ML20237G636

Amend 6 to License DPR-21,revising Current Min Critical Power Ratio,Maplhgr to Reflect Cycle 12 Operation

ML20237G636
Person / Time
Site:	Millstone
Issue date:	08/06/1987
From:	Thomas C Office of Nuclear Reactor Regulation
To:
Shared Package
ML20237G625	List: ML20237G623 ML20237G636 ML20237G648
References
NUDOCS 8708240166
Download: ML20237G636 (17)

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[ o g UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION

$ WASHINGTON, D. C. 20555

\.....);E NORTHEAST NUCLEAR ENERGY COMPANY

, DOCKET N0. 50-245 MILLSTONE NUCLEAR POWER STATION, UNIT N0. 1 l

AMENDMENT TO FACILITY OPERATING LICENSE cet ho bP-21

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by the Northeast Nuclear Energy Company, (the licensee) dated May 21, 1987, as supported by submittals dated June 30 and July 14, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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2. Accordingly, the license is amended by. changes to the Technical j Specifications as indicated in the attachment to this license 1

. amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-21.is hereby amended to read as follows:

(2) Technical Specifications (The Technical Specifications contained in Appendix A, as revised 3through Amendment No. 6 , are hereby incorporated in the license. The licensee shall. operate the facility in accordance j with the Technical Specifications. .-

3. This' license amendment is effective as of the date of its issuance. :j FOR THE NUCLEAR REGULATORY COMMISSION .]

i 0nL0, Q$mW '

Cecil 0. Thomas, Director Integrated Safety Assessment Project Directorate Division of Reactor Projects - III, IV, V and Special Projects

Attachment:

Charges to the Technical i Specifications Date of Issuance: August 6, 1987 l l

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ATTACHMENT TO LICENSE AMENDMENT N0. 6 FACILITY OPTRATING LICENSE N0. DPR-21 l

DOCKET NO. 50-245 Revise Append'ix A Technical Specifications by removing the pages indentified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

F.EMOVE INSERT 1-1* 1-1*

1-2 1-2 2-3* 2-3*

i 2-4 2-4 2-5 2-5 2-6* 2-6*

3/4 11-1 3/4 11-1 3/4 11-f 3/4 11-2 3/4 11-3 3/4 11-3 3/4 11 4 3/4 11-4 3/4 11 5 3/4 11-5 I

3/4 11-6* 3/4 11-6*

3 3/4 11-7 3/4 11-7 ,

1 B 3/4 11-2 B 3/4 11-2 l i 1 l

00venleaf page provided to maintain document completeness. No changes conttined on these pages.

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.. c .i i.0 l DEFINITIONS f.

j- The succeeding frequently used terms are explicitly defined. sol that a I . uniform interpretation,of the Specifications may be achieved, y

A. . Fire Suppression Water System ,.

A FI,RE' SUPPRESSION. WATER SYSTEM shall consist of: a water source (s);.

gravity. tank (s) or pump (s); and distribution piping with associated sectionalizing control or isolation valves. ~Such valves shall-include yard hydrant curb valves, and the first valve ahead of the water flow alarm device; on each sprinkler,' hose standpipe, or spray system riser.

B. Alterat' ion of the Reactor Core The act of moving any component in the ' region above the core support plate, below the upper grid and within the. shroud with the exception of normal control rod motion.

C, Hot Sta'ndby HOT. STANDBY means operation with the reactor critical, system pressure less.than 600 psig, and the main steam isolation valves closed.

D. Immediate IMMEDIATE means that the required action will be initiated as soon as practicable considering the safe operation of the' unit and the importance of the required' action.~

E. Instrument Calibration l

An INSTRUMENT CALIBRATION means the adjustment of an instrument signal output so that it corresponds, within acceptable range, accuracy and response time, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the i I

entire instrument including actuation, alarm or trip.

F. Instrument Functional Test An INSTRUMENT FUNCTIONAL TEST means'the injection of a simulated f signal into the instrument primary sensor to verify the proper instrument channel response, alarm, and/or initiating action.

G. Instrument Check An INSTRUMENT CHECK is qualitative determination of operability by l observ'ation of behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

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q Millstone Unit 1 1-1 j l I l

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l 1.0 H. Minimum Critical Power Ratio (MCPR) )

Minimum Critical Power Ratio (MCPR) is the value of critical power ratio associated with the most limiting assembly in the reactor core. ]

Critical Power Ratio (CPR) is the ratio of that power in a fuel 1

, assembly, which is calculated by application of the GEXL correlation I to cause some point in the assembly to experience boiling transition, to the actual assembly operating power.  ;

I. Mode .!

• b The reactor mode is that which is established by the mode-selector- {

switch.

