ML20043B036

Amend 116 to License DPR-36,modifying Tech Specs to Reflect Operating Limits for Cycle 12 Core Reload

ML20043B036
Person / Time
Site:	Maine Yankee
Issue date:	05/17/1990
From:	Wessman R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20043B034	List: ML20043B033 ML20043B036 ML20043B037
References
NUDOCS 9005240106
Download: ML20043B036 (17)

Text

- -

+j je UNITED STATES g

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p, NUCLEAR REGULATORY COMMISSION

j W ASHINGT ON, D. C. 20$$$

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MAINE YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-309 i

MAINE YANKEE ATOMIC POWER STATION

(

AMENDMENT TO FACILITY OPERATING LICENSE i

Amendment No.116 License No. DPR-36 t

1.

The Nuclear Regulatory Comission (the.Comission or the NRC) has found that:

A.

The application.for amendment filed by the Maine Yankee Atomic Power Company (the licensee) dated January 16, 1990 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 t

Comission; C.

There is reasonable assurance: (i) 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 Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the comon

. 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 L

the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.B.6.(b) of Facility Operating License No. DPR-36 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.116, are hereby incorporated in the license.

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

9005240106 900517 ADOCK0500g}OP DR

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3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION-w~.

Richard H. !!assman, Director l

Project Directorate I-3 Division of Reactor Projects-I/II Office of Nuclear. Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

May 17, 1990 e

i 5

e I

i

ATTACHMENT TO llCENSE AMENDMENT NO.116 FACit1TY OPERATING LICENSE NO. OPR-36 DOCKET NO. 50-309 Replace the following pages of the Appendix A Technical Specifications with the attached pages..The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert 2.1-1 2.1-1 2.1-4 2.1-4 2.1-5 2.1-5 2.2-1 2.2-1 3.10-1 3.10-1 3.10-2 3.10-2 3.10-3 3.10-3 3.10-4 3.10-4 3.10-5 3.10-5 3.10-7 3.10-7 3.10-12 3.10-12 3.10-13 3.10-13 3.10-15 3.10-15 3.10-19 3.10-19 1

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2.1 LIMITING SAFETY SYSTEM SETTING - REACTOR PROTECTION SYSTEM Apolicability Applies-to reactor trip settings and bypasses for the instrument channels monitoring the process variables which influence the safe operation of the pl ant.

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 Qs, + BTc + C, or 1835 psig (whichever is larger).

Where cold leg temperature, *F Te

=

2053.2-

\\

A 17.9 B

-10053.0

{

C

=

A x QR Qa.

i 3

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

3 3

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 AmendmentNo.4!/E,EE,TE,//,

JB, EE,173,116

4 l'*

WHERE: 0

= A

• QR DNB 1

1 TRIP AND PVAR = 2053.2 0DNB+ T7.9TC 10053.0 T

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1.35-I l

1.30 J '

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-0.1 0.0 0.1 0.2 0.3 0.4 0.5 Excere Symmetric Offset Y, = A*((U-L)/(U+L))+B MAINE YANKEE Thermol Mergin/ Low Pressure Trip Setpcint Figure Technical Part 1 2.1-1c Spscification (A versus Y) 2.1-4 Amendment No. 29, JE, AD, #3, EE, /#. /E, EE, //4, i 16

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1 TRIP i

- 10053.0 AND P

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0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermol Power MAINE Y ANKEE Th'ermal Margin / Low Pressure Figure Technical Trip Setpoint Port 2 2.1-1b Specificction (CR)versus Fraction of Roted Thermal Power) 2.1-5

. Amendment No. 29, 3E, /0, 48, EE, EB, 74, 78, EE, JJ3,116

I t

e 2.'2 SAFETY LIMITS - REACTOR CORE Anolicability Applies to the limiting combinations of reactor power, and Reactor Coolant System flow, temperature, and pressure during operation.

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

Soecifications A.

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

DNBR - 1.20 using the YAEC-1 DNB heat flux correlation 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:

Fuel Type LHGR Limit, kw!ft EDE LOL M

20.8 20.0 P

21.1 20.0 0

22.1 20.6 R

23.5 22.4 where the LHGR limit for each fuel type decreases linearly with Cycle Average Burnup (CAB), and the EOC 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 I

greater than the coolant saturation temperature. The upper boundary of the I

nucleate boiling regime is termed " Departure from Nucleate Boiling" (DNB).

l 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 Amendment No. 79, EE, 74, 7$, $E, SE, J07, JJ3,116

E a

3.10 CEA GROUP. POWER DISTRIBUTION. MODERATOR TEMPERATURE COEFFICIENT LIMITS AND 4

COOLANT COND1T10NS Aeolicability:

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

L Qbiective:

To insure (1) core suberiticality 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:

A.

CEA Operational Limits

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

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 SA and SB.

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 group by more than 10 steps.

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

The indicated subgroup positions of Subgroup 5A and 5B differ by more than 15 steps.

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

t 1.

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 j-be within the limits specified in 3.10.C.2 using the latest unrodded radial peaking factor.

l 3.10-1 Amendment No, M, EE, EE, 116

r 1.

L

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

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 />.

