram and rod block settings shall be reduced to the values given by the equations in Specifications 2.1.A.1 and 2,1.B.

This may be accomplished by increasing APRM gains as described therein.

The power distribution shall be changed such that the FDLRC no longer exceeds 1.0 DRESDEN III DPR-25

• Amendment No. "JI:>, ~, r1 4.5 SURVEILLANCE REQUIREMENT (Cont'd.)

K.

Transient Linear Heat Generation Rate (LHGR)

The fuel design limiting ratio for centerline melt (FDLRC) shall be checked daily during reactor operation at greater than or equal to 25% rated thermal power 3/4.5-17 1997b

llW -

J -

... s:

..:i a:

c 5

~

13.5 13.0 12.5 12.0 11.s 11.0 10.5 10.0 0

DRESDEN III DPR-25 Amendment No. ~. ¢ MAPLHGR LIMIT VS. BUNDLE AVERAGE EXPOSURE ANF 8x8 FUEL

~

" !'I..

10,000 20,000 30,000 BUNDLE AVERAGE EXPOSURE (MWD/MTU)

The above graph is based on the following MAPLHGR summary for ANF 8x8 fuel design:

Bundle Average Exposure (MWD/MTU).

1997b 0

10,000 15,000 18,000 20,000 25,000 30,000 35,000 Figure 3.5-1 (Sheet 1 of 2) 3/4.5-18 MAPLHGR Limit (kw/ft) 13.0 13.0 13.0 12.85 12.60 11.95 11.20 10.45

DRESDEN III DPR-25 Amendment No. ~* ~* ~

11.S 11.0 10.s

10.0 J 9.5

!£. 9.0

~

~. 8.S

~. 8.0 1.s 1.0 6.S 6.0 0

MAPLHGR LIMIT VS. BUNQLE AVERAGE EXPOSURE ANF 9x9 FUEL 10,000 20,000 BUNDLE AVERAGE EXPOSURE (MWD/MTU) 30,001)

The above graph is based on the following MAPLHGR summary for ANF 9x9 fuel design:

Bundle Average Exposure (MWD/~TU) 1997b 0

5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 Figure 3.5-1 (Sheet 2 of 2) 3/4.5-19 MAPLHGR Limit (kw/ft) 11.40 11.75 11.40 10.55 9.70 8.85 8.00 7.15 6.30 40,000_

---

e -_,

D a:

a D -

D

. u.*

D * =...

D..

c:

1997b 8 x 8 Fuel Exposure 0.00 25.40 42.00 LGHR 16.00 14..10 9.30 DRESDEN III Amendment No.

PLANAR EXPOSURE (GWD./MTU)

Figure 3.5-lA 9 >< 9 Fuel Exposure 0.00 5.00 25.20 48.00 LHGR 14.50 14.50 10.80 7.20 STEADY STATE LINEAR HEAT GENERATION RATE LIMIT (SLHGR)

VS. NODAL EXPOSURE 3/4.5-20 DPR-25

20 19

~18

, :-11

.E

~ 16 11:11 ai: 1 5 c

0 *--; 14 c :: 13 11:11

~ 12 1997b DRESDEN III DPR-25 Amendment No. Y.,, ~

~

~

~

"' ' ~

• ~

~

~

~

10 20 30 40 PLANAR EXPOSURE (GWD/MTU)

Exposure LHGR 0.00 19.20 25.40.

16.90 43.20 10.80 48.00 10.00 Figure 3.5-18 TRANSIENT LINEAR HEAT GENERATION RATE LIMIT (TLHGR)

VS. NODAL EXPOSURE FOR ALL FUEL TYPES 3/4.5-21 50

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

1-1997b L.

Minimum Critical Power Ratio (MCPR)

During steady state operation at rated core flow, MCPR shall be gr.eater than or equal to 1.39.

• For core flows other than rated, the MCPR Operating Limit shall be as follows:

1.

Manual Flow Control the MCPR Operating Limit shall be the value from Figure 3.5-2 sheet 1 or the above rated flow value, whichever is greater.

2.

Automatic Flow Control -

the MCPR Ope~ating Limit is the greatest of the following:

a.

The above rated flow value;

b.

The value from Figure 3.5-2 sheet 1; or

c.

The interpolated value from Figure 3.5-2 sheets 2 and

3.

3.

During Single Loop Operation. the rated flow MCPR operating limit shall be increased by an additive factor of. 0.01.

3/4.5-22 DRESDEN III Amendment No.

DPR-25

~.~

4.5 SURVEILLANCE REQUIREMENT (Cont'd.)

L.

Minimum Critical Power Ratio (MCPR)

• MCPR shall be determined daily during a reactor power operation at greater than or equal to 25% rated thermal power and following any change in power level or d~stribution that would cause operation with a limiting control rod pattern as described in the bases for Specification 3.3.8.5.

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

.1~

1997b If at any time during steady state power operation, it is determined that the limiting vaiue for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady state MCPR 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 />.

Surveil-lance and corresponding action shall continue until reactor operation is within the prescribed limits..

3/4.5-23 DRESDEN III DPR-25 Amendment No. f'j, ¢ 4.5 SURVEILLANCE REQUIREMENT (Cont'd.)

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

M.

Condensate Pump Room Flood Protection

1.

The system is installed to prevent or mitigate the consequences of flooding of the condensate pump room and shall be operable prior to startup of the reactor.

3/4.5-24 1997b DRESDEN III DPR-25 Amendment No. Y.,. ¢ 4.5 SURVEILLANCE REQUIREMENT (Cont'd.)

M.

Condensate Pump Room Flood Protection

1.

The following surveillance requirements shall be observed to assure that the condensate pump room flood protection is operable.

a.

The testable pene-trations through the walls of CCSW pump vaults shall be checked during each operating cycle by pressuriz-ing to 15 plus or minus 2 psig and checking for leaks using a soap bubble solution.

The cri-teria for acceptance should be no visible leakage through the soap bubble solu-tion.. The bulkhead door shall be checked during each operating cycle by hydrostatical ly test"ing the door at

• 15 plus or minus 2 psig and checking to verify that leakage around the door is less than one gallon per hour.

.... J. 70 0

CIC --e 1

• so 0 -'

~ 1.50

,Z c

II -

1. 40 u z:

~ 1. 30 CIC a..

0 Q

u

~ 1.10 CIC CIC a..

u

=-1.0~0 1997b

~18 9x9 ~

~

~

~

~

DRESDEN.III Amendment No.

~-

. OPR-25

~

~

30 40 50 60 70 80 TOTAL CORE FLOW (% RATED, 98 MLB/HR)

The above curves are based on the following MCPR Limit summary for reduced Total Core flow:

Total Core flow

(% Rated) 100 90 80 70 60 50 40 MCPR Limit 8x8 9x9 1.10 1.15

1. 21 1.28 1.36 1.46 1.60 1.09 1.14 1.20

. 1. 26 1.34 1.44 1.57 Figure 3.5-2 (Sheet 1 of 3)

MCPR Limit for reduced Total Core Flow 3/4.5-25

~

90 100 p

DRESDEN III Amendment No.

OPR-25 1.ao..-----..-~i.-39'!'"""'"'1~----,....----~----..,...----..,..-----r-----..

1.36 e 1

• 1 o a----....- 1.32

=-

1.28

~ 1. 60L--~--4__;.~-4-~~~~~+-~~t--~-+~~-r-~----;

Z:. -

31.501----~---l----+.;;~~~-~~;--;~:::--r---r---1

~

0 a::

~1.401---~---1-----~~------~-----+----~q..~~~r--__;;:==-o::::1:::::--=--~

~

0 a::

~ 1." 3 Ol------+----1----4---+----f---+----i::::....""""==::::::::::::--t u..

0

~1.20L------+-----1-~--+----+-----t---"1-----t---~

c e1. 10L-------l-------+----__;.~------+-------t-------+-------t-----~

=

c 1. o~ 0 1997b 30 40 50 60 70 80 TOTAL CORE FLOW (% RATED, 98 MLB/HR)

The above 8x8 curves are based on the following MCPR operating limit summary for Automatic Flow Control:

MGPR Operating Limit Total Core Flow for 8x8 fuel*

(% Rated) 1.28

1. 32

1. 36

1. 39 100 1.28

1. 32

1. 36

1. 39 90

1. 31

1. 35

1. 39 1.43 80

1. 34

1. 39 1.43 1.46 70

1. 39 1.44 1.48

1. 52 60 1.45

1. 49 1.54

1. 58 50 1.52

1. 56

1. 61

1. 65 40

1. 66

1. 71

1. 76 1.80

• Column headers are MCPR Operating Limits at rated flow.

Figure 3.5-2 (Sheet 2 of 3) 8x8 MCPR Operating Limit For Automatic Flow Control 3/4.5-26 90 100

~..

1.90

1. 80 e 1. 70

~

=-

~ 1

• 60 A.

u

1.50 0 --

~ 1.40

-=

0

~ 1. 30 u

~

c

~* 1. 2 0 c

1. 10 1

• o~ 0 1997b 1

• J 9

~

1

• 35 1

• 31 ~

i--...._

~ t:::::::

~

~

~

DRESDEN III Amendment ~Jo.

F=::::: 1---

~- ---

DPR-25

~

~

30 40 50 60 70 80 90 100 TOTAL CORE FLOW (% RATED, 98 MLB/HR)

The above 9x9 curves are based on the following MCPR operating limit summary for Automatic flow Control:

Total Core flow

(% Rated)'

100 90 80 70 60 50 40 MCPR Operating Limit for 9x9 fuel*

1.31 1.35 1.39

1. 31 1.34

1. 36 1.41 1.46

1. 53 1.68

1. 35

1. 38 1.41 1.45.

1. 51

1. 58 1.73

1. 39 1.42 1.45 1.49 1.55 1.62

1. 78

• Column headers are MCPR Operating Limits at rated flow.

figure 3.5-2 (Sheet 3 of 3) 9x9 MCPR Operating Limit for Automatic flow Control 3/4.5-27

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

3/4.5-28 1997b DRESDEN III DPR-25 Amendment No. -Y.,, ¢ 4.5 SURVEILLANCE REQUIREMENT (Cont'd.)

b.

The CCSW Vault Floor drain shall be checked during each operating cycle by assuring that water can be run through the drain line and actuating the air operated valves by operation of the following sensor:

i.

loss of air ii. high leve 1 in the condensate pump room (5'0")

c.

The condenser pit five foot trip shall have a trip setting of less than or equal to five feet zero inches.

The five foot trip circuit for each channel shall be checked once every three months.

The 3 and 1 foot alarms shall have a setting of less than or equal to three feet*zero inches and less than or equal to 1 foot 0 inches.

A logic system functional test, including all alarms, shall be performed during the refueling outage.

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

1997b

2.

The condenser pit 1.iater level switches shall trip the condenser circulating water pumps and alarm in the control room if water level in the condenser pit exceeds a level of 5 feet above the pit floor. If a failure occurs in one of these trip and alarm circuits, the failed circuit shall be immediately placed in a trip condition and reactor operation shall be permissible for the follot.Jing *seven days unless the circuit is sooner made operable.

