L-76-418, Transmittal of Request to Amend Appendix a of Facility Operating License

Transmittal of Request to Amend Appendix a of Facility Operating License

ML18227D847
Person / Time
Site:	Turkey Point
Issue date:	12/09/1976
From:	Robert E. Uhrig Florida Power & Light Co
To:	Stello V Office of Nuclear Reactor Regulation
References
L-76-418
Download: ML18227D847 (27)

Text

NRC FORM 195 U.S. NUCLCAR REGULATORY COMh<ISSION DOCKET NUMBER V

{2.76) I 50-251 FILE NUMBER NRC DISTRIBUTION FOR PART 60 DOCI<ET MATERIAL TO: FROM: OATE OF OOCUMENT Mr. Victor Stello Florida Power & Light Company 12/9/76 Miami, Florida OATE RECEIVED Mr. Robert E. Uhrig 12/9/76 RLETTE R RNOTOR IZE 0 PROP INPUT FORM NUMBER OF COPIES RECEIVED

/@ORIGINAL QUN C LASS I F I E O three signed DCOP Y 40 copies encl recvd.

OESCQIPTION ~ ENCLOSU RE Ltr. notorized 12/9/76.... trans the following: Amdt. to ol/change to Appendix A tech specs

....submitted as a result of a re-evaluation of ECCS cooling performance.

(S-P) 'ACKNOWLEDGED PLANT NAME: DP NOT REMOVE .

Turkey Point Unit 04 SAFETY FOR ACTION/INFORMATION 12 10 76 ASSXGNED PD:

N Lear PR ECT MAN E Elliott PROJECT lfANAGER'IC L C ASST Parrish ASST lfcGough INTERNAL D I ST R I BUT ION REG F SYSTEMS SAFETY PLANT SYSTEMS S TE SAFE NRC PDR HEXNEMAN TEDESCO I&E SCHROEDER OELD GOSSXCK & STAFF ENGINEERXNG XPPOLXTO MXPC ERNST CASE KNIGHT HANAUER SIHMEIL OPERATING REACTORS SPANGLER HARLESS PAWL CK STELLO PROJECT MANAGEMENT REACTOR SAFE OPERATING TECH GAlfMILL BOYD ROSS EISENHUT STEPP PE COLLINS NOVAK HOUSTON ROSZTOCZY PETERSON CHECK BUTLE SITE ANALYSIS.

MELTZ VOLLMER HELTEMES AT&I BUNCH SKOVHOLT SALTZlfAN J COLLINS RUTBERG KREGER EXTERNAL DISTRIBUTION CONTROL NUiiilBE R LPDR ~ Miami Fla. NAT LAB B 00 TIC: REG V,XE ULRXKSON OR NSIC: LA PDR ASLB: CONSULTANTS:

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NRC FORM 196 (2 76)

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~+ gc g FLORIDA POWER & LIGHT COMPANY December 9, 1976 L-76-418 Office of Nuclear Reactor Regulation Attention: Mr. Victor Stello, Director Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Stello:

Re: Turkey Point Unit 4 Docket No. 50-251 Proposed Amendment to Facilit 0 crating License DPR-41 i I In accordance with 10 CFR,50.30, Florida Power 6 Light Company submits herewith three >>(3) signed orig'inals and forty (40).

copies of a request to amend Appendix A of Facility, Operating License DPR-41.

i, This proposal is being submitted as a result of a re-evaluation of HCCS coolingperformance calculated in accordance wi'th an approved Westinghouse Evaluation Model. The proposed change described below and shown on the accompanying Technical

's Specification pages bearing the date of this letter in the lower right hand corner.

Page 3.2-3 is designated applicable to Unit 3 only.

New page 3.2-3a is designated applicable to Unit 4 only.

The new page contains a revision to Specification 3.2.6.a such that the limit on the Heat. Flux Hot Channel Factor (F<) for Unit 4 is reduced from 2.32 to 2.24.

Pa es B3.2-4 and 83.2-6 Pages B3.2-4 and B3.2-6 are designated applicable to Unit 3 only.

~Pa es B3.2-4a and B3.2-6a New pages B3.2-4a and B3.2-6a are designated applicable to Unit 4 only. The new 'pages present the basis for the revised Unit 4 limit on F<.

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Office of Nuclear Reactor Regulation Attention: Mr. Victor Stello, Director Page Two Page 3.4-1 is designated applicable to Unit 3'nly.

