ML20206N644

Amend 170 to License NPF-49,modifying TS 3/4.2.2 to Be IAW NRC-approved Westinghouse Methodologies for Heat Flux Hot Channel Factor - Fq(Z)

ML20206N644
Person / Time
Site:	Millstone
Issue date:	05/10/1999
From:	Clifford J NRC (Affiliation Not Assigned)
To:
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ML20206N650	List: ML20206N644 ML20206N659
References
NUDOCS 9905180024
Download: ML20206N644 (12)

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.. ja " coq UNITED STATES f

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 is...../

NORTHEAST NUCLEAR ENERGY COMPANY. ET AL.

DOCKET NO. 50-423 MILLSTONE NUCLEAR POWER STATION. UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.170 License No. NPF-49

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

A. The application for a;.sendment by Northeast Nuclear Energy Company, et al. (the licensee) dated.'.inuary 18,1999, complies with the standards and requirements of the Atomic Energe Act of 1954, as amended (the Act), and the Commission's rules and regulation.,et forth in 10 CFR Chapter I; B. The facility will opcste in conformity with the applichtion, the provisions of the Act, and the rules and regulations of the Commission; 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 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.

i 9905190024 990510 1

PDR ADOCK-05000423 j

PDR p

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-49 is hereby amended to read as follows:

(2) TechnicalSoecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.170, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION J mes W. Clifford, Chief, Section 2 Project Directorate l Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: Ny 10,1999 i

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ATTACHMENT TO LICENSE AMENDMENT NO.170 l

FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO, 50-423

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Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

I Remove Insert

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3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2 7 3/4 2-8 3/4 2-8 3/4 2-10 3/4 2-10 3/4 2-11 3/4 2-11 6 20 6-20 6-20a 6-20a B 3/4 2-4 B 3/4 2-4 L '.

P' WER ' DISTRIBUTION LIMJji 0

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F fZ) e 1

FOUR LOOPS OPERATING LINITING CONDITION FOR OPERATION 3.2.2.1 Fo(Z) shall be limited by the following relationships:

p*RTP F (Z) S K(Z) for P > 0.5 a

P p*are F (Z) s K(Z) for P s 0.5 o

0.5 F$" = the Fo limit at RATED THERMAL POWER (RTP) provided in the core operating limits report (COLR).

MRMM M Where:

P=

RATED THERMAL POWER, and K(Z) - the normalized F (Z) as a function of core height o

specified in the COLR.

APPLICABILITY:

MODE 1.

ACTION:

I With Fo(Z) exceeding its limit:

I For RA0C' operation with Specification 4.2.2.1.2.b not being satisfied a.

or for base load operation with Specification 4.2.2.1.4.b not being satisfied:

(1)

Reduce THERMAL POWER at least 1% for each 1% Fa(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip setpoints have been reduced at least 1% for each 1% Fo(Z) exceeds the limit, and NILLSTONE - UNIT 3 3/4 2-5 Amendment No. Jip, pp. 77, //p,170 0607

)

POWER DISTRIBUTION LINITS LINITING CONDITION FOR OPERATION (Continued) l (2)

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by item (1) above; THERMAL POWER may then be increased provided Fo(Z) is demonstrated through incore mapping to be within its limits.

b.

For RAOC operation with Specification 4.2.2.1.2.c not being satisfied, one of the following actions shall be taken:

(1) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the CORE OPERATING LIMITS REPORT by at least 1% AFD for each percent F (Z) j o

exceeds its limits.

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm I

setpoints to these modified limits, or

)

4 (2) Verify that the requirements of Specification 4.2.2.1.3 for base load operation are satisfied and enter base load operation.