3.

If Specification 3.5.M.1 and 2 cannot be met, reactor startup shall not commence or if operating, an orderly shutdown shall be initiated and the.

reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3/4.5-29 DRESDEN III DPR-25 Amendment No.

~, p 4.5 SURVEILLANCE REQUIREMENT (Cont'd.)

• '*.1 DRESDEN III DPR-25 Amendment No. ~, ¢ 3.5 LIMITING CONDITION FOR OPERATION BASES 1997b A.

Core Spray and LPCI Mode of the RHR System - This specification assures that adequate emergency cooling capability is available.

Based on the loss of coolant analyses included in References (1) and (2) in accordance with 10CFR50.46 and Appendix K, core cooling systems provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident, to limit the calculated peak clad temperature to less than 22oo*r, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to less than 1%, and to limit the calculated local metal-water reaction to less than 17%.

The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate.

The method and concept are described in Reference (3).

Using the results developed in this reference, the repair period is found to be less than 1/2 the test interval. This assumes that the core spray and LPCI subsystems constitute a 1 out of 3 system, however, the combined effect of the two systems to limit excessive clad temperatures must also be considered.

The test interval specified in Specification 4.5 was 3 months.

Therefore, an allowable repair period which maintains the basic risk considering single failures should be less. than 45 days and this specification is* within this period.

For multiple failures, a shorter interval is specified and to improve the assurance that the remaining (1) ~Loss of Coolant Accident Analyses Report for Dresden Units 2, 3 and Quad-Cities Units l, 2 Nuclear Power Stations," NED0-24146A, Revisions l, April 1979.

(2) NED0-20566, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50*

Appendix K.

(3) APED-"Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards" - April 1969, I.M.

Jacobs and P.W. Marriott.

B 3/4.5-30

DRESDEN III DPR-25 Amendment No.

~. f}lf 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

B.

1997b systems will function, a daily test is called for.

Although it is recognized that the information given in reference 3 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach prevents complete acceptance of this method at this time.

Therefore, the times stated in the specific items were established with due regard to judgement.

Should one core spray subsystem become inoperable, the remaining core spray and the entire LPCI system are available should the reactor core cooling arise.

To assure that the remaining core spray and LPCI subsystems and the diesel generators are available they are demonstrated to be operable immediately.

This demonstration includes a manual initiation of the pumps and associated valves and diesel generators.

Based on judgements of the reliability of the remaining systems; i.e. the core spray and LPCI, a 7-day repair period was obtained.

Should the loss of one LPCI pump occur, a nearly full complement of core and containment cooling equipment is available.

Three LPCI pumps in conjunction with the core spray _subsystem wi 11 perform the core cooling function.

Because of the availability of the majority of the core cooling equipment, which will be demonstrated to be operable, a 30-day repair period is justifiep. If the LPCI subsystem is not available, at least 2 LPCI pumps must be available to fulfill the containment cooling function.

The 7-day repair period is set on this basis.

Containment Cooling Service Water - The containment heat removal portio~ of the LPCI/containment cooling subsystem is provided to remove heat energy from the containment in the event of a loss of coolant accident.

For the flow specified, th~ containment long-term pressure is limited to less 1than 8 ps ig and, therefore, is more than ample to provide the required heat removal capability.

(Ref. Section 5.2.3.2 SAR).

The containment cooling subsystem consists of two sets of 2 service water pumps, 1 heat exchanger and 2 LPCI pumps.

Either set of equipment is capable of performing the containment cooling function.

Loss of one containment cooling service water pump does not seriously jeopardize the containment cooling capability as any 2 of the remaining three pumps can satisfy the cooling requirements.

Since there is some redundancy left a 30-day repair period is adequate.

Loss B 3/4.5-31

DRESDEN III DPR-25 Amendment No. f'j, ~

• 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

1997b of 1 containment cooling subsystem leaves one remaining system to perform the containment cooling function.

The operable system is demonstrated to be operable each day when the above condition occurs.

Based on the facts that when one.

containment cooling subsystem becomes inoperable only one system remains which is tested daily.

A 7-day repair period was specified.

C.

High Pressure Coolant Injection - The high pressure coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI and core spray subsystems can protect the core.

The HPCI meets this requirement without the use of off-site electrical power.

For the pipe breaks for which the liPCI is intended to function the core never uncovers and is continuously cooled and thus no clad damage occurs.

(Ref.

Section 6.2.5.3 SAR).

The repair times for the limiting conditions of operation were set considering the use of the HPCI as part of the isolation cooling system.

D.

Automatic Pressure Relief - The relief valves of the automatic pressure relief subsystem are a back-up to the HPCI subsystem.

They enable the core spray and LPCI to provide protection against the small pipe break in the event *of HPCI failure, by depressurizing the reactor vessel rapidly eno~gh to actuate the core sprays and LPCI.

The core spray and LPCI provide sufficient flow of coolant to adequately cool the core.

Analyses have shown that only four of the five valves in the Automatic Oepressurization System are required to operate.

Loss of one of the relief valves does not significantly affect the pressure-relieving capability, therefore continued operation is acceptable provided the appropriate MAPLHGR reduction factor is applied to assure compliance with the 2200°F PCT limit.

Loss of more than one relief valve significantly reduces the pressure relief capability of the ADS; thus: a seven day repair period is specified with the HPCI available, and.a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> repair period otherwise.

E.

Isolation Cooling System - The turbine main condenser is normally available.

The isolation condenser is provided for core decay heat removal following reactor isolation and scram.

The isolation condenser has a heat removal capacity sufficient to handle the decay heat production at 300 seconds following a scram.

Water will be lost from the reactor vessel through the relief valves in the 300 seconds following isolation and scram.

This represents a minor loss relative to the vessel inventory.

B 3/4.5-32

DRESDEN III DPR-25 Amendment No. ~, ~

3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

1997b The system may be manually initiated at any time.

The system is automatically initiated on high reactor pressure in excess of 1060 psig sustained for 15 seconds.

The time delay is provided to prevent unnecessary actuation of the system during anticipated turbine trips.

Automatic initiation is provided to minimize the coolant loss following isolation from the main condenser.

To be considered operable the shell side of the isolation condenser must contain at least 11,300 gallons of water.

Make-up water to the shell side of the isolation condenser is provided by the condensate transfer pumps from the condensate storage tank.

The condensate transfer pumps are operable from on-site power.

The fire protection system is also available as !)lake-up water.

An alternate method of cooling the core upon isolation from the main condenser is by using the relief valves and HPCI subsystem in a feed and bleed manner.

Therefore, the high pressure relief.function and the HPCI must be available together to cope with an anticipated transient so the LCO for HPCI and relief valves is set upon this function rather than their function as depressurization means for a small pipe *break.

F.

Emergency Cooling Availability - The purpose of Specification D is to assure a minimum of core cooling equipment is available at all times. If, for example, one core spray were

• out of service and the diesel w!'l_ich power.ed the opposite core spray were out of service, only 2 LPCI*pumps would be available.

Likewise, if 2 LPCI pumps were out of service and 2 containment service water pumps on the opposite side were also out of service rio containment cooling would be available. It is during refueling outages that major maintenance is performed *and during such time that all low pressure core cooling systems may be out of service. This specification provides that should this occur, no work will be performed on the primary system which could lead to draining

_the vessel. This work would include work on certain control rod drive components and recirculation system.

Thus, the specification precludes the events which could require core cooling.

Specification 3.9 must also be consulted to determine other requirements for the diesel generators.

Dresden Units 2 and 3 share certain process systems such as the makeup demineralizers and the radwaste system and also some safety systems such as the standby gas treatment system, batteries,. and diesel generators.

All of these systems have been sized to perform their intended function considering the simultaneous operation of both units.

B 3/4.5-33

.. ~...

• • ~

DRESDEN III DPR-25 Amendment No. -Y.,, l}1 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

G.

H.

I.

1997b For the safety related shared features of each plant, the Technical Specifications for that unit contain the operability and surveillance requirements for the shared feature; thus, the level of operability for one unit is maintained independently of the status of the other.

For example, the shared ~iesel (2/3 diesel) would be mentioned in the specifications for both Units 2 and 3 and even if Unit 3 were in the Cold Shutdown Condition and needed no diesel power, readiness of the 2/3 diesel would be required for continuing

. Unit 2 operation.

Specification 3.5.F.4 provides that should this occur, no work will be performed which could preclude adequate emergency cooling capability being available.

Work is prohibited unless it is in accordance with specified procedures which limit the period that the control rod drive housing is open and assures that the worst possible loss of coolant resulting from the work will not result in uncovering the reactor core.

Thus, this specification assures adequate core cooling.

Specification 3.9 must be consulted to determine other requirements for the diesel generator.

Specification 3.5.F.5 provides assurance that an adequate supply of coolant water is immediately available to the low

.pressure core cooling systems and that the core will remain covered in the event of a loss of coolant accident while the reactor is depressurized with the head removed.

Maintenance of Filled Discharge Pipe - If the discharge p1p1ng of the core spray, LPCI, and HPCI are not filled, a water hammer can develop in this piping when the pump and/or pumps are started.

Average Planar LHGR This specification assures that the peak cladding temperature following a postulated design basis loss-of-coolant accident will not exceed the 2200°F limit specified in 10CFR50 Appendix K considering the postulated affects of fuel pellet densification.

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average LHGR of all the rods in a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within a fuel assembly.

Since expected B 3/4.5-34

DRESDEN III DPR-25 Amendment No. "¢, ¢ 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

U)

(2)

(3)

(4) 1997b local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than plus or minus 20°F relative to the peak temperature for-a typical fuel design, the limit on the average planar LHGR is sufficient to assure that calculated temperatures are below the 10CFR50, Appendix K limit.

The maximum average planar LHGRs shown in Figure 3.5.1 are based on calculations employing the models described in References (1) '* (2) and (3).

Power operation with APLHGRs at or below those shown in Figure 3.5.1 assures that the peak cladding temperature following a postulated loss-of-coolant accident will not exceed the 2200°F limit.

ANF has analyzed the effects Single Loop Operation has on LOCA events (Reference 4).

For breaks in the idle loop, the above Dual Loop Operation results are conservative (Reference 1).

For breaks in the active loop, the event is more severe primarily due to a more rapid loss of core flow.

By applying a multiplicative 0.91 reduction factor to the results of the previous analyses, ~11 applicable criteria are met. *

"Lossof Coolant Accident Analyses Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nuclear Power Stations," NED0-2414GA, Revision l, April, 1979.

XN-NF-81-75 "Dresden Unit 3 LOCA Model Using the ENC EXEM Evaluation Model MAPLHGR Results" XN-NF-85-63 "Dresden Unit 3 LOCA-ECCS Analysis MAPLHGR results for 9x9 fuel", dated September 1985.

ANF-84-111, "LOCA-ECCS Analysis for Dresden Units During Single Loop Operation with ANF Fuel", September 1987.

B 3/4.5-35

• -1

3.5 1997b DRESDEN III Amendment No.

LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

J.

Local Steady State LHGR DPR-25

-,.,, ¢ This specification assures that the maximum linear heat generation rate in any fuel rod is less than the design linear heat generation rate even if fuel pellet densification is postulated. This provides assurance that the fuel end-of-life steady state criteria are met.

K.

Local Transient LHGR This specification provides assurance that the fuel will neither experience centerline melt nor exceed 1% plastic cladding strain for transient overpower events beginning at any power and terminating at 120% of rated thermal power.

L.

Minimum Critical Power Ratio (MCPR)

The steady-state values for MCPR specified in the Specifica-tion were determined using the THERMEX thermal limits methodology described in XN-NF-80-19, Volume 3.

The safety limit implicit in the Operating limits is established so that during sustained operation at the MCPR safety limit, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.

The Limiting Transient delta CPR implicit in the operating limits was calculated such that the occur-rence of the limiting transient from the operating limit will not result in violation of the MCPR safety limit in at least 95% of the random statistical combinations of uncertainties.

Transient events of each type anticipated during operation of a BWR/3 were evaluated to determine which is most restrictive in terms of therma1 margin requirements.

The generator load rejection/turbine trip without bypass is typically the limit-ing event.

The thermal margin effects of the event are evaluated with the THERMEX Methodology and appropriate MCPR

• limits consistent with the XN-3 critical power correlation are determined.. Several factors influence which transient results in the largest reduction in critical power ratio, such as* the cycle-specific fuel loading, exposure and fuel type.

The cur-rent cycle's reload licensing analyses identifies the limiting transient for that cycle.

As described in Specification 4.3.C.3 and the associated Bases, observed plant data were used to determine the average scram performance used in the transient analyses for determin-ing the MCPR Operating Limit. If the current cycle scram time performance falls outside of the distribution assumed in the analyses, an adjustment of the MCPR limit may be required to maintain margin to the MCPR Safety Limit during transients.

Compliance with the assumed distribution and adjustment of the MCPR Operating Limit will be performed as directed by the nuclear fuel vendor in.accordance with station procedures.

B 3/4.5-36

DRESDEN III DPR-25 Amendment No. ¢, l}Q 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

M.

1997b For core flows less than rated,.the MCPR Operating Limit established in the specification is adjusted to provide protection of the MCPR Safety Limit in the event of an uncontrolled recirculation flow increase to the physical limit of pump flow.

This protection is provided for manual and automatic flow control by choosing the MCPR operating limit as the value from Figure 3.5-2 Sheet 1 or the rated core flow value, whichever is greater.

For Automatic Flow Control, in addition to protecting the MCPR Safety Limit during the flow run-up event, protection is provided against violating the rated flow MCPR Operating Limit during an automatic flow increase to rated core flow.

This protection is provided by the reduced flow MCPR limits shown in Figure 3.5-2 Sheet 2 or 3 where the curve corresponding to the current rated flow MCPR limit is used (linear interpolation between the MCPR limit lines depicted is permissible).

Therefore, for Automatic Flow Control, the MCPR Operating Limit is chosen as the value from figure 3.5-2 Sheet l, Sheet 2, Sheet 3 or the rated flow value, whichever is greatest.

Analyses have demonstrated that transient events in Single Loop Operation are bounded by those at rated conditions; however, due to the increase in the MCPR fuel cladding integrity safety limit in Single Loop Operation, an equivalent adder must be uniformly applied to all MCPR LCO to maintain the same margins to the MCPR fuel cladding.

integrity safety limit.

• flood Protection Condensate pump room flood protection will assure the availability of the containment coo Ung service water system (CCSW) during a postulated incident of flooding in the turbine building.

The redundant level switches in the condenser pit will preclude any postulated flooding of the turbine building to an elevation above river water level.

The level switches provide alarm and circulating water pump trips in the event a water level is detected in the condenser pit.

B 3/4.5-37

-1

DRESDEN III DPR-25 Amendment No. Y.,, ¢ 4.5 SURVEILLANCE REQUIREMENT BASES 1997b (A thru F)

The testing interval for the core and containment cooling systems is based on quantitative reliability analysis, judgement and practicality. The core cooling systems have not been designed to be fully testable during operation.

For example the core spray final admission valves do not open until reactor pressure has fallen to 350 psig thus during operation even if high drywell pressure were stimulated the final valves would not open.

In the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable.

The systems can be automatically actuated during a refueling outage and this will be done.

To increase the availability of the individual components of the core and containment cooling systems the components which make up the system i.e.*,

instrumentation, pumps, valve operators, etc., are tested more frequently.

The instrumentation is functionally tested each month.

Likewise the pumps and motor-operated valves are also tested each month to assure their operability.

The combination of a yearly simulated automatic actuation test and monthly tests of the pumps and valve operators is deemed to be adequate testing of these systems.

With components or subsystems out-of-service overall core and containment cooling reliability is maintained by demonstrating the operability of the remaining cooling equipment.

The degree of operability to be demonstrated depends on the nature of the reason for the out-of-service equipment.. For*routine out-of-service periods caused by preventative maintenance, etc., the pump and valve operability checks will be performed to demonstrate operability of the remaining components.

However, if a failure, design deficiency, etc., caused the out-of-service period, then the demonstration*of operability should be thorough enough to assure that a similar problem does not exist on the remaining components.

For example, if an out-of-service period were caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test in addition to the operability checks.

The requirement of 180 psig at 3500 gpm at the containment cooling service water (CCSW) pump discharge provides adequate margin to ensure that the LPCI/CCSW system provides the design B 3/4.5-38

DRESDEN III DPR-25 Amendment No. y:,, ¢ 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

1997b bases cooling water flow and maintains 20 psig differential pressure at the containment cooling heat exchanger.

This differential pressure precludes reactor coolant from entering the river water side of the containment cooling heat exchangers.

The verification of Main Steam Relief Valve operability during manual actuation surveillance testing must be made independent of temperatures indicated by thermocouples downstream of the relief valves. It has been found that a temperature increase may result with the valve still closed.

This is due to steam being vented through the valve actuation mechanism during the surveillance test.

By first opening a turbine bypass valve, and then observing its closure response during relief valve actuation, positive verification can be made for the relief

• valve opening and passing steam flow.

Closure response of the turbine control valves during relief valve manual actuation would likewise serve as an adequate verification for relief valve opening.

This test method may be performed over a wide range of reactor pressure greater than 150 psig.

Valve operat_ion below 150 psig is limited by the spring tension exhibited by the relief valves.

G.

Deleted H.

Maintenance of Filled Discharge Pipe I.

The surveillance requirements to assure that the discharge piping of the core spray, LPCI, and HPCI systems are filled provides for a visual observation that water flows from a high point vent.

This ensures that the line is in a full con-dition.

Between the monthly intervals at which the lines are vented, instrumentation has been provided to monitor the presence of water in the discharge piping.

This instrumenta-tion will be calibrated on the same frequency as the safety system instrumentation.

This period of periodic testing ensures that during the int~rvals between the monthly checks the status of the discharge piping is monitored on a con-tinuo_us basis.

Average Planar LHGR At core thermal power levels less than or equal to 25 per cent, operating plant experience and thermal hydraulic analyses indicate that the resulting average planar LHGR is below the maximum average planar UiGR by a considerable B 3/4.5-39

.f**

DRESDEN III DPR-25 Amendment No.

~, ¢ 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

1997b margin; therefore, evaluation of the average planar LHGR below this power level is not necessary.

The daily requirement for calculating average planar LHGR above 25 percent rated thermal power is sufficient since power distribution shifts are slow when there have not been significant power or control rod changes.

J.

Local Steady State LHGR The LHGR for all fuel shall be checked daily during reactor operation at greater than or equal to 25 percent power to determine if fuel burnup or control rod movement has caused changes in power distribution.

A limiting LHGR value is precluded by a considerable margin when employing a permissible control rod pattern below 25% rated thermal power.

K.

Local Transient LHGR The fuel design limiting ratio for centerline melt (FDLRC) shall be checked daily during reactor operation at greater than or equal to 25% power to determine if fuel burnup or control rod movement has caused changes in power distribution.

The FDLRC limit is designed to protect against centerline melting of the fuel during anticipated operational occurrences.

L.

Minimum Critical Power Ratio (MCPR)

M.

At core thermal power levels less than or equal to 25 percent; the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicates that the resulting MCPR value is in excess of requirements by a considerable margin.

With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.

The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

Flood Protection The watertight bulkhead door and the penetration seals for pipes and cables penetrating the vault walls have been designed to withstand the maximum flood conditions.

To assure that their installation is adequate for maximum flood conditions, a method of testing each seal has been devised.

B*3/4.5-40

DRESDEN III DPR-25 Amendment No. ~. ¢ 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

1997b To test a pipe seal, another test seal is installed in the opposite side of the penetration creating a space between the two seals that can be pressurized.

Compressed air is then supplied -to a fitting on the test seal and the space inside the sleeve is pressurized to approximately 15 psi.

The outer face of the permanent seal is then tested for leaks using a soap bubble solution.

On compl~tion of the. test, the test seal is removed for use on other pipes and penetrations of the same size.

In order to test the watertight bulkhead doors, a test frame must be installed around each door.

At the time of the test, a reinforced steel box with rubber gasketing is clamped to the wall around the door.

The fixture is then pressurized to approximately 15 psig to test for leak tightness.

Floor drainage of each vault is accomplished through a carbon steel pipe which.penetrates the vault.

When open, this pipe will drain the vault floor to a floor drain sump in the condensate pump room.

Equipment drainage from the.vault coolers and the CCSW pump bedplates will also be routed to the vault floor drains.

The old equipment drain pipes will be permanently capped to pre-clude the possibility of back-flooding the vault.

As a means of preventing backflow from outside the vaults in the event of a flood, a check valve and an air operated valve are installed in the 2" vault floor drain line 6'0" above the floor of the condensate pump room.

The check valve is a 2" swing check designed for 125 psig service.

The air operated valve is a control valve designed for a 50 psi differential pressure.

The control valve will be in the normally open position in the energized condition and will close upon any one of the following: *

a.

Loss of air or power

b.

High level (5'0") in the condensate pump room Closure of the air operated valve on high water level in the condensate pump room is effected by use of a level switch set at a water level of 5'0".

Upon actuation, the switch will close the control valve and alarm in the control room.

B 3/4.5-41

DRESDEN III DPR-25 Amendment No. 1'. ¢ 4.5 SURVEI~LANCE REQUIREMENT BASES (Cont'd.)

1997b The operator will also be aware of problems in the vaults/

condensate pump room if the high level alarm on the equipment

• drain sump is not terminated in a reasonable amount of time.

It must be pointed out that these alarms provide information to the operator but that operator action upon the above alarms is not a necessity for reactor safety since the other provisions provide adequate protection.

A system of level switches has been installed in the condenser pit to indicate and control flooding of the condenser area.

The fol lowing switches are installed:

Level

a.

l '0" (1 switch)

b.

3'0" (1 switch)

c.