J. Operable - Operability l A system, subsystem, train, component or device shall be OPERABLE or I

have OPERABILITY when it is capable of performing its specified function (s). Implicit 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 l function (s) are also capable of performing their related'aupport l function (s).

K. Operating Operating means that a system or component is performing its intended function in its required manner.

L. Operating Cycle Interval between the end of one refueling outage and the end of the  !

next subsequent refueling outage.

M. Fraction of Limiting Power Density l

The ratio of the linear heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type. The design LHGR is 13.4 kW/ft for BP8x8B (GE-7B) fuel bundles and 14.4 kW/ft for GE8x8EB (GE-BB) fuel bundles.

MaximumFractionof(,imitingPowerDensity The Maximum Fraction of Limiting Power Density (MFLPD) is the highest v'alue existing in 'the core of the Fraction of Limiting Power Density (FLPD).

N. P'rimary Containment Integrity Primary containment integrity means that the drywell and pressure suppression chambgr are intact and all of the following conditions are satisfied.

su m ad===.h is. 3 b2 Amendment No. 6

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l SAFETY LIMITS

.i i, 2.1.1 FUEL CLADDING INTEGRITY B. When the reactor pressure is less than or equal to 800 psia or reactor flow is 1.ess than 10% of design, the reactor thermal power transferred to the coolant shall'not exceed 25% of rated.

C. 1. To assure that the Limiting 3afety System Settings established in Specifications 2.1.2A and 2.1.2B are not exceeded, each required scram shall be initiated by its primary source signal. The Safety Limit shall be assumed to be exceeded when scram is accomplished by a means other than the Primary Source Signal.

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 2.1.2A and a control rod scram does not occur. _

D. Whenever the reactor is in the cold 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 fuel when it is seated in the core. This level shall be continuously monitored.

LIMITING SAFETY SYSTEM SETTINGS 2.1.2.A.1.a. where:

S = Setting in percent of rated thermal power (2011 MWt)

W= Total recirculation flow in percent of design. See Note (1) l The trip setting shall not exceed 90 percent of rated power during generator load rejections from an initial generator ,

power greater than 307 MWe, The APRM scram setdown shall be 90% of rated within 30 seconds after initiation of full load rejection.

b. In the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:

5 5 (0.58 W + 62) r FRP

'MFLPD) p where, FRP = fraction of rated thermal power (2011 MWt)

! Note (1) Design flog to be defined as the recirculation flow (not to exceed 33.48 x 10 lbs/hr.) needed to achieve 100% core flow.

Millstone Unit 1 2-3

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LIMITING SAFETY SYSTEM SETTINGS (Continued) 2.1.2 FUEL CLADDING INTEGRITY i

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A.1.b. MFLPD z maximum fraction of limiting' power density where the limiting power density is 13 4 kW/ft for BP8x8R (GE-7B)

. fuel bundles and 14.4 kW/ft

• for GE8x8EB (GE-8B) fuel bundles.

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the j" design value of 1.0, in which case the actual operating value will be used.

c. During power ascensions with power levels less than or equal to 90%, APRM Flux Scram Trip Setting adjustments may be made as described below, provided that the change in scram setting adjustment is less than 10%

and, a notice of the adjustment is posted on the reactor control panel.

The APRM meter indication is adjusted by: J ARPRM p

_FRP_

where:

APRM = APRM Meter Indication

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P z 5 Core Thermal Power For no combination of loop recirculation flow rate and core theraal power shall the APRM flux scram trip setting be allowed to exceed 120% of RATED THERMAL POWER.

2. APRM Reduced Flux Trip Setting (Refuel or Startup/ Hot Standby Mode)

When the mode switch is in the REFUEL ar STARTUP/ HOT STANDBY position, the APRM scram shall be setdown to less than or equal to 15% of RATED THERMAL POWER. The IRM scram trip setting shall l

not exceed' 120/125 of full sr. ale.

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2-4 Amendment No. 6 m i

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• LIMITING SAFETY SYSTEM SETTINGS (Continued) 2.1.2 FUEL CLADDING INTEGRITY B. 1. AFRH Rod Block Trip Setting

a. The APRM rod block trip setting shall be as shown in Figure 2.1.2 and shall be: (Run Mode)-

SRB I +'

where:

S = Rod block setting in percent RB of rated thermal power (2011 MWt).