I Where it is necessary to calculate the percent that Fo(Z) exceeds the limits for item (1) above, it shall be calculated as the maximum percent over the core height (Z),

consistent with Specification 4.2.2.1.2.f, that f (Z) exceeds its limit by the j

o following expression:

0

-1 x 100 for P > 0. 5 Fo" x K(Z) p Lj

~1 x 100 for P.4 0. 5 0.5 For base load operation with Specification 4.2.2.1.4.c not being c.

satisfied, one of the following actions shall be taken:

(1)

Place the core in an equilibrium condition where the limit in 4.2.2.1.4.c is satisfied, and remeasure F "(Z), or o

NILLSTONE - UNIT 3 3/4 2-6 Amendment No. pp, J79,170 0607 J

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POWER DISTRIBUTION LINITS LINITING CONDITION FOR OPERATION (Continued)

(2) Reduce THERMAL POWER at least 1% for each 1% Fo(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds its limit o

shall be calculated as the maximum percent over the core height (Z), consistent with Specification 4.2.2.1.4.f, by the following expression:

i

"' ~1 x 100 for P a APL ^*

x K(2) p SURVEILLANCE REQUIRENENTS 4.2.2.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.1.2 For RAOC operation, Fo(Z) shall be evaluated to determine if F (Z) is within its limit by:

o Using the movable incore detectors to obtain a power distribution map a.

at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.

Evaluate the computed heat flux hot channel factor by performing both of the following:

(1) Determine the computed heat flux hot channel Factor, F "(Z) by o

increasing the measured Fo(Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5%

to account for measurement uncertainties, and (2) Verify that F "(Z) satisfies the requirements of Specification o

j 3.2.2.1 for all core plane regions, i.e. 0-100% inclusive.

1 i

MILLSTONE - UNIT 3 3/4 2-7 Amendment No. 59, #9 77, 170 I

0607 l

i

PbWERDISTRIBUTIONLINITS SURVEILLANCE REQUIRENENTS (Continued) '

Satisfying the following relationship:

c.

Ff(Z) s F*""' x K(Z) for P > 0.5 PxW(Z)

F;(Z) S F*""' x K(Z) for P S 0. 5

~\\

W(Z) x 0.5 where Fy(Z) is the measured Fo(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, FS" is the Fa limit, K(Z) is the normalized Fa(Z) as a function of core height, P is the relative THERMAL POWER, and W(Z) is the cycle-dependent function that accounts for power distribution transients encountered during normal operation.

FS"", K(Z), and W(Z) are specified in the CORE OPERATING LIMITS REPORT as per Specification 6.9.1.6.

d.

Measuring F"a(Z) according to the following schedule:

(1) Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWt1 at which Fo(Z) was last determined,* or (2) At least once per 31 Effective Full Power Days, whichever occurs I

first.

4 e.

With the maximum value of l

.i F;(Z)

K(2) i j

over the core height (Z) increasing since the previous determination of Fy(Z), either of the following actions shall be taken:

(1)

Increase Fo"(Z) by an appropriate factor specified in the COLR and 4

verify that this value satisfies the relationship in Specification 4.2.2.1.2.c, or i

During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map outlined.

NILLSTONE - UNIT 3 3/4 2-8 Amendment No. pp, JP, 97, JEP,170 0607

POWER DISTRIBUTION LIMITS

. SURVEILLANCE REQUIREMENTS (Continued) b.

-During base load operation, if the THERMAL POWER is decreased below APL" then the conditions of 4.2.2.1.3.a shall be satisfied before reentering base load operation.

j 4.2.2.1.4 During base load operation F (Z) shall be evaluated to determine if o

i Fo(Z) is within.its limit by:

{

l Using the movable incore detectors to obtain a power distribution a.

map at any THERMAL POWER above APL".

b.

Evaluate the computed heat flux hot channel factor by performing both of the following:

(1) Determine the computed heat flux hot channel factor, Fo"(Z), by increasing the measured F "(Z) component of the power o

distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties, and (2) Verify that Fo"(Z) satisfies the requirements of Specification 3.2.2.1 for all core plane regions, i.e., 0 - 100% inclusive.

c.

Satisfying the following relationship:

i Ff(Z) s F*" x K(Z) for P > APL""

\\

P x W(Z),

where: Fy(Z) is the measured F (Z) increased by the allowances o

formanufacturingtolerancesandmeasurementuncertainty,FS'"

is the Fo limit, K(Z) is the normalized F (Z) as a function of o

core height, P is the relative THERMAL POWER, and W(Z)g is the cycle-dependent function that accounts for limited power distribution transients encountered during base load operation.

