ML19354D845

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Proposed Tech Specs,Reflecting Changes to Operating Parameters for Cycle 12
ML19354D845
Person / Time
Site: Maine Yankee
Issue date: 01/16/1990
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Maine Yankee
To:
Shared Package
ML19354D844 List:
References
NUDOCS 9001220256
Download: ML19354D845 (20)
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!.e ATTACHMENT B i.

Summary Description of Cycle 12 Technical Specification Changes CDF9005.LTR 9001220256 900116 3

PDR ADOCK 05000309 P

PDC

Item No.

Technical Specification Description of Chance Reason for Chance 1.

2.1.1.b a) TM/LP trip A-a) Reflects Cycle 12 pages 2.1-1,.2.1-4 and coefficient modified power distributions 2.1-5 and RPS setpoints.

b) Figure 2.1-la b) Reflects Cycle 12 -

modified power distributions and RPS setpoints.

c) Figure 2.1-Ib c) Reflects Cycle 12 modified power distributions and RPS setpoints.

2.

2.2 a) Steady-state peak a) Reflects Cycle 12 linear heat rates Specified Acceptable modified Fuel Design Limits (SAFDL) for prevention of noterline melting (Section 3.2.2 of YAEC-1713).

3.

3.10 a) Technical Specification a) Corrects a typo-pages 3.10-1 to 3.10-5, 3.10.A.3 modified graphical error.

3.10-7, 3.10-8, 3.10-12, 3.10-13, 3.10-15, and b) Technical Specification b) Reflects 3.10-19 3.10.C.1 conservative application i

of axial fuel densification and thermal expansion factor to all fuel' types.

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Item No.

-Technical Specification Descriotion of Chance Reason for Chance 3.

3.10 continued c). Technical Specification c) Simplifies actions 3.10.C.2.2.2 modified for this Technical Specification.

d) Technical Specification d) Reflects Cycle 12 3.10.C.3.1.1 modified, power distributions and RPS setpoints.

e) Technical Specification e) Clarifies wording in 3.10.C.4.2.1 and this Technical 3.10.C.4.2.2 modified Specification.

f) Technical Specification

'f) Corrects a typo-3.10.C.7 modified graphical error.

g) Basis on page 3.10-7 g) Corrects a. typo-modified graphical error and changes basis wording to be consistent with the change to Technical i

Specification i

3.10.C.2.2.2.

l h) Figure 3.10-4 modified h) Reflects Cycle 12 radial l

peaking (Section 4.3 of YAEC-1713).

i) Figure 3.10-5 modified i) Reflects Cycle 12 power

' distributions and RPS setpoints.

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. Item No.

Technical Specification-

- Description of Chance Reason for Chanee -

3.

3.10 continued-

~

j) ' Figure 3.10-7 modified

' j) Reflects increase.in

. shutdown margin required by Cycle 12 analyses (Section 5 of--

YAEC-1713).1

. k) Figure 3.10-11 modified k) Reflects Cycle-12 LOCA analysis results (Section 5 of YAEC-1713).

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ATTACHMENT C Proposed Technical Specification Chanaes I

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f 2.1 LIMITING SAFETY SYSTEM SETTING - REACTOR PROTECT 10_N_.$1$,1 W Aeolicability Applies to reactor trip settings and bypasses for the instrument channels monitoring the process variables which influence the safe operation of the plant.

Ob.iective To provide automatic protective action in the event that'the process variables approach a safety limit.

Specification The Reactor Protective System trip setting limits and bypasses for the required operable instrument channels shall be as follows:

2.1.1 Core Protection-a) Variable Nuclear Overpower:

Less than or equal to Q + 10, or 106.5 (whichever is smaller) for Q greater-than or equal to 10 and less than or equal to 100, and less than'or equal to 20 for Q 1ess than or equal to 10.

Where Q = percent thermal or nuclear power, whichever is larger, b) Thermal Margin / Low Pressure:

Greater than or equal to: A 03. + BTc + C, or 1835 psig (whichever is larger).

Where cold leg temperature

  • F Tc 2053.2

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A 17.9 B

-10053.0 C

=

A x QR Qm.