5'0" (2 redundant switch pairs)

Function Alarm, Panel Hi-Water-Condenser Pit Alarm, Panel High-Circ.

Water Condenser Pit Alarm and Circ. Water Pump Trip Level (a) indicates water in the co*ndenser pit from either the hotwell or the circulating water system.

Level (b) is above the hotwell capacity and indicates a probable circulating water failure.

Should the switches at level (a) and (b) fail or the operator fail to trip the circulating water pumps on alarm at level (b), the actuation of either level switch pair at lev~l (c) shall trip the circulating water pumps automatically and alarm in the control room.

These redundant level switch pairs at level (c) are designed and installed to IEEE-279, "Criteria for Nuclear Power Plant Protection Systems."

As' the cir-culating water pumps are tripped, either manually or auto-matically, at level (c) of 5' 0", the maximum water level reached in _the ~-0ndenser pit due t<L,pu!J!Pil'.19 wi 11 be at the 491'0" elevation no* above condenser.pit floor elevation 481'0"; 5' plus an additional 5' attributed to pump coastdown).

In order to prevent overheating of the CCSW pump motors, a vault cooler is supplied for each pump.

Each vault cooler is designed to maintain the vault at a maximum 105°F temperature during operation of its respective pump.

For example. if CCSW pump 26-1501 starts, its cooler will also start and compensate B 3/4.5-42

DRESDEN III DPR-25 Amendment No. -ps, r; 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

1997b for the heat supplied to the vault by the 2B pump motor keeping the vault at less than 105°F.

Each of the coolers is supplied with cooling water from its respective pump's discharge line.

After the water has been passed through the cooler, it returns to its respective pump's suction line.

In this way, the vault coolers are supplied with cooling water totally inside the vault. The cooling water quantity needed for each cooler is approximately 1% to 5% of the design flow of the pumps so that the recirculation of this small amount of heated water will not affect pump or cooler operation.

Operation of the fans and coolers is required during pump operability testing and thus additional surveillance is not required.

Verification that access doors to each vault are closed, following entrance by personnel, is covered by station operating procedures.

B 3/4.5-43

• r 3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

e.

The suction valve in the idle loop shall be closed and electrically isolated except when the idle loop is being prepared for return to service; and

f. If the tripped pump is out of service for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, imp le-ment the following additional restrictions:

i.

The flow biased RBM Rod Block LSSS shall be reduced by 4.0%

(Sped fication 3.2.C.l);

ii. The flow b1ased APRM Rod Block.

LSSS shall be reduced by 3. 5%

(Sped f ication 2.1.B);

iii. The flow biased APRM scram LSSS shall be reduced by 3.5%

(Sped fication 2.1.A.l);

iv.

The MCPR Safety Limit shall be increased-by 0.01 (Specification 1.1.A);

v.

The MCPR Operating Limit shall be increased by 0.01 (Sped fication 3.5.L.3);

3/4.6-15 1997b DRESDEN III DPR-25 Amendment No. Y.,, ~, ~

4.6 SURVEILLANCE REQUIREMENT (Cont'd.)

3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

1997b vi.

The MAPLHGR Operating Limit shall be reduced by a multiplicative factor of 0.91 (Specification 3.5.I). If, concurrently, one Automatic Pressure Relief Subsystem relief valve is out-of-service, the MAPLHGR Operating Limit shall be reduced by a multiplicative factor of 0.89 for 8x8 fuel and 0.76 for 9x9 fuel.

4.

Core thermal power shall not exceed 25% of rated without forced recircu-lation. If core thermal power is greater than 25%

of rated without forced recirculation, action shall be initiated within 15 minutes to restore operation to within the prescribed limits and core thermal power shall be returned to within the prescribed limit Within two (2) hours.

I. Snubbers (Shock Suppressors) 3/4.G-lG DRESDEN III DPR-25 Amendment No. V6, ~, fl 4.G SURVEILLANCE REQUIREMENT (Cont'd.)

I. Snubbers (Shock)

Suppressors)

The following surveillance requirements apply to safety related snubbers.

DRESDEN III DPR-25 Amendment No. ~. ~

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

1997b In addition, during the start-up of Dresden Unit 2, it was found that a flow mismatch between the two sets of jet pumps caused by a difference in recirculation loops could set up a vibration until a mismatch in speed of 27% occurred.

The 10%

and 15% speed mismatch restrictions provide additional margin before a pump vibration problem will occur.

Reduced flow MCPR Operating Limits for Automatic Flow Control are not applicable for Single Loop Operation.

Therefore, sustained reactor operation under such conditions is not permitted.

Regions I and II of Figure 3.6.2 represent the areas of the power/flow map with the least margin to stable operation.

Although calculated decay ratios at the intersection of the natural circulation flow line and the APRM Rod Block line indicate that substantial margin exists to where unstable operation could be expected.

Specifications 3.6.H.3.b.,

3.6.H.3.c. and 4.6.H.3. provide additional assurance that if unstable operation should occur, it wi 11 be detected and corrected in a timely manner.

During the starting sequence of the inoperable recirculation pump, restricting the operable.recirculation pump speed below 65% of rated prevents possible damage to the jet pump riser braces due to exc*essive vibration.

The closure of the suction valve in the idle loop preverits the loss of LPCI through the idle recirculation pump into the

• downcomer.

Analyses have been performed which support indefinite operation in single loop provided the restrictions discussed in Specification 3.6.H:3.d. are implemented within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The LSSSs are corrected to account for backf low through the idle jet pumps above 20-40% of rated recirculation pump speed.

This assures that the original drive flow biased rod block and scram trip settings are preserved during Single Loop Operation.

The MCPR safety limit has been increased by 0.01 to account for core flow and TIP reading uncertainties which are used in the statistical analysis of the safety limit.

In addition, the MCPR Operating Limit has also been increased by 0.01 to maintain the same margin to the safety limit as during Dual Loop Operation.

B 3/4.6-36

DRESDEN III Amendment No.

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

DPR-25

~.Pf The mulplicative 0.91 reduction of MAPLHGR Operating Limit accounts for the more rapid loss of core flow during Single Loop Operation than during Dual Loop Operation.

The more conservative MAPLHGR reduction factors of 0.89 for 8x8 fuel and 0.76 for 9x9 fuel are applied if one relief.valve and one recirculation loop are inoperable at the same time.

The small break LOCA is the concern for one relief valve out-of-service; the large break LOCA is the concern for Single Loop Operation.

Selecting the more conservative MAPLHGR multipliers will cover both the relief valve out-of-service and Single Loop Operation.

Specification 3.6.H.4. increased the margin of safety for thermal-hydraulic stability and for startup of recirculation pumps from natural circulation conditions.

I.

Snubbers (Shock Suppressors)

Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient while allowing normal thermal motion during startup and shutdown.

The consequence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads. It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be operable during reactor operation.

Because the snubber protection is required only during low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replacements.

In case a shutdown is required, the allowance of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to reach a cold shutdown condition will permit an orderly shutdown consistent with standard operating procedures.

Since plant startup should not commence with knowingly defective safety related equipment, Specification 3.6.I.4 prohibits startup with inoperable snubbers.

When a snubber is found inoperable, a review shall be performed to determine the snubber.mode of failure.

Results of the review shall be used to determine if an engineering evaluation of the safety-related system or component is necessary.

The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the support component or system.

All safety related hydraulic snubbers are visually inspected for overall integrity and operability. The inspection will include verification of proper orientation, adequate hydraulic fluid level and proper attachment of snubber to piping and structures.

All safety related mechanical snubbers are visually inspected for overall integrity and operability.

The inspection will include verification of proper orientation and attachments to the piping and anchor for indication of damage or impaired operability.

.* B 3/4.6 1**;p;;, D 3.10 LIMITING CONDITIONS FOR OPERATION (Cont'd.)

1997b G.

Fuel Storage Reactivity Limit

1.

The new fuel storage facility shall be such that the Keff dry is less than 0.90 and flooded is less than 0.95.

2.

Whenever a fuel assembly is stored in the spent fuel storage pool, the peak assembly reactivity in a reactor lattice distribution shall be limited to less than or equal to the following values:

Assembly Type Kinf GE 7x7 1.26 GE axe

1. 32 ANF 8x8

1. 33 ANF 9x9 1.27 Whenever storing other assembly types or fuel rods in the spent fuel storage pool, their peak reactivity shall be bounded by the most limiting Kinf value listed above.

H.

Loadi Over Spent Fuel Storage Pool No loads heavier than the weight of a ~ingle spent fuel assembly and handling tool shall be carried over fuel stored in the spent.fuel storage pool.

3/4.10-8 DRESDEN III DPR-25 Amendment No. ¥>, ¢ 4.10 SURVEILLANCE REQUIREMENTS (Cont'd.)

G.

Fuel Storage Reactivity Limit

1.

Prior to storing Fuel in the new fuel storage facility, an analysis must be performed to demonstrate that the criteria in 3.10.G.1 are satisfied.

2.

Prior to storing Fuel in the the spent fuel storage pool, an analysis must be performed to demonstrate that the criteria in 3.10.G.2 are satisfied.

ENCLOSURE C

SUMMARY

OF RELOAD AND PROPOSED LICENSE AMENDMENT This swnmary describes the fuel design, reload analyses, Technical Specification changes, Extended Load Line Limit Analysis results (ELLLA),

coastdown operation and Single Loop Operation (SLO) provisions for Dresden 3 Cycle 11, including a new LOCA-ECCS analysis for SLO and Relief Valve Out-of-Service (RVOOS) analyses.

Th~ evaluation is divided into six sections as follows:

I.

Reload Fuel and core Design II. Transient and Accident Analyses (including RVOOS)

III. Single Loop Operation (SLO)

IV.

Extended Load Line Limit Analysis (ELLLA)

v.

Description of Technical Specifications/License Changes VI.

Summary I.

RELOAD FUEL AND CORE DESIGN The Dresden 3 Cycle 11 Reload will consist of 72 ANF 9x9 reload fuel bundles designated as type XN-4H and 96 ANF 9x9 reload bundles designated as XN-4L.

These bundles have a central region enrichment_ of 3.35\\ and natural uranium in the last 6 inches of each end to yield an average assembly enrichment of 3.13\\.

The Dresden 3 Cycle 11 core loading will consist of the following fuel types:

Fuel Type Enrichment*

Number of Bundles XN-1 8x8 2.69\\

196 XN-2 8x8 2.83\\

184 XN-3 9x9 3.13\\

160 XN-3A 9x9 3.13\\

16 XN-4H 9x9 3.13\\

72 XN-4L 9x9 3.13\\

96 Total 724 All residual General Electric fuel is scheduled to be discharged during the EOClO outage.

Dresden 3 Cycle 11 will therefore represent the first Dresden cycle with an entire core of ANF fuel.

The core will be operated under the Single Rod Sequence (SRS) mode of operation as currently in use for Dresden 3 Cycle 10.

A.

Fuel Mechanical Design The mechanical design of the XN-4H and XN-4L 9x9 reload fuel is described in References l and 2.

In general, design criteria are established to limit the stress, strain and overall duty on the fuel rod or bundle during normal, transient and accident operation.