W z Total recirculation flow in percent of design. (Note 1, Page 2-3).

b. In the event of operation with a maximum fraction limiting power density (MPLPD) greater than the fraction of rated power (FRP),

the setting shall be modified as follows:

S RB g (0.5BW + 50) FRP MFLPD j where:

I FRP = fraction of rated thermal power (2011 MWt)

MFLPD maximum fraction of limiting power density where the~

limiting power density is

, 13.4 kW/ft for BP8x8R (GE-7B)

, fuel bundles and 14.4 kW/ft I

for GEBx8EB (GE-8B) fuel bundles.

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.

c. During power ascensions with power levels less than or.

t equal to 90%, APRM Rod Block Trip Setting adjustments may be made as described below, provided that the change in scram setting adjustment is less than 10% and a notice of the adjustment is posted on the reactor cor. trol' panel:

1tinsteam mit.1 2-5 Amendment No. 6 i

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j LIMITING SAFETY SYSTEM SETTINGS (Continued)

'2.1.2 FUEL CLADDING INTEGRITY B .1. c. The APRM meter indication is adjusted by: i APRM = [Hp Dj p where: )

APRM = APRM Meter Indication P' =  % Core Thermal Power

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2. The APRM rod block trip setting for the refuel and startup/ hot standby mode shall be less than or equal to 12% rated thermal power. l

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C. The Reactor Low Water Level Scram trip setting shall be greater than or 6

equal to 127 inches above the top of the active fuel.

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D. The Reactor Low Low Water Level ECCS initiation trip point shall not be greater than 83 inches nor less than 79 inches.

1 E The Turbine Stop Valve Scram ~ trip setting shall be less than or equal to '

ten percent valve closure from full open. .

F. The Turbine Control Valve Fast Closure Scram shall trip upon actuation of ,

the acceleration re'ay in conjunction with failure cf selected bypass I valves to start opening within 280 milliseconds.

The maximum setting of the time delay relays which bypass this scram shall be 280 milliseconds. 1 G. The Main Steam Isolation Valve Closure Scram trip settings shall be less than or equal to ten percent valve closure from full'open.

H. The Main Steam Line Low Pressure trip, which initiates main steam line  !

isolation valve closure, shall be greater than or equal to 825 psig.

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LIMITING CONDITI0d FOR OPERATION 3.11 REACTOR FUEL ASSEMBLY ,

h Applicability L ,

The Limiting Conditions for Operation associated with .the fuel rods apply- to those parameters which monitor the fuel rod operating conditions.

Objective, The Objective of the Limiting Conditions for Operation is to assure the performance of the fuel rods.

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

1. During power operation, the APLHGR,.i.e., the LCO, for each type.of fuel as a function of axial location and average planar exposure, shall not exceed limits based on applicable APLHGR limit values that have been approved.for the respective fuel and lattice types, as determined by the approved methodology described in~GESTAR II. (This approval is based on and limited to the GESTAR II methodology.) If

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hand calculations are required, the APLHGR for each type of fuel as a function of average planar exposure shall not exceed the limiting value for the most limiting lattice :(excluding natural U) shown in Figure 3 11.1.

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2. If at any time during operation it is determined by normal surveillance that the limiting value for APLHGR specified in Section 3.11.A.1 is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If I the APLHGR 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.

SURVEILLANCE REQUIREMENT , 4.11 REACTOR FUEL ASSEMBLY Applicability 1 The Surveillance Requirements apply to the parameters which monitor the fuel rod operating' conditions.

l Objective The Objective of Surveillance Requirements is to specify the type and frequency '

of surveillance to be applied to the fuel rods. i Millstone Unit 1 3/4 11-1 knendment No. 6 i

i e LIMITING CONDITION FOR OPERATION (continued) i Specifications A. Average Planar Linear Heat Generation Rate (APLBGR) te APLHGR for each type of fuel, as a function of average planar exposure shall be determined daily during reactor operation at 125% RATED THERMAL POWER.

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t LIMITING CONDITION FOR OPERATION 1

3 11 REACTOR FUEL ASSEMBLY B. Linear Heat Generation Rate (LHGR) I During steady state power operation, the linear heat generation' rate (LHGft) of any rod in any fuel assembly at any axial location shall not exceed the maximum allowable LHGR of 13.4 kW/ft for BP8x8R (GE-7B) fuel i bundles and 14.4 kW/ft for GE8x8EB (GE-8B) fuel bundles.