New page 3.4-1a is designated applicable to Unit 4 only. The accumulator water volume in Specification 3.4.1.3 is revised from 825-841 ft to 875-891 ft Although the above amendments are being proposed now, they will not be applicable until after the Spring 1977 Unit 4 refueling outage. NRC approval of these proposed amendments is requested during the spring refueling outage, however, the amendments are not necessary for the conduct of the refueling, or return to op-eration following the refueling. For the remainder of the current core Cycle 3, the accumulator minimum. water volume will remain:at 825 cubic feet. During the refueling outage, the accumulators will be modified to increase the minimium water volume to 875 cubic feet in conformance with the ECCS re-evaluation. Therefore, for the re-mainder of Cycle 3, the maximum allowable nuclear peaking factor(Fg) will be limited to 2.08 December in'accordance with the Order for Modifi-cation of License dated 3, 1976.,issued by the Commission for Turkey Point Unit 4.

The proposed amendment has been reviewed by the Turkey Point Plant Nuclear Safety Committee and the Florida Power & Light Company'Nu-clear Review Board. They have concluded that safety question. written it safety does not. involve evaluation is at-an unreviewed A tached.

Very tr y ours,'

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Robert E. Uhrig ~

Vice President REU/MAS/cpc/ls Attachments cc: Mr. Norman C. Moseley Robert Lowenstein, Esquire

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STATE OF FLORIDA )'"

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COUNTY OF DADE )

being first duly sworn, deposes and says:

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of Florida Power &

Light Company, the ><<nsee herein; That he has executed the foregoing document; that the state-ments made in this said document are true and correct to the best of his knowledge, information, and belief, and that, he is authorized to execute the document, on behalf of said >>ce>>ee-E. A. Adomat Subscribed and sworn to before me this ctay of , F9~7 L

NOTAR PUBLIC, in and for the County of Dade, State of Florida QQ74Ry pI IoI Ir <47E OF FLORIDA 4> i >>rF, 1979 c:L MY COMMISSION EXP IRKS!4N. 26, UNDERWRILEi5 BONDED TIIRl! GENERAL INSURANCE r,"~'Q 7a, '

My commission expires:

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Unit 3 reactivity insertion upon ejection greater than 0.3% A k/k at rated power. Inoperable rod worth shall be determined within 4 weeks.

b. A control rod shall'be considered inoperable if (a) the rod cannot be moved by the CRDM, or (b) the rod is misaligned 'from its bank by more than 15 inches, or (c) the rod drop time is not met.

c. If a control rod cannot be moved by the drive mechanism, shutdown margin shall be increased by boron addition to compensate for the withdrawn worth of the inoperable rod.

5. CONTROL ROD POSITION INDICATION If either the power range channel deviation alarm or the rod deviation monitor alarm are not operable rod positions shall be logged once per shift and after a load change greater than 10% of rated power. If both alarms are inoperable for two hours or more, the nuclear overpower tri'p shall be reset to 93% of rated power.

6. POSER DISTRIBUTION LIMITS

a. At all times except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:

Fq(Z) < (2 '2/P) x K(Z) for P ~ ~ 5 Fq(Z) < (4 ~ 64) x K(Z) for P < ~ 5 FNA < 1 55 [1 + 0.2 (1-P) ]

where P is the fraction of design power at which the core is operating. K(Z) is the function given in Figure 3.2-3 and Z is the core height location of Fq.

Following. initial loading before the reactor is

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b.

operated above 75% of rated power and at regular effective full rated power monthly intervals thereafter, power distribution maps, using the movable detector system shall be made, to conform that the hot channel factor. limits of the specifica-tion are satisfied. For the purpose of this comparison, Unit 3 3~ 2 3

Si Unit 4 reactivity insertion upon ejection greater than 0.3% A k/k at rated power. Inoperable rod worth shall be determined within 4 weeks.

b. A control rod shall be considered inoperable if (a) the rod cannot be moved by the CRDM, or (b) the rod's misaligned from its bank by more than 15 inches, or (c) the rod drop time is not met.

c. If a control rod cannot be moved by the drive mechanism, shutdown margin shall be increased by boron addition to compensate for the withdrawn worth of the inoperable rod.

5. CONTROL ROD POSITION INDICATION If either the power range channel deviation alarm or the rod deviation monitor alarm are not operable rod positions shall be logged once per shift and after a load change greater than 10% of rated power. If both alarms are inoperable for two hours or more, the nuclear overpower trip shall be reset to 93% of rated power-

6. POWER DISTRIBUTION LIMITS

a. At all times except during low power physics tests, the hot channel factors defined in the'asis must meet the following limits:

Fq(Z) < (2.24/P) x K(Z) for P > 5 ~

Fq(Z) < (4.64) x K(Z) for P < .5

< 1'55 [1 + 0.2 (1-P) ]

FAH where P is the fraction of design power at which the core is operating. K(Z) is the function given in Figure 3.2-3 and Z is the core height location of l

Fq.

b. Following, initial loading before the reactor is operated above 75% of rated power and at regular effective full rated power monthly intervals thereafter, power distribution maps, using the movable detector system shall be made, to conform that the hot channel factor limits of the specifica<<

tion are satisfied. For the purpose of this comparison, 30