In addition, the fuel is designed to be mechanically compatible with reactor internals, fuel handling equipment and existing fuel.

• Bundle Average Enrichments

• The Cycle 11 reload batch size is 168 fuel assemblies including 72 XN-4H assemblies and 96 XN-4L assemblies.

With the exception of a change in the Gd2o3 concentration in the central region of the gadolinia-bearing fuel rods, the design of the XN-4H fuel is identical to the XN-4L fuel.

Both the XN-4H and XN-4L assemblies contain nine gadolinia-bearing fuel rods, with 3.0\\ Gd2o3 in the top six inches and bottom twelve inches of the enriched region of these rods.

In the XN-4H assemblies, the central region of the gadolinia-bearing rods contains 4.5\\.Gd2o3; the XN-4L assemblies contain 4.0\\

Gd203 in this region.

The 9x9 reload assemblies contain 79 fuel rods and two water rods.

Eight of the fuel rods are tie rods.

B.

Thermal Hydraulic Design Component hydraulic resistances for the constituent fuel types in the Dresden Unit 3 Cycle 11 core have been determined in single phase flow tests of full scale assemblies.

The hydraulic demand curves for ANF 8x8 and ANF 9x9 fuel in the Dresden 3 core is illustrated in Figure 3.1 of Attachment 1.

The MCPR Safety Limit has been calculated by ANF to be 1.05 for all fuel types for Dresden 3 Cycle 11 operation.

The analysis has considered each of the constituent fuel types, conservative local power distributions for each type and the worst (bounding) radial power distribution at which each type is expected to operate during the cycle.

This limit is therefore independent of fuel type designation.

c.

Fuel Linear Heat Generation Rate (LHGR) constraints

1.

Fuel Steady State LHGR Constraints Dresden's onsite core monitoring code (POWERPLEX) utilizes the parameter, Fuel Design Limiting Ratio for Exxon fuel (FDLRX), to assure that end-of-life steady state design criteria are met.

One way that FDLRX assures acceptable end-of-life conditions is by limiting the release of fission gas to the cladding plenum.

Assumed in the analysis are several transient overpower events; therefore, no power biasing is required for this limit.

FDLRX is defined as:

FDLRX =

LHGR SLHGR where LHGR = calculated local pin power (linear heat generation rate)

SLHGR = steady state linear heat generation rate limit

3 -

While the ANF steady state LHGR limits have not changed for this reload, the parameter FDLRX has now been explicitly defined in Section 3.5.J.

2.

centerline Melting and Cladding Strain A new subsection has also been added to sections 3.5/4.5 of the Tech Specs to reflect the POWERPLEX parameter (FDLRC) for monitoring compliance with another ANF thermal mechanical design criterion, i.e., that fuel design and operation will be such that fuel centerline melting and 1\\

plastic cladding strain are not expected for anticipated opera~ional occurrences (transients) throughout the life of the fuel.

To demonstrate compliance with this criterion, fuel rod centerline temperatures are determined at 120\\ overpower conditions as a check against calculated centerline melt temperatures.

Calculations have shown that the maximum fuel centerline temperature for the 9x9 fuel occurs at an exposure of~5000 MWD/MTU and decreases more rapidly with exposure than does the fuel melt temperature.

Thus the minimum margin to melt is also at 5000 MWD/MTU.

The peak centerline temperature at 120\\ overpower and 5000 MWD/MTU was calculated to be 4115°F with a corresponding melting point of 5050°F, so that the minimum margin to centerline melt is about 935°F..

The above results are for a urania rod (U02) in ANF's 9x9 design for Dresden 3 Cycle 11.

Although gadolinia rods cuo2-Gd2o3> have lower fuel melt temperature than a urania rod, the peak fuel temperatures for the gadolinia rods inherently operate at significantly lower power levels due to their gadolina content (high neutron absorption) and design enrichments.

ANF's transient LHGR limit, Fuel Design Limiting Ratio for centerline melt. (FDLRC) is therefore incorporated to protect the above criteria.

The LHGR limit must be protected at all power levels considering events which could cause the reactor power to increase as high as 120\\

of rated thermal power.

FDLRC is defined as:

FDLRC = LHGR

• 1.2 TLHRG

• FRP where LHGR = calculated local generation rate)

TLHGR = Tr.ansient Linear Limit pin power (linear heat.

Heat Generation Rate FRP

= Fraction of Rated Core Thermal Power In the event FDLRC is greater than 1.0, either the APRM flow biased scram must be reduced by a factor of FDLRC or the APRM gain adjust factors must be increased by a factor of FDLRC.

D.

Nuclear Design The specific neutronic design parameters and pin enrichments are contained in Section 3.0 of Reference 2.

The Dresden 3 Cycle 11 core loading pattern given in Figure 4.4 of Attachment 1 was analyzed assuming an EOC 10 core average exposure of 21,625 MWD/MTU.

The calculated core average exposure at BOC 11 is 14,548 MWD/MTU and EOC 11 is 23,429 MWD/MTU with a cycle exposure increment of 8,881 MWD/MTU.

The results of the neutronic analyses are provided in the following sections.

1.

Core Reactivity The calculated BOC 11 cold core keff for the all-rods~in condition is 0.9605.

The BOC 11 cold core keff with the strongest rod out and all other rods in is 0.9876.

This translates to a Shutdown Margin (SOM) of 1.24\\'1.k at BOC 11.

A minimum calculated shutdown margin of l.ll\\4k occurs at cycle exposures of 2000 MWD/MTU, 7000 MWD/MTU, and 8000 MWD/MTU.

Hence, the value of R in the required Technical Specifications shutdown margin of 0.25\\Ak + R would be 0.17\\~k (1.24\\.A,k -

1.11\\ + 0.04\\).

The 0.04\\J1k conservatively accounts for the effect of B4c settling in any residual inverted absorber tubes of the control blades.

The projected shutdown margin is there-fore in conformance with the minimum Technical Specifications limit.

The Standby Liquid Control System was calculated to provide a SOM of 6.19\\~k for cold, xenon free conditions with all control rods in their full power positions.

This meets the SOM Technical Specification requirement of 3.0\\Ak for Anticipated Transients Without scram (ATWS) conditions requiring use of standby Liquid Control.

2.

Core Stability The stability of the Dresden 3 Cycle 11 core was determined analytically using the NRC approved COTRAN code at various power and flow conditions.

The maximum decay ratio for natural recirculation flow at the 100\\ FCL is 0.35 (47.6\\ rated power, 31.5\\ rated flow); at the APRM rod block intercept, the decay ratio is 0.55 (58\\ rated power).

Since both of these decay ratios are less than the NRC SER surveillance criterion of 0.75 as calculated by COTRAN, no stability Technical Specification surveillance requirement is needed for Cycle 11 operation.

5 -

A stability comparison between dual loop operation (DLO) and single loop operation (SLO) was performed during Cycle 10 on December 3, 1986.

Based on the results of this test, it was

• concluded that the operating region of concern exhibits adequate margin to power/flow instabilities in SLO and DLO.

As expected, DLO was significantly more stable than SLO, demonstrating that stability monitoring Technical Specifications are not required during DLO.

Furthermore, the current SLO stability surveillances required by Technical Specifications are adequate for detecting any core wide or local instabilities while operating with a recirculation loop out-of-service.

3.

Fuel Storage Vault/Pool Criticality The criticality safety analysis of storing ANF 9x9 reload fuel in the Dresden high density spent fuel storage racks in the pool has been previously addressed in Reference 3.

The 9x9 reload fuel for D3Cll has been verified to meet the Technical Specification reactivity limits for fuel storage in both the spent fuel pool and the new fuel vault.

4.

ASEA-ATOM Control Blades Dresden 3 was the first U.S. BWR to utilize ASEA-ATOM (A-A) control blades.

As part of an EPRI-sponsored demonstration, eight A-A CR-82 design control blades were inserted in SRS locations during cycle 9.

References 4 and 5 provide an extensive description of the design and performance of the A-A designed control blades.

During the upcoming refueling outage, several of these control blades will be inspected by ASEA-ATOM personnel as part of the EPRI-sponsored program.

For Dresden 3 Cycle 11, 12 new CR-82B design A-A blades will be used.

While generally very similar to the CR-82 design, these blades incorporate several enhancements.

Reactivity characteris-tics are improved by replacing th~ upper six inches of the cover bar with hafnium instead of using stainless steel. Also, the radius of the center cut-out has been increased, making the cut-out more rounded than square.

A cylindrical filling piece has been used instead of a square piece, where the blade wings are welded together.

This has changed the throat of the weld to 0.157 inches from 0.177 inches.

The last change in design is the plugging of the helium fill hole (used for leakage testing) before seal welding.

ASEA has shown that these changes from the CR-82 design do not have any significant impact on the mechanical

)

characteristics of the control rod, and reactivity matching to existing GE blades is improved.

Therefore, use of the blades during Cycle 11 does not create any safety concerns.

• II.

TRANSIENT AND ACCIDENT ANALYSES A.

Corewide Transients

1.

Limiting HCPR Transients The following transients were evaluated to establish the limiting core wide Anticipated Operational occurrence:

a.

Generator Load Rejection without Bypass (LR w/o BP)

b.

Feedwater Controller Failure (FWCF)

c.

Loss of Feedwater Heating (LoFWH)

The analyses were performed using the same plant transient analysis methodology (References 6 and 7) as was used to establish thermal margin requirements for cycle 10 operation.

The approved XCOBRA-T hot channel model was used to calculate the limiting ACPRs.

In accordance with the ANF methodology as given in Reference 6, possible limiting transients have been evaluated; the limiting corewide transient is identified as the LR w/o BP.

Since this is the limiting rapid pressurization event, the treatment of several of the calculational uncertainties in the determination of MCPR operating limits is statistically applied.

This methodology includes a conservative deterministic multiplier of 110\\ on the calculated transient power to account for COTRANSA code uncertainties. Other variables are treated either in a bounding fashion or in a statistkal manner as in previous analyses.

The~CPR results of the plant transient analyses for the above events (as shown in Table 2.1 of Attachment 2) are clearly bounded by the proposed cycle 11 MCPR LCO of 1.39.

This conservative value, which is the limiting MCPR LCO for current (cycle 10) operation, was chosen for the following reasons:

a)

It allows removal of the Technical Specification provision for scram time dependency of the MCPR LCO,(as discussed in Section 4 below).

This not only simplifies the Tech.spec.

but also simplifies the related procedures and the process of performing daily MCPR surveillance.

b)

Results of cycle 10 operation and core management projections for cycle 11 indicate that adequate MCPR operating margin will continue to exist with a 1.39 Leo.

c)

Use of a conservative value increases the probability that 10 CFR 50.59 could be applied for subsequent reloads.

• 2.

Compliance to ASME Pressure Vessel Code The most limiting event for reactor vessel over-pressurization is the Main Steamline Isolation Valve (MSIV) closure without direct scram on Valve position.

As required by the ASHE code, no credit is taken for the electromatic relief valves.