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A and QR are given in Figures 2.1-la and 2.1-lb, respectively.

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This trip may be bypassed below 10% of rated power.

c)

The symmetric offset trip function shall not exceed the limits shown in Figure 2.1-2 for three loop operation. This trip may be bypassed below 15% of rated power.

2.1-1 CDF9005.LTR 1

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Excore Symmetric Offset Y, = A*((U-L)/(U+L))+B MAINE YANKEE Thermal Margin / Low Pressure Trip Setpoint Figure Technical Part 1 2.1-1a Specification (A versus Y;)

2.1-4

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Technical Trip Setpoint Port 2 2.1-1b l

Specification (QR msus Fracdon of 3

Roted Thermal Power) 2.1-5

,m 2.2 SAFETY LIMITS - REACTOR CORE Apolicability l

Applies to.the limiting combinations of reactor power, and Reactor Coolant System flow, temperature, and pressure during operation.

Ob.iective To maintain the integrity of the fuel cladding and prevent the release of significant amounts of fission products to the reactor coolant.

Specifications A.

The reactor and the Reactor Protection System shall be operated such that the following Specified Atceptable fuel Design Limit (SAFDL) on the departure from nucleate boiling heat flux ratio (DNBR) is not exceeded during normal operation and anticipated operational occurrences, i

i DNBR = 1.20 using the YAEC-1 DNB heat flux correlation p

B.

The reactor and the Reactor Protection System shall be operated such that the following SAFDLs for prevention of fuel centerline melting are not exceeded.during normal operation and anticipated operational occurrences.

A steady-state peak linear heat generation rate (LHGR) equal to:

l Fuel Tvpl LHGR Limit. kw/ft E0L f.0L M.

20.8 20.0

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21.1 20.0

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t -

Q 22.1 20.6

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R 23.5 22.4

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where the LHGR limit for each fuel type decreases linearly with l

Cycle Average Burnup (CAB), and the E0C Burnup for the purposes of establishing a linear relationship is 14,500 MWD /MTV CAB.

Basis To maintain the. integrity of the fuel cladding, thus preventing fission product release to the Primary system, it is necessary to prevent overheating of the cladding. This is accomplished by operating within the nucleate boiling regime of heat transfer, and with a peak linear heat rate that will not cause fuel centerline melting in any fuel rod.

First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed " Departure from Nucleate Boiling" (DNB).

At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperature and the possibility of cladding failure.

2.2-1 1.

l CDF9005.LTR l

3.10' CEA GROUP. POWER DISTRIBUTION. MODERATOR TEMPERATURE COEFFICIENT llMITS AND COOLANT CONDITIONS Aeolicabilitv:

Applies to insertion of CEA groups and peak linear heat rate during operation.

' Ob.iect ive : -

l To insure (1) core subcriticality after a reactor trip, (2) limited potential reactivity insertions from a hypothetical CEA ejection, and (3) an acceptable core power distribution, moderator temperature coefficient, core inlet temperature, and reactor coolant system pressure during power operation.

Specification

  • I i

A.

CEA Operational Limits

1. When the reactor is critical., except for physics tests and CEA l

exercises, the shutdown CEAs (Groups A, B, and C) shall be fully withdrawn and the regulating CEAs (Groups 1 through 5) shall be no further inserted than the limits shown in Figure 3.10-1 for 3 loop operation.

CEA Group 5 consists of two subgroups designated Subgroup 5A and 58.

2. A CEA is considered fully withdrawn if the CEA is withdrawn to 4 steps or less from its upper electrical limit.
3. Except during physics testing, a CEA misalignment is considered to be any one of the following:

A CEA in Group A, B, C, 1, 2, 3, or 4 that is out of position from the remainder of the grouo by more than 10 steps.

A CEA in Subgroup 5A or SB that is out of position from the remainder of the subgroup by more than 10 steps.

The indicated subgroup positions of Subgroup SA and SB differ l

by more than 15 steps.