FSTP K(Z), and W(2)g are specified in the COLR as per Specification 6.9.1.6.

d.

Measuring Fs(Z) in conjunction with target flux difference determi-nation according to the following schedule:

(1)

Prior to entering base load operation after satisfying Sec-tion 4.2.2.1.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermal power having been maintained above APL" for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to mapping, and (2) At least once per 31 Effective Full Power Dajs.

MILLSTONE - UNIT 3 3/4 2-10 Amendment No. Jip, JP, #7,170 0607

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) e.

With the maximum value of Ff(Z)

K(2) 1 over the core height (Z) increasing since the previous determination of Fy(2), either of the following actions shall be taken:

(1)

Increase Fo"(Z) by appropriate factor specified in the COLR and verify that this value satisfies the relationship in Specification 4.2.2.1.4.c, or (2)

Ff(Z) shall be measured at least once per 7 Effective Full Power Days until 2 successive. maps indicate that the maximum value of F;(Z)

K(2) over the core height (Z) is not increasing.

f.

The limits specified in 4.2.2.1.4.c and 4.2.2.1.4.e are not applicable in the following core plane regions:

-(l)

Lower core region 0% to 15%, inclusive.

(2)

Upper core region 85% to 100%, inclusive.

4.2.2.1.5 When Fo(Z) is measured for reasons other than meeting the require-ments~ of Specifications 4.2.2.1.2 or 4.2.2.1.4, an overall measured F (Z) shall o

be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, l

l MILLSTONE - UNIT 3 3/4 2-11 Amendment No. pp, pp. 77, Jgp,170 l

0607

ADMINISTRATIVE CONTROLS CORE OPERATING LINITS REPORT (Cont.)

2.

Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, 3.

. Control Rod Insertion Lir.its for Specification 3/4.1.3.6, 4.

_ Axial Flux Difference Limits, target band, and APL"" for Specifica-tions 3/4.2.1.1 and 3/4.2.1.2, 5.

Heat Flux Hot Channel Factor, K(z), W(z), APL", and W(z)a for Specifications 3/4.2.2.1 and 3/4.2.2.2.

'6.

Nuclear Enthalpy Rise Hot Channel Factor, Power Factor Multiplier

for Specification 3/4.2.3.

'7.

ShutdownMarginMonitorminimumcountrateforSpecification3/4.3.5.l 6.9.1.6.b The' analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1.

WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"

July 1985 (W Proprietary). (Methodology for 5pecifications 3.1.1.3--

Moderator Temperature Coefficient, 3.1.3.5--Shutdown Bank Insertion Limit, 3.1.3.6--Control Bank Insertion Limits, 3.2.1--Axial Flux Difference, 3.2.2--Heat Flux Hot Channel factor, 3.2.3--Nuclear Enthalpy Rise Hot Channel Factor.)

l 2.-

T. M. Anderson to'K. Kniel (Chief of Core _ Performance Branch, NRC),

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January 31, 1980--Atta.hment: Operation and Safety-Analysis Aspects of an_ Improved Load Follow Package.

3.

NUREG-800, Standard Review Plan, U'.S. Nuclear Regulatory Commission, l

Section 4.3, Nuclear Design, July 1981 Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC),

Revision 2, July 1981.

4.

WCAP-10216-P-A-RIA, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," Rev. 1, February 1994 (W Proprietary).

(Methodology for Specifications 3.2.1--Axial Flux Difference,(Relaxed Axial Offset Control] and 3.2.2--Heat Flux Hot Channel Factor [W(z) surveillance requirements for Fo Methodology].)

5.

WCAP-9561-P-A, ADD. 3, Rev. 1,."BART A-1: A COMPUTER CODE FOR THE l

-BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS--SPECIAL REPORT:

THIMBLE MODELING W ECCS EVALUATION MODEL," July 1986 (W Proprie-tary).

(Methodology for Specification 3.2.2--Heat Flux Hot Channel factor.)

{

6..

WCAP-10266-P-A, Addendum 1, Rev. 2-P-A, "THE 1981 VERSION OF THE WESTINGH0USE ECCS EVALUATION MODEL USING THE BASH CODE," March 1987 (W Proprietary).