During power operation, the LHGR shall not exceed the limiting value. If at any time during operation it is determined, by normal surveillance, that the limiting value for LHGR is being exceeded,_ action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits 3 within two (2) hours, the reactor shall be brought to COLD SHUTDOWN l

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. ]

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SURVEILLANCE REQUIREMENT .

4.11 REACTOR FUEL ASSEMBLY l

B. Linear Heat Generation Rate (LHGR) l The LHCR shall be checked daily during reactor operation at 8255 RATED THERMAL POWER. '

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e n=*== autt.1 3/4 11- 5 Amendment No. 6

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LIMITING CONDITION FOR OPERATION 3 11 REACTOR FUEL ASSEMBLY C. Minimum Critical Power Ratio (MCPR)

During power operation, HCPR shall be as shown in Table 3 11.1. If at any time during operation it is determined by normal surveillance that the ligfting value 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 />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

For core flows other than rated, the MCPRs in Table 311.1 shall be multiplied by K f, where Kf is as shown in Figure 3.11.2.

D. If any of the limiting values identified in Specifications 311. A B, or C, are exceeded, even if corrective action is taken, as prescribed, a Reportable Occurrence report shall be submitted.

SURVEILLANCE REQUIREMENT 4.11 REACTOR FUEL ASSEMBLY C. Minimum Critical Power Ratio (MCPR)

1. MCPR shall be determined daily during reactor power operation at

> 25% RATED THERMAL POWER and following any change in power level or distribution that would cause operation with a limiting control rod pattern as described in the bases for specification 3.3.B.5.

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2. Utilization of Option B Operating limit MCPR values requires the scram time testir.g of 15 or more control rods on a rotating basis every 120 operating days.

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M111=+a== amit 1 3/4 11. 6

l TABLE 3.11.1 OPERATING LIMIT MCPRS FOR CYCLE 12 (OPTION B)

BOC 12 TO EOC EOC 12 TO 705 COASTDOW FUEL TYPE i 1.34 , 1.34 BP8 x 8R (GE-78) 1 34 1.34 GE8 x BEB (GE-8B)

OPERATING LIMIT HCPR'S FOR CYCLE 12 (OPTION A)

BOC 12 TO EOC EOC 12 TO 705 COASTDOW FUEL TYPE 1.39 1 39 BPB x 8R (GE-78) 1.39 1 39 GE8 x BEB (GE-8B) I 1

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Minstame sait 1 3/4 m7 Amendment No. 6

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f 3 11 REACTOR FUEL ASSEMBLY j i

BASES )

! I C. The st'eady state value for MCPR was selected to provide a margin to l accomodate transients and uncertainties in monitoring the core operating- l state as well as uncertainties in the critical power correlation itself.

This value ensures that:

1.[ For the initial conditions of the LOCA analysis, a HCPR of 1.18 is ,

sa tisfied. For the low flow ECCS analysis, an initial MCPR of 1.24 j is assumed, and

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2. For any of the special transients, or disturbances, caused by single operator error or single equipment malfunction the value of MCPR is conservatively assumed to exist prior to the initiation of the transient or disturbance.

At core thermal power levels 4 25%, the reactor will be operating at l minimum recirculation pump speed, and moderator void content will be very '

! small. For all designated control rod patterns which may be employed at this power, thermal hydraulic analysis indicates that the resultant MCPR value is in excess of requirements. 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 calculation of MCPR at greater than 25% RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

, The use of the Option B operating limit MCPR requires additional SCRAM l time testing and verification in accordance with GE letter, A. D. Vaughn l to R. M. Matheny, April 21, 1987, regarding Potential Technical l Specification Changes for Implementation of Advanced Methods.

D. Reporting Requirements The LCO's associated with monitoring the fuel rod operating conditions are required to be met at all times or corrected to within the limiting values of MAPLHGR, LHGR, and MCPR within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the time the plant is determined to be exceeding them. It is a requirement, as stated in Specifications 3.11. A, B, and C, that if at any time during power operation it is determined that the limiting values for MAPLHGR, LHGR, or MCPR are exceeded, action is then initiated to restore operation to within the prescribed limits. This action is to be initiated within 15 minutes if normal surveillance indicates that an operating limit has been reached.

Each event involving operation beyond a specified limit shall be logged

, , and a reportable occurrence issued. It must be recognized that there is always an action which would return any of the parameters (MAPLHGR, LHGR, or MCPR) to within prescribed limits, namely power reduction. Under most circumstances, this will not be the only alternative.

e mme== 1tnit.1 33/411-2 Amendment No. 6 s

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