If a CEA misalignment is not corrected within 15 minutes, L

operation with a CEA misalignment is permitted for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provided:

a. Thermal power is reduced by at least 10% of rated power within one-half hour by at least 20% of rated power within one hour of identification of the misalignment.

The CEA insertion limits specified for the initial thermal power must be maintained.

b. Within two hours after realignment, the peak linear heat rate will be shown to be within the limits specified in 3.10.C.1 and the total radial peaking factor will be shown to be within the limits specified in 3.10.C.2 using the latest

)

unrodded radial peaking factor.

3.10-1 i

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4. If the CEA deviation alarms from both the computer pulse counting system and the reed switch indication system are not available, individual CEA J

positions shall be logged and misalignment checked every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, i

5. Operation of the CEA's in the automatic mode is not permitted.

B.

Shutdown Margin Limits

1. When the reactor is critical, the shutdown margin will not be less than that shown in Figure 3.10-7, except during low power physics tests when the shutdown margin will not be less than 2% in reactivity.
2. A trippable CEA is considered inoperable if it cannot be tripped.

A CEA o

L that cannot be driven shall be assumed not able to be tripped until it is f

proven.that it can be tripped. Operation with an inoperable CFA is permitted provided:

a.

The shutdown margin specified in 3.10.B.1 is satisfied without the reactivity associated with the inoperable CEA within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of identification of the inoperable CEA.

b.

Except for low power physics tests and CEA exercises, only one CEA is inoperable.

3. A trippable CEA is considered to be a slow CEA if the drop time from de-energizing its holding coil to reaching 90% of its full insertion exceeds 2.7 seconds at operating temperature and 3 pump flow.

Operation with a slow CEA is permitted provided:

a.

The shutdown margin specified in 3.10.B.1 is satisfied without 1.5 times the reactivity associated with the slow CEA after 2.5 seconds of drop time.

C.

Power Distribution Limits 1.

The peak linear heat rate with appropriate consideration of normal flux peaking, measurement-calculational uncertainty (8%), engineering factor (3%), increase in linear heat rate due to axial fuel densification and thermal expansion (0.3%), and power measurement

)

uncertainty (2%) shall not exceed the limits shown in Figure 3.10-11 as a function of core height.

Should any of these limits be exceeded, immediate action will be taken to restore the linear heat rate to within the appropriate limit specified in Figure 3.10-11, 3.10-2 CDF9005.LTR

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2.

The total radial peaking factor, defined as i

1 P

F - F. (1 + T,)

shall be evaluated at least once a month during power operation above 50%

of rated full power.

2.1 Fl is the latest available unrodded radial peak determined from the incore monitoring system for a condition where all CEAs are at or above the 100% power insertion limit.

T, is given by the following expression:

)f Pa-Pc)'+

(Pb-Pd)'

T, = 2 (Pa+Pb + Pc4Pd)*

where Pi is the relative quadrant power determined from the incore system for quadrant i, when the incore system is o perable, if the incore system is not operable, the Pi are tie signals from excore detector channels 2.2 If the measured value of F' exceeds the value given in Figure 3.10-4, perform one of the following within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

1.

Reduce the allowable PDIL insertion (Figure 3.10-1) symmetric offset LCO (Figures 3.10-8 and 3.10-9) and trip band (Figure 2.1-2), thermal margin low pressure trip limit (Figures 2.1-1 a and b and Technical Specification 2.1), linear heat rate limits (Figure 3.10-11) and excore LOCA monitoring limits (Figures 3.10-2 and 3.10-3) by a factor greater than or equal to:

[F' measured] / [F' figure 3.10-4)

E:

2.

Reduce thermal power at a rate of at least 1%/ hour to bring the combination of thermal power and % increase in F', to within the limits of Figure 3.10-5, while maintaining CEAs at or above the 100% power insertion limit.

Reduce the linear heat rate limits (Figure 3.10-11) by the allowable percent increase in F',

]

corresponding to 100% power in Figure 3.10-5.

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E:

3.

Be in at least HOT SHUTDOWN.

3.10-3 l-CDF9005.LTR

i 3.

Incore detector alarms shall be set at least weekly.