The maximum calculated value of the pressure sensed in the steam dome was 1297 psig which corresponds to a maximum vessel pressure of 1324 psig at the lower plenum.

These values are less than the Technical Specification limit of 1345 psig (as measured by the steam dome pressure indicator) and the 1375 psig ASME vessel pressure limit, respectively.

3.

Reduced Flow Operation

4.

ANF has determined the required reduced flow MCPR operating limit for off-rated conditions to assure that MCPR complies with the Cycle 11 MCPR full flow operating limits during Automatic Flow Control (AFC) operation. This was determined by evaluating the MCPR for a given reactor power distribution at varying total reactor power and flow conditions. presents the reduced flow MCPR operating limits for AFC in Figures 5.1 and 5.2 as curves and Tables 5.4 and 5.5 in tabulated form.

When in Manual Flow Control at reduced flow conditions, ANF determined that two-pump excursions were the limiting pump run-up events.

The evaluation of the two recirculation pump flow excursion showed that this event will also bound single pump failures.

The reduced flow MCPR calculations were performed for events initiated from both the APRM rod block line and the 100\\ FCL.

Furthermore, the analysis conservatively assumed the reactor reached 120\\ rated thermal power at 110\\

rated flow. presents the results of the bounding two-pump run-up analyses for manual flow control in Figure 5.3 (as a curve) and.Table 5.7 (in tabulated form).

Control Rod scram Insertion Time surveillance Section 3.3.C of the Dresden 3 Technical Specifications currently requires surveillance measurements and limits on the average of individual control rod drive scram insertion times in order to assure adequate CRD scram performance.

If the average 90\\ scram insertion time is greater than 3.5 seconds, the unit must be in cold shut down within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

In addition, Section 3.5.K currently requires that if the average 90\\ scram insertion time exceeds another specified limit (2.73 seconds for Cycle 10), a MCPR penalty is applied based on the measured average insertion time observed during the surveillance.

By conserva-tively continuing to use the limiting MCPR LCO from the previous cycle (i.e., 1.39), not only are all licensing basis events clearly bounded for Cycle 11, but a desirable simplification of the MCPR Technical Specification becomes possible.

That is, the scram time dependency of the MCPR limit can be eliminated as shown in the following discussion.

8 -

The limiting scram time MCPR penalty for Cycle 11 (based on 9x9 fuel) would be (Tav-2.62)*0.08, where Tav is the average scram time observed during the surveillance.

Using 3.5 seconds

.(the Section 3.3.C Tech Spec limit) as Tav yields 0.07 for a MCPR scram time penalty.

Adding this to the LR w/o BP MCPRLco for 9x9 fuel.results in an adjusted MCPRLCO of 1.38, which is less than the proposed LCO of 1.39.

Therefore, the MCPR penalty associated with the scram insertion time surveillance can be deleted from the Technical Specifications as shown in Enclosure E.

It should be noted that the deletion of this provision in Section 3.5.K in no way affects the existing provisions on allowable CRD scram times in Section 3.3.C.

5.

Analysis With one Relief Valve out-of-Service The ANF analyses consider the effects of a relief valve out-of-service in two portions: the effects on the Loss of Coolant Accident (LOCA) and the effects on the plant transients.

LOCA Analysis (with RVOOS)

In evaluating the effect of operation with a RVOOS, ANF assumed the worst case single failure to be the HPCI system inoperable.

This leaves only ADS available to relieve high pressure during a small break LOCA~ With the number of operable relief valves reduced to four, the time to depressurize the reactor will be longer and the fuel peak cladding temperature (PCT) will be higher.

ADS operation is not required for large break LOCAs.

If the break is large enough, the reactor will quickly depressurize through the break itself without need of ADS.

A General Electric analysis (Reference 9) of RVOOS for Quad Cities Station indicated the limiting small break to be a 0.05 ft2 recirculation line break.

As Quad Cities and Dresden are similar plants, the 0.05 ft2 limiting break was also used in the ANF analysis for Dresden.

Based on the results of the ANF analysis, the following MAPLHGR multipliers were calculated:

Fuel Type 8x8 9x9 Multiplier 0.89 0.76 Reference 10 11 These multipliers are therefore included in the* proposed Tech.

Spec.changes as discussed in Enclosure E.

Transient Analysis (with RVOOS)

The Load Reject without Bypass (LR w/o BP) yields the limiting CPR as stated in the Limiting HCPR Transients section.

Several calculations were performed in the ANF analysis to evaluate the effect of a RVOOS on this transient~ Calculations assumed the highest capacity or first opening RVOOS and nominal or worst case plant parameters.

In all the cases the lowest CPR occurred before any of the relief valves had started to open.

Since the relief valves play no part in the transient until after the limiting CPR, the reduced ADS capacity with a RVOOS will not affect the A CPR results.

Since the ACPR results calculated assuming all relief valves operable can be applied to operation with one RVOOS, no change is required to the MCPR Technical spec if ica t ions.

The limiting overpressurization transient for Dresden is MSIV closure without flux trip as discussed in the Compliance to ASHE Pressure Vessel Code section in this report.

Because the ASME code does not allow credit for relief valve operation, operation with a RVOOS does not affect the overpressurization transient analysis.

The RVOOS has a negligible effect on the peak pressure from LR w/o BP.

With all valves operable the peak pressure was calculated as 1270 psia.

With the highest capacity RVOOS, the peak pressure increased by 1 psi. With the first opening RVOOS, the peak pressure increased by 5 psi.

These results demonstrate the insignificant effect on peak pressure from LR w/o BP when

.operating with one RVOOS, compared to the ASME pressure vessel criteria of 1375 psia for peak pressure and the corresponding Technical Specification limit of 1345 psia for peak steam dome pressure.

Allowable Repair Time with Two RVOOS The Reference 10 RVOOS analysis (Attachment 3) assumes one RVOOS, but also assumes HPCI inoperable.

As can be seen from Figures 1 and 2 of Reference 9, the clad temperature rises quickly until LPCI initiates.

LPCI cannot initiate until ADS reduces pressure, and the time to depressurize is a function of the number of RVs available (although the time to uncover the fuel is also a function of the break's blowdown capacity}.

Since the peak cladding temperature (PCT) in the Reference 10 analysis approaches the 2200°F PCT limit, allowing two RVOOS may exceed the PCT limit and the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time is therefore specified.

To allow a longer repair time HPCI operability must be credited.

Since HPCI must be tested upon finding two RVs inoperable, the Dresden Unit 3 Technical Specification changes (Enclosure E) use the same 7 day period as the recent amendment for Quad Cities Unit 1 Cycle 10 (Reference 12), provided HPCI is shown to be operable.

The ANF analysis for operation at Dresden Unit 3 with one RVOOS therefore supports the proposed Technical Specification changes.

B.

Local Transients

1.

Rod Withdrawal Error (RWE)

The Dresden Unit 3 Rod Block Monitor (RBM) upper setpoint at full flow is currently set at 110\\.

At this setting, the RWE would result in aACPR of 0.31 for 9x9 fuel and 0.30 for 8x8 fuel.

Continuing to use the limiting Cycle 10 MCPR LCO of 1.39 (for both fuel types) in Cycle 11 therefore bounds the RWE results.

c.

Coastdown with a Feedwater Heater out-of-Service (FHOOS)

Additional transient analyses were performed to support coastdown operation for EOC 11.

The analyses were performed with a Feedwater Heater out-of-Service (FHOOS).

A feedwater temperature reduction of 100°F was used to model the FHOOS.

The results in Attachment 2 show that the~CPRs for the transients analyzed with FHOOS are bounded by the.6CPRs for transients at normal feedwater temperature.

The above coastdown analyses were performed at a reduced d0me pressure of 992.1 psia.

To ensure that the results bounded coastdown while maintaining full pressure, analyses were also performed at the rated dome pressure of 1020 psia which demonstrated that the reduced pressure results are conservative.

D.

Accidents

1.

Fuel Loading Error The limiting fuel loading error ACPR of 0.19 is the same for the mis-located and the mis-oriented fuel assembly events.

The limiting mis-located result is for an XN-3A assembly and the limiting mis-oriented result is for an XN-4L assembly.

These events are also bounded by the proposed MCPR LOC of 1.39.

2.

Control Rod Drop Accident (RDA)

The maximum deposited fuel rod enthalpy for the RDA is calculated to be 187 cal/gm which is well below the Technical Specification limit of 280 cal/gm, provided the provisions of the "Banked Position Withdrawal Sequence" (BPWS) are followed below 20\\ power.

11 -

3.

Loss of Coolant Accident (LOCA)

ANF has previously performed design basis LOCA analyses for Dresden 3 utilizing 8x8 fuel (Reference 8) and 9x9 fuel

• (Reference 11).

These analyses provided MAPLHGR limits which remain applicable for the ANF 8x8 and 9x9 fuel in Cycle 11 during Dual Loop Operation (DLO).

These previous NRC-approved LOCA results for DLO demonstrated compliance with 10 CFR 50.46 criteria on peak cladding temperature (2200°F), local clad oxidation (<17\\) core wide cladding reaction (<l\\), maintaining coolable geometry, and long term cooling capability.

LOCA results for SLO are addressed in the following section.

III. SINGLE LOOP OPERATION (SLO)

ANF has performed analyses to support a decrease in the MCPR Safety Limit and MCPR LCO adjustment factors for SLO from 0.03 to 0.01.

These analyses are described in Attachment 2.

New LOCA-ECCS analyses have also been performed for SLO and the results support a change in the MAPLHGR SLO multiplier from 0.70 to 0.91 for all fuel types in the D~Cll core.

The previous multiplier had been based on the conservative application of earlier GE analyses of SLO.

The details of the LOCA-ECCS analysis and results for SLO are in Attachment 4.

IV.

EXTENDED LOAD LINE LIMIT ANALYSIS (ELLI.A)

ANF performed analyses to support operation above the 100\\ flow control line (or "load line") up to the APRM rod block line as part of the Dresden 3 Cycle 10 reload analyses.

The results of the previous analyses are applicable to Dresden 3 Cycle 11 since the cycle specific analyses for the new cycle have been performed consistently with respect to power/

flow region assumptions.

Therefore, operation in the expanded power/flow operating region of Figure 1.1 of Attachment 2 is supported.

V.

DESCRIPTION OF TECHNICAL SPECIFICATIONS/LICENSE Enclosure E provides Technical Specification changes to support D3Cll operation with a mixed core of ANF 8x8 and ANF 9x9 fuel consistent with the analyses provided in Attachments 1, 2,3, and 4.

The following sections outline the major areas requiring revisions and identifies the associated sections of the Technical Specifications.

A.

General Throughout the Technical Specifications and bases, references to Exxon Nuclear Company (ENC) have been changed to Advanced Nuclear Fuels Corporation (ANF), except in titles of documents.

Also, various

-. sections have been revised to reflect the appropriate ANF methodo-logies and to delete General Electric (GE) methods and references where appropriate.

All references to GE fuel, except those dealing with spent fuel storage, have been deleted, since no GE fuel will remain in the Dresden Unit 3 core.

For each revised specification identified below, the corresponding section in the bases has been revised as required.