(Methodology for Specification' 3.2.2--Heat Flux Hot Channel Factor.)

MILLSTONE - UNIT 3 6-20 Amendment No. 7J, 77, #7, 77,' pf,170

oeoe

179,

ADMINISTRAT>?E CONTROLS I

GORE OPERATING LINITS REPORT (Cont.)

7.

WCAP-11946, " Safety Evaluation Supporting a More Negative E0L Moderator Temperature Coefficient Technical Specification for the l

Millstone Nuclear Power Station Unit 3," September 1988 (W Proprie-tary).

8.

WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL.17 l

USING THE NOTRUMP CODE," August 1985 (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

9.

WCAP-10079-P-A, "NOTRUMP - A N0DAL TRANSIENT SMALL BREAK AND GENERAL l

NETWORK CODE," August 1985 (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

10.

WCAP-12610, " VANTAGE + Fuel Assembly Report," June 1990 (W l

Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

11.

Letter from V. L. Rooney (USNRC) to J. F. Opeka, " Safety Evaluation for Topical Report, NUSCO-152, Addendum 4, ' Physics Methodology for PWR Reload Design,' TAC No. M91815," July 8, 1995.

12.

Letter from E. J. Mroczka to the USNRC, " Proposed Changes to Technical Specifications, Cycle 4 Reload Submittal - Boron Dilution Analysis,"

B13678, December 4, 1990, 13.

Letter from D. H. Jaffe (USNRC) to E. J. Mroczka, " Issuance of Amendment (TAC No. 77924)," March 11, 1991, 14.

Letter from M. H. Brothers to the USNRC, " Proposed Revision to Technical Specification, Shutdown Margin Requirements and Shutdown Margin Monitor Operability for Modes 3, 4, and 5 (PTSCR 3-16-97),

B16447, May 9, 1997.

15.

Letter from J. W. Anderson (USNRC) to M. L. Bowling (NNECO), " Issuance of Amendment - Millstone Nuclear Power Station, Unit No. 3 (TAC No.

M98699)," October 21, 1998.

MIgLSTONE-UNIT 3 6-20a Amendment No. pf,170

POWER DISTRIBUTION LIMITS RAW %

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FL0tl RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

Margin is maintained between the safety analysis limit DNBR and the design limit DNBR.

This margin is more than sufficient to offset any rod bow penalty and transition core penalty.

The remaining margin is available for plant design flexibility.

When an Fa measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.

The heat flux hot channel factor, F (Z), is measured periodically using the o

incore detector system. These measurements are generally taken with the core at or near steady state conditions.

Using the measured three dimensional power distributions, it is possible to derive Fo"(Z), a computed value of Fa(Z).

However, because this value represents a steady state condition, it does not include the variations in the value of F (Z) that are present during a

nonequilibrium situations.

To account for these possible variations, the steady state limit of Fa(Z) is adjusted by an elevation dependent factor appropriate to either RA0C or base load operation, W(Z) or W(Z)a, that accounts for the calculated worst case transient conditions.

The W(Z) and W(Z)a, factors described above for normal operation are specified in the COLR per Specification 6.9.1.6.

Core monitoring l

and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion. Evaluation of the steady state Fo(Z) limit is j

performed in ' Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation nonequilibrium limits are performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c.

)

i When RCS flow rate and F"m are measured, no additional allowances are l

necessary prior to comparison with the limits of the Limiting Condition for Operation. Measurement errors of 2.4% for four loop flow and 2.8% for three loop flow for RCS total flow rate and 4% for F"m have been allowed for in determination of the design DNBR value.

The measurement error for RCS total flow rate is based upon performing.a precision heat balance and using the result to calibrate the RCS flow rate indicators.

Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner.

Therefore, a penalty of 0.1% fer undetected fouling of the feedwater venturi will be added if venturis are not inspected and cleaned at least once for 18 months.

Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending i

various plant performance parameters.

If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

MILLSTONE - UNIT 3 B 3/4 2-4 Amendment No. M, g,170 0609

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