Alarms will be based on the latest power distribution obtained, so that the linear heat rate does not exceed the linear heat i

rate limit defined in Specification 3.10.C.1.

If four or more i

l coincident alarms are received, the validity of the alarms shall be t

immediately determined, and, if valid, power shall be immediately decreased below the alarm setpoint.

3.1 If the incore monitoring system becomes inoperable, perform one of the following within 4 effective full power hours:

1. Initiate a power reduction at a rate of at least 1% per hour to a power level less than or equal to the power level given by the following expression for the limiting location:

P = [R-0.8 S) (LHR (limit)/ LHR (measured)), where:

)

l P = % of rated power, R = 75 for symmetric offset between +0.05 and 0.10, or

)

81 for symmetric offset between 0.00 cnd 0.05, or

]

96 for symmetric offset between 0.00 and -0.05, or

)

90 for symmetric offset between -0.05 and -0.10.

)

S = Number of steps the CEAs deviate from the CEA position existing when the linear heat rate measurement was taken.

LHR (limit) - Linear heat rate permitted by Specification 3.10.C.1, and LHR (measured)

Linear heat rate last measured corrected to 100% power.

The CEAs shall be maintained above the 100% power dependent insertion limit and symmetric offset shall be monitored once per shift to ensure that it remains within the above range.

This method may be used for up to 14 effective full power days from the time when the linear heat rate measurement was taken; or

2. Comply with the LCO given in Figure 3.10-2 while maintaining the CEAs above the 100% power insertion limit.

If a power reduction is required, reduce power at a rate of at least 1%

power hour; or

3. Comply with the LC0 in Figure 3.10-3.

If a power reduction is required, reduce power at a rate of at least 1% per hour.

3.10-4 CDF9005.LTR

i 1

4.2 If the measured value of T, is greater than 0.10, operation may proceed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as long as F. is maintained within the provisions of Specification 3.10.C.2.

Subsequent operation for the purpose of measurement and to identify the cause of the tilt is allowable provided

1. The thermal power level is restricted to less than or equal to 20% of rated power

)

2. Operation is in accordance with Specification 3.10.C.2.2.

]

5. The incore detector system shall be used to confirm power distribution, such that the peaking assumed in the safety analysis is not exceeded, after initial fuel loading and after each fuel reloading, prior to operation of the plant at 50% of rated power.
6. If the core is operating above 50% of rated power with an excore nuclear channel out of service, then the azimuthal power tilt shall be determined once per shift by at least one of the following means:
a. Neutron detectors (at least 2 locations per quadrant),
b. Core-exit thermocouples (at least 2 thermocouples per quadrant).
7. Whenever the reactor is operating above 20% of rated power the excore symmetric offset shall be within the bounds for symmetric offset LCO shown in Figure 3.10-8.

When the turbine is operating in the IMPIN control mode, the excore symmetric offset shall be within the bounds for symmetric offset LC0

)

shown in Figure 3.10-9.

l D.

Moderator Temperature Coefficient (MTC):

[

Except during low power physics testing the MTC shall be less positive than that shown in Figure 3.10-10.

E.

Coolant Conditions

1. Except for low power physics testing, the reactor coolant pressure and the reactor coolant temperature at the inlet to the reactor vessel shall be maintained within the limits of Figure 3.10-6 during steady-state operation whenever the reactor is critical.
2. Except for low power physics testing, the reactor coolant flow rate shall be maintained at or more than a nominal value of 360,000 gpm during steady-state operation whenever the reactor is critical.

Exception: The requirements of 3.10.E.2 may be modified during initial testing to permit power levels not to exceed 10% of rated power with three loops operating on natural circulation.

3.10-5 CDF9005.LTR

under either the more conservative excore symmetric offset LCO envelope or at a sower level consistent with maintaining an appropriate margin to the peak linear 1 eat rate assumed in the LOCA.

Both these functions ensure that operation is within the limiting peak linear heat rates assumed as initial conditions for the loss of Coolant Accident (LOCA).

Further, since rod position information is not available to this excore system, this function assumes the most limiting radial power distributions permitted at each power level.

The split excore detectors monitor the axial component of the power distribution.