B.

MCPR TS Section 3.5.K, 3.5.L 1.1.A, 3.5.L.3, (previously 3.5.K),

3.6.H.3.f.lv, 3.6.H.3.f.v 3.5.L (previously 3.5.K) 3.5.L.l (previously 3.5.K.l)

Description Sections 3.5.K and 3.5.L are changed to 3.5.L and 3.5.M, respectively, to accorrunodate insertion of new Section 3.5.K for transient LHGR (TLHGR) limits.

The following changes in Section 3.5.L are for the section that previously was "3.5.K".

In Single Loop Operation, the MCPR safety Limit and MCPRLCO adder is changed to 0.01 from 0.03.

MCPRLcO is changed to 1.39 for both 8x8 and 9x9 fuel.

Revised Figure 3.5-2 (Sheets 1, 2, and 3) to incorporate the following changes in reduced flow MCPR values:

For Sheet 1:

Total Core Flow

(\\ Rated) 100 90 80 70 60 50 40 MCPR Limit 8x8 9x9 1.10 1.15 1.21

1. 28.

1.36 1.46 1.60 1.09 1.14 1.20 1.26 1.34 1.44 1.57

• For Sheet 2:

Total Core Flow

(\\ Rated) 100 90 80 70 60 50 40 For Sheet 3:

Total Core Flow

(\\ Rated) 100 90 80 70 60 50 40

• Column headers are MCPR Operating Limits at rated flow.

C.

LHGR TS Section Description MCPR Operating Limit for 8x8 fuel*

1.28 1.32 1.36 1.28 1.32 1.36 1.31 1.35 1.39 1.34 1.39 1.43 1.39 1.44 1.48 1.45

1. 49.

1.54 1.52 1.56 1.61 1.66

1. 71

1. 76 MCPR Operating Limit for 9x9 fuel*

1.31 1.35 1.39 1.31 1.34

1. 36.

1.41 1.46 1.53 1.68 1.35 1.38 1.41 1.45 1.51 1.58

1. 73 1.39 1.42 1.45 1.49 1.55 1.62

1. 78 1.0 The definitions of the Fraction of Limiting Power Density (FLPD) and the Maximum Fraction of Limiting Power Density (MFLPD) are deleted and replaced by the definitions of the Steady State Linear Heat Generation Rate (SLHGR), the Fuel Design Limiting Ratio for Exxon Fuel (FDLRX), the Transient Linear Heat Generation Rate (TLHGR) and the Fuel Design Limiting Ratio for centerline Melt (FDLRC).

2.1.A.l, 2.1.B, Changed references to MFLPD, MFLPD/FRP, and Table 3.2.3 FRP/MFLPD to the indicated FDLRC or l/FDLRC.

3.1.A.2, 4.1.A.2.a 3.5.I Deleted Figure 3.5-1, sheets 3,4, and 5

3.5.J, 4.5.J 3.5.K (new)

D.

MAPLHGR TS Section Changed Section 3.5.J title to "LOCAL STEADY STATE LHGR."

Added references to FDLRX in these sections.

In addition, deleted GE LHGR design value of 13.4 KW/ft from Figure 3.5-lA, as well as replotting the same ANF limits for clarity and adding "steady state" to the title.

Added section on local transient LHGR and FDLRC.

Added Figure 3.5-lB, Transient LHGR Limit curve.

Description 3.5.I, 3.6.H.3.f.vi The SLO MAPLHGR multiplier is changed to 0.91 from 0.70.

o E.

Control Rod scram Time surveillance TS section 3.5.L (previously 3.5.K)

Description Deleted the paragraph that requires a MCPR penalty based on scram time performance.

F.

Relief Valve Out-of-Service TS Section 3.5.D, 4.5.D, 3.5.I, 3.6.H.3.f.vi Description Changed wording to allow extended operation with one relief valve out-of-service and limited (7 days) operation with two RVOOS provided HPCI is operable.and MAPLHGR adjustment factors are applied.

G.

Feedwater Heaters Out-of-Service Deleted the condition requiring a safety evaluation each time that off-normal feedwater heating is necessary during coastdown from Section 3.E in the Dresden 3 Facility Operating License (License No. DPR-25)

VI.

SUMMARY

The proposed Dresden 3 Cycle 11 MCPR operating limit (MCPRLco> at rated conditions has been conservatively chosen to bound all design basis abnormal operating transients and therefore precludes violations of the fuel cladding integrity safety limits for both ANF 9x9 and ANF 8x8 fuel.

The preceding discussions have addressed all major features of the ANF 9x9 XN-4 reload for Dresden 3 Cycle 11.

The mechanical, thermal hydraulic and neutronic design of the 9x9 XN-4L and XN-4H reload fuel have been analyzed to be compatible with the ANF 8x8 fuel design.

Thus, Commonwealth Edison concludes that operation of Dresden 3 Cycle 11 with an additional reload of ANF 9x9 fuel is safe and acceptable provided the following items are completed prior to start-up:

A.

The attached Technical Specifications are NRC approved; and B.

Dresden procedures are modified to be consistent with the revised Technical Specifications.

Based on the above discussion, the analyses in Attachments l, 2, 3 and 4, and the Significant Hazards Evaluation in Enclosure D, opera-tion of Dresden 3 Cycle 11 with ANF XN-4 reload 9x9 fuel does not represent an unreviewed safety question.

That is, for the Cycle 10 reload and associated Technical Specification changes:

a)

Neither the probability nor the consequences of an accident or malfunction of safety-related equipment, as previously evaluated in the safety analysis report, is increased.

b)

The possibility for an accident or malfunction of a different type than previously evaluated in the safety analysis report is not created.

c)

The margin of safety, as defined in the basis for any Technical Specification, is not reduced.

4321K

BNCLOSURB D SIGNIFICANT HAZARDS EVALUTION Commonwealth Edison proposes to amend License DPR-25 (Dresden Unit 3) to support the Cycle 11 core reload.

The proposed revisions include three basic types of changes; changes specific to the Cycle 11 reload fuel and analyses including new LOCA-ECCS analysis for Single Loop Operation, changes resulting from analyses performed to allow one relief valve out~of-service, and changes that are administrative or provide clarification.

Although a general description of the changes follows, more detailed discussion of the changes and their technical bases can be found in Enclosure c.

DESCRIPTION OF AMENDMENT REQUEST The proposed change in the Dresden Unit 3 License (DPR-25) is as follows:

1)

Deletion of the condition requiring a safety evaluation for coastdown operation with off-normal feedwater temperature from Section 3.E of the license.

The Technical Specification changes specific to the Cycle 11 reload fuel and analyses include:

2)

Revision of the Minimum Critical Power Ratio (MCPR) operating limit for Cycle 11.

3)

Reduction of the single Loop Operation (SLO) MCPR adder to 0.01 (from 0.03).

Also, a reduction in the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) reduction factor for SLO to 0.91 (from 0.70).

4)

Deletion of the paragraph that requires a MCPR penalty based on scram time performance.

5)

Incorporation of transient Linear Heat Generation Rate (TLHGR) limits.

6)

Revision of reduced flow MCPR limits.

The Technical Specification changes resulting from analyses*performed to allow equipment out-of-service include:

7)

Revision of the relief valve.Technical Specifications to require action only after two relief valves are found to be inoperable, provided HAPLHGR reduction factors are implemented.

Encl. D The Technical specification changes provided for clarification or as administrative changes include:

8)

Removal of all references to GE fuel, except in spent fuel storage Technical Specifications.

9)

Changing references to Exxon Nuclear Company (ENC) to Advanced Nuclear Fuels Corporation (ANF), except in titles of earlier documents and definitions of nuclear limits.

10)

Defining Transient LHGR (TLHGR), Steady State LHGR (SLHGR), LHGR, Fuel Design Limiting Ratio for Centerline melt (FDLRC) and Fuel Design Limiting Ratio for Exxon Fuel (FDLRX).

BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION commonwealth Edison has evaluated the proposed Technical Specifications and determined that they do not represent a significant hazards consideration.

Based on the criteria for defining a significant hazard established in 10CFR50.92(c), operation of Dresden Unit 3 Cycle 11 in accordance with the proposed changes will not:

A.

Involve a significant increase in the probability or consequences of an accident previously evaluated because:

1)

ANF has performed analyses with NRC approved methodologies to ensure that transients occurring under coastdown conditions with off-normal feedwater temperature are bounded by transients at rated conditions.

2)

The incorporation of the proposed MCPR operating limits noted above is explicitly provided to establish limits on reactor operation which ensure that the core is operated within the assumptions and initial conditions of the transient analyses. Operation within these limits

• will ensure that the consequences of a transient or accident remain within the results of the analyses.

The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are significantly affected and the MCPR safety limit continues to be protected.

3)

The incorporation of the proposed MCPR and MAPLHGR limits during single loop operation establishes limits on reactor operation to ensure thermal-mechanical integrity of the fuel and cladding.

Neither the consequences nor the probability of an accident is affected by this change because the design basis transients and accidents were considered when establishing these operating limits.

Encl. D 3 -

4)

The proposed MCPR operating limit is conservatively chosen such that it is not affected by scram time performance.

A MCPR penalty for scram time performance is not required because the most severe pressurization transient (load reject without bypass) is bounded even when the worst case scram performance penalty is applied.

Therefore, the proposed MCPR change does not adversely affect the consequences of previously evaluated events.

The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident have been significantly modified and the MCPR safety limit continues to be protected.

5)

The incorporation of the proposed TLHGR limits establishes limits on reactor operation to ensure thermal-mechanical integrity of the fuel under transient overpower conditions, consistent with the fuel vendor's design criteria and the surveillance method already performed in the onsite core monitoring computer software.

Consequences of previously evaluated events are therefore not affected.

The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are affected and the cladding integrity will be maintained during overpower events.

6)

The incorporation of the proposed reduced flow MCPR limits maintains limits on reactor operation to ensure that thermal limits will not be violated during transients initiated during off-rated core flows.

Therefore, consequences of postulated events are unaffected.

The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are affected and the MCPR safety limit will be protected during overpower events initiated at off-rated core flows.

7)

ANF has performed analyses with NRC approved methodologies to ensure that reactor limits are not violated during limiting transients and accidents with one relief valve out-of-service.

Event consequences are therefore not affected by this change.

The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are significantly affected.

8,9 These changes are administrative in nature and have no

&10) impact on any systems or limits on reactor operation.

B.

Create the possibility of a new or different kind of accident from any accident previously evaluated because:

1)

ANF has determined that transients occurring at off-rated feedwater heating during coastdown are bounded by those initiated at rated, full power conditions.

Furthermore, there is no impact or physical modification to systems or components whose failure could initiate a new or different kind of accident.

Encl. D 2,3, The proposed MCPR, MAPLHGR, and LHGR limits represent limits 4,5 on core power distribution which do not directly affect the

& 6) operation or function of any system or component.

As a result, there is no impact on or addition of any systems or equipment whose failure could initiate a new or different kind of accident.

7)

Operation is allowed with one RVOOS provided appropriate MAPLHGR reductions are implemented.