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 that cannot be driven shall be assumed not able to be tripped until it is proven that it can be tripped. Operation with an inoperable CEA is permitted provided:

a.

The shutdown margin specified in 3.10.8.1 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 clow 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.

l

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3.10-2 Amendment No. 2E, 4, EE, (E, PE,107,116 l

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• 2.

The total radial peaking factor, defined as 1

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:

a

)I Pa-Pc )*+

(Pb-Pd)'

T, - 2 z

(Pa+Pb + Pc+Pd)'

where Pi is the relative quadrant power determined from the incore system for quadrant i, when the incore system is

operable, if the incore system is not operable, the Pi are the 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 LC0 (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.

E:

3.

Be in at least HOT SHUTDOWN.

s 3.10-3 l

Amendnant No. #, 6, EE, EE, /E, PE, Jp7,116 l

b q

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c

+

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,44

-3. Lincore detector alarms shall be set at least weekly.

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

rate limit defined in Specification 3.10.C.1.

If four or more ecoincident alarms are received, the validity of the alarms shall be immediately determined, and, if valid, power shall be immediately 1

?

' decreased below the alarm setpoint.

3.1.

If the:in' core monitoring system becomes inoperable, perform one of;the following within 4 effective full power hours:

1 ~1nitiate 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:

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

81 for symmetric offset between 0.00 and 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 boat rate permitted by Specification 3 '.,. t, c. u' rate last measured corrected to.

L LHR (measured) <

.m-

00% r w t.

The CEAs shall be tmtsined c.,,ve the 100% power dependent insertion limit anu,: -trir offset shall be monitored once per shift to ensure that it remains within the above range.

+

L 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 LC0 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.

4 3.10-4 Amendment flo, /$, pp, f>h /E,PE,JS/,JD7,116 s

f

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4.2L If the measured value of Tc 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:

i-

1. The thermal power levf1 A nstricted to less than or equal to 20% of rated po d

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

5. The incore detector' system shall be used to confirm powor 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 i

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

'c 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

'l.

shown in Figure 3.10-9.

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 Amendment No. 3E, 40, 6E, EE, 7E, 707,116 n.

b

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under either the more conservative excore symmetric offset LCO envelope or at a

>ower level consistent with maintaining an appropriate margin to the peak linear 3-1 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 1

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

'.g 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 LC0 on the radial power 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 l

limit on a monthly basis. Should the measured steady-state unrodded total L

radial peak including uncertainties exceed the limit of Figure' 3.10-4 l

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', (measured)/F', (Figure 3.10-4)]

l-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 l-LC0 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 I

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 L

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 RPS LSSS and LOCA analyses remain valid.

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

The L

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 Amendment No. 38, M, EE, EE, 78, 116

(.

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NOTE: 1. THIS CURVE INCLUDES 10% CALCULATIONAL UNCERTAINTY

2. F

=F X 1.0 3 R

T

3. MEA 5URED F SHOULD'BE AUGMENTED BY MEASUREMENT-R UNCERTAINTY (8%) BEFORE COMPARISON TO THIS CURVE.

1.79 1

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R (0.00,1.747)

(0.50,1.747)

(1.00,1.742)

-(2.00,1.743) 1.77 (4,00,1,741)

[s,co,1,733) 18.00,1.723)

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9 10 11 12 13 14 15 CYCLE AVERAGE EXPOSURE (KMWD/MT)

MAINE YANKEE Allowoble Unrodded Rcdial Peck Versus Figure Technical Cycle Average Burnup 3.10 - 4 Specificctier Amendment No. 116 l

3.10 - 12

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NOTE: CEA's are molntoined at or above 100% power insertion limit when applying 3.10.C.2.2.2 110

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10 12 14 16 18 20 22-34 26 as T

Allowable % increase in F ( bove Figure 3.10-4)

R

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

amenament No. es, AE, H M, M. EE, /E,

,g,JJ3,116

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1. J l I

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7.5 7.0 4.

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t 3,3 t.

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l il P = 20 l'

lij 1

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2.0 SDM = 5.20 - 0.00350C + 0.0200P l

when C is less then 400 PPM lil 1.5 SOM = 3.80 + 0.0200P l

h-when C is greater then or equel to 400 PPM

'l c

i,0 where i

I SDM is the required shutdown rnorgin in percent reactivity j

C is the RCS boron concentration in PPM l

0.5 l:

t----r-P is the power levelin percent of rated power i

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0.0 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 0

100 i

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ACTUAL RCS BORON CONCENTRATION (PPM)

MAINE YANKEE Required Shutdown Morgin Figure l

Technicc!

Versus 3.10 -7 Specification RCS Baron Concentration Amendment No. 40/ /E, J07, 116 0-2

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18 17 UNACCEPTABLE OPERATION d

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o.Qs COORDINATES

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( 0,16.0) zg6

( 6 5, 14.5) 5

( 7 3, 14.3) to 4

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

9 MAINE YANKEE Linear Heat Generation Rote (LHGR) Limits Figure Techn.ical Versus 3.1 0 - 11 Specification Core Height wi.enwaen i. iiu. Jid7, ild, ii6 3.10 -19

-