The signal generated from the excore detectors is provided as input to both the Symmetric Offset and Thermal Margin / Low Pressure Trip Systems.

Limiting Safety System Settings (LSSS) are, therefore, generated as a function of the excore detector response.

The radial component of the power distribution is monitored as a limiting Condition of Operation (LCO) by Technical Specification 3.10.C.2.

The

)

intent of the Specification is to monitor the radial component of the power distribution and to ensure that assumptions made in the generation of Reactor Protective System (RPS) LSSS remain valid. The LCO on the radial powar distribution is specified in Figure 3.10-4 in the form of a steady

-state unrodded total radial peak (F', and provides indication that the core power distribution is behaving as predicted.

Figure 3.10-4 includes 10% for calculation uncertainties.

The measured steady-state value of F'., augmented by 8% for measurement uncertainty, is compared to this limit on a monthly basis.

Should the measured steady-state unrodded total radial peak including uncertainties exceed the limit of Figure 3.10-4 at any time in the cycle, specific action is to be taken to assure that the LSSS remain valid. The specific action includes a) the reduction of RPS LSSS and LCO by the ratio of [F'a (measured)/F',, (Figure 3.10-4)]

to directly compensate for the higher radial peaks, or b) the imposition of additional restrictions on power and CEA position (Figure 3.10-5) to assure that the assumptions made in establishing he RPS LSSS and LCO remain valid.

Figure 3.10-5 in conjunction with restricted CEA insertion allows for an increase in the steady-state unrodded total radial peak above the limits of Figure 3.10-4 without a modification of the RPS LSSS.

The allowed increase in radial peak is derived from the

]

difference between the radial peaks assumed in the RPS setpoints for rodded conditions at reduced power and the radial peaks reflected in the CEA insertion limit at 100% power. To accommodate the increased radial peaking, LOCA linear heat generation rate limits are decreased

]

by the allowable radial peaking increase at 100% power in Figure 3.10-5.

]

This assures that the radial peaking factors versus power assumed in the

)

RFS LSSS and LOCA analyses remain valid.

]

The power distribution in the core can be determined in two ways.

The normal method is through analysis of the fixed and movable neutron detector signals with the on-line computer. The alternative is to determine the radial and axial peaking factors by hand.

The radial peaking factor can be determined from the core exit thermocouples, the 3.10-7 CDF9005.LTR

+.

l NOTE: 1. THIS CURVE INCLUDES 107. C/.LCULATIONAL UNCERTAINTY

2. F

=F X 1.0 3 R

3. MEASURED F SHOULD BE AUGMEnlTED BY MEASUREMENT R

UNCERTAINTY (87.) BEFORE COMPARISON TO THIS CURVE.

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MAINE YANKEE Allowable Unrodded Radiol Peck Versus Figure Technical Cycle Average Burnup 3.10 -4 Specificotion 3.10 - 12

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NOTE: CEA's are molntoined at or above 100% power Insertion limit wheft opplying 3.10.C.2.2.2 i

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MAINE YANKEE Allowable Power Level vs. Increase in Figure Technical Total Radial Peck 3.10 - 5 Specification 3.10 -13

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100 200 300 400 500 ACTUAL RCS BORON CONCENTRATION (PPM)

MAINE YANKEE Required Shutdown Margin Figure Technical Versus 3.10 -7 Specification RCS Boron Concentration 3.10 -15

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l Lu 0, 16.0)'

zg' 6

( 65, 14.5) 5

( 73, 14.3) _

b4

( 85, 13.0)

L I

3

( 93, 11.9) h2 (10 0, 9.2)

--~

Z

--}

i t

l 0

L 0

10 20 30 40 50 60 70 80 90 10 0 CORE HEIGHT (%)

l.

MAINE YANMEE Linear Heat Generation Rote (LHGR) Limits Figure Techn.ical Versus 3.1 0 - 11 Spec.ficat. ion Core Height l

i 3.10 - 19 m

I o

7s' i

f ATTACHMENT D Cycle 12 Core Performance Analysis Report Q '.

.CDF9005.LTR 4

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