This change in no way impacts the function of the remaining operable valves or other equipment and since the appropriate requirements to test HPCI are included, this change does not create a new or different kind of accident.

8,9 These changes are administrative in nature and have no

&10) impact on or modification to any system or equipment whose failure could initiate an accident.

c.

Involve a significant reduction in the margin of safety because:

1)

The analysis supporting this change shows that transients during coastdown with off-normal feedwater temperature are bounded by transients at rated conditions, therefore no reduction in the margin to safety occurs.

2,3, These changes have been analyzed to demonstrate that the 4,5 consequences of transients or accidents are not increased,

& 6) using the specified restrictions, beyond those previously evaluated and accepted at Dresden.

The analyses show that the MCPR safety limit, fuel thermal-mechanical limits, and reactor pressure limits are not violated during postulated transients.

7)

The analysis supporting this change shows that the point of minimum MCPR occurs before any relief valves open, indicating that the assumption of one relief valve out-of-service will not reduce the margin to safety for anticipated abnormal operating transients.

For LOCA, analysis has shown that with the specific MAPLHGR restrictions, all criteria of 10 CFR 50.46 are satisfied for the limiting small break.

Large breaks are unaffected.

8,9 These changes are administrative in nature, either deleting

&10) information that is no longer applicable or providing clarification to current specifications

• Based on the above discussion, conunonwealth Edison concludes that the proposed amendments do not represent a significant hazards consideration.

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ATTACHMENT 5 DISCUSSION OF PREVIOUS SER TOPICS The following discusses several topics raised in previous generic and reload SERs for Dresden and is based on information provided by Advanced Nuclear Fuels.

A.

Rod Bow Considerations During the review of the previous reload submittal, the NRC SER placed an exposure cap on 8x8 and 9x9 fuel due to rod bow considerations.

The limit was set at 30,000 MWD/MR for 8x8 fuel and 23,000 MWD/MT for 9x9 fuel (batch average exposure).

Supplementary information has been submitted to

• the NRC and found to be adequate by the technical reviewer.

It is anticipated that an SER will be received which will increase the assembly average exposure limit to 35,000 MWD/MTU and 40,000 MWD/MTU for 8x8 and 9x9 fuel, respectively.

The projected peak assembly exposures at ECX::ll for the 8x8 and 9x9 assemblies are 34,400 and 23,500 MWD/MTU, respectively.

B.

Extended Burnup Approval Status (XN-NF-82-06)

QUalification of Exxon Nuclear Fuel for Extended Burnup (XN-NF-82-06(P)(A) w/supplements 2, 4 and 5, Revision 1) was approved in July of 1986.

Extended burnup qualification for 9x9 fuel was not sought until additional in reactor performance data was available.

The 9x9 document (Supplement l to 82-06) was submitted in January 1987 and the issuance of an SER is anticipated prior to the startup of Dresden 3 cycle 11.

c. Conditions on Critical Power Methods ((XN-NF-524, Rev. 1)

In the safety Evaluation prepared by the core Performance Branch covering Revision l to XN-NF-524(P)(A), "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors", the NRC Staff identified four conditions to be met during the application of the subject methodology under the generic approval granted by the SER.

The steps taken during the analysis to assure compliance with these conditions are described below.

CONDITION 1:

Each plant specific application must contain the data used to generate the uncertainties employed in the methodology.

The uncertainties used in the Dresden unit 3 MCPR safety limit calculations are the same as the uncertainties which have been used in previous Dresden analyses.

The two loop uncertainties are discussed below; the single 10op uncertainties are the same except as described in the cycle 11 reports.

ATT. 5 Plant measurement uncertainties which are not fuel-dependent were taken from approved NSSS supplier generic documents applied to Dresden.

As identified in XN-NF-524(P)(A), Revision 1, specific uncertainty values used in the analysis were a feedwater flow rate uncertainty of 1.76\\, a feedwater temperature uncertainty of 0.76\\, a core pressure uncertainty of 0.5\\, and a total core flow rate uncertainty of 2.5\\.

The generic core inlet temperature uncertainty of 2.0\\ was conservatively replaced with an uncertainty of 2.4\\ on the core inlet enthalpy.

These approved values were used as one-sigma uncertainties consistent with NE00-24011.

The nominal values, uncertainties, and statistical treatment of these measured plant parameters are swnmarized in Table 1.

The uncertainties associated with the XN-3 Critical Power Correlation are based on data contained in XN-NF-512(P)(A), Revision_!, and XN-NF-734(P)(A).

The safety limit analysis was based on a one-sigma uncertainty value of' 4.11\\ for the XN-3 correlation, consistent with the source documents noted above and with XN-NF-80-19(P)(A), Volume 4, Revision 1, which provides a generic description of the overall reload analysis.

The correlation statistics were developed from the XN-3 data base, which includes test geometries which encompass the Dresden 8x8 and 9x9 fuel designs.

XN-NF-734(P)(A) was issued explicity to validate the XN statistics for application to 9x9 fuel.

Power distribution measurement uncertainties are based on data contained in XN-NF-80-19(P)(A), Volume 1.

The safety limit calculation was based on one-sigma uncertainties of 5.28\\ on radial peaking factor and 2.46\\ on local peaking factor consistent with the reference report.

These uncertainties were developed based on analytical predictions of measured data for BWR fuel.

The same methods used for the analytical predictions were used for the nuclear design analyses for Dresden; hence the generic uncertainty values are applicable to Dresden.

The correlation and power distribution measurement uncertainties and their statistical treatment for the Dresden analysis are swnmarized in Table 2.

CONDITION 2:

All plant parameters that are not statistically convoluted must be placed at their limiting value.

In_ the performance of plant transient analyses, ANF uses design values for major process parameters for consistency with the FSAR analyses which are superseded by the ANF transient analyses.

Design values are established by the plant designer as conservative predictions of the boundaries of the plant operating envelope, and may not be accurate predictions of actual plant operation.

These values are used to assure a conservative calculation of the transient effects.

Nominal values are best-estimate predictions of plant operating conditions.

The use of nominal conditions is appropriate for the statistically treated parameters in the Monte Carlo analysis.

ATT. 5 3 -

Input to the Monte Carlo calculation consists of three major classifications of data:

heat balance information, power distribution information, and fuel geometric information.

Heat balance information consists of feedwater temperature and flow rate, core pressure and total flow rate, and core inlet enthalpy.

All of these variables are considered statistically in the Monte Carlo analysis.

Power distribution information is taken from the fuel management analysis and consists of radial, axial, and local peaking factors.

Radial and local peaking factors are considered statistically in the Monte Carlo analysis.

For power distributions characterized by bottom-peaked core average axial power shapes, a limiting center-peaked axial distribution is used.

Fuel geometric information consists of fuel dimensions and hydraulic demand curves.

Small variations in fuel dimensions within manufacturing tolerances are considered in the ANF pressure drop methodology and contribute to the flow distribution uncertainty.

Hydraulic demand curves are used to determine fuel assembly flow rates as a function of bundle power; indivudual assembly flow rates are treated statistically in the Monte Carlo analysis.

CONDITION 3:

Each application should demonstrate that the uncertainties in plant parameters are treated with at least a 95\\

probability at a 95\\ confidence level in accordance with Acceptance Criterion 1.0 of Standard Review Plan section 4.4.

The magnitude and nature of the uncertainties used in the Monte Carlo analysis have been established generically during the Staff review of ANF topical report XN-NF-524(P)(A), Revision 1, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors".

A detailed review of the XN-3 correlation statistics was included in the review of ANF topical report XN-NF-512(P) (A), Revision 1, "The XN-3 Correlation".

A detailed review of power distribution measurement uncertainties was included in the review of ANF topical report XN-NF-80...'..19(P)(A), Volume 1, "Neutronics Methods for Design and Analysis".

The conclusion that these uncertainties may be conservatively treated as normally distributed was addressed during the generic review.

Uncertainties in the measurement of plant parameters were taken from the NSSS supplier's generic reload submittal.

Based on the Staff's approval of these uncertainties for use in the MCPR safety limit calculation, the ANF analyses used the published values as 95\\ confidence statistics.

Process measurement uncertainties are generally characterized by a normal distribution; therefore, a normal distribution was used in the ANF analysis.

ATT. 5 The Monte Carlo analysis was performed to demonstrate that during sustained operation at the MCPR safety limit, at least 99.9\\ of the fuel rods in the core would be expected to avoid boiling transition at a confidence level of 95\\.

This conclusion conservatively assures that the boiling transition limitation will be protected during anticipated operational occurrences in which the MCPR safety limit is protected.

The referenced Standar~ Review Plan section identifies this method as an acceptable approach to the 95/95 treatment of uncertainties.

CONDITION 4:

Each application must present a goodness-of-fit analysis for the fitting of the Pearson curve in order to assure that the number of Monte Carlo trials used in establishing the safety limit MCPR are sufficient.

In the original ANF MCPR safety limit methodology, the first four statistical moments of the Monte Carlo output were used to define an output frequency distirbution through fitting of Pearson functions.

This approach was taken to minimize the number of trials necessary in the Monte Carlo analysis.

Revision 1 to XN~NF-524 abandoned this approach in favor of a distribution-independent method of assigning tolerance limits.

The new approach required a larger number of Monte Carlo trials, but the end result was a conclusion which was independent of the Pearson functions.

Since the statistical analysis involved no fitting of standard functions to the Monte Carlo output, no goodness-of-fit analysis was provided.

In the case of the Dresden analysis, 500 Monte Carlo trials were performed.

In the non-parametric tables, an expected value may be established at a confidence level of 95\\ with as few as 50 trials.

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TABLE l PLANT MEASUREMENT UNCERTAINTIES Nominal Uncertainty Parameter Units Value

\\ Nominal Treatment Feedwater Flowrate Mlbm/hr 12.41*

l. 76 Convoluted Feedwater Temperature deg F 340.l 0.76 Convoluted Core Pressure psia 1035 0.50 Convoluted Total Core Flow Mlbm/hr 98.0 2.50 Convoluted Core Inlet Temperature 2.00 Replaced by core inlet enthalpy Core Inlet Enthalpy Btu/lbm 522.3 2.40 convoluted Core Power MW 3200*

Allowed to

• vary with heat balance

• Feedwater f lowrate and core power were increased above design values to attain desired core MCPR for safety limit evaluation, consistent with XN-NF-524(P)(A), Revision l.

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TABLE 2 FUEL-RELATED UNCERTAINTIES Source Uncertainty Statistical Parameter Document

\\ Nominal Treatment XN-3-Correlation XN-NF-512(P} (A}

4.11 Convoluted XN-NF-734(P}(A}

Radial Peaking Factor XN-NF-80-19(P}(A}

5.28 Convoluted Volume l Local Peaking Factor XN-NF-80-19(P}(A}

2.46 Convoluted Volume l Axial Peaking Factor XN-NF-80-19(P}(A}

2.99 Limiting Volume l Value Assembly Flowrate XN-FN-79-59(P}(A}

2.80 convoluted 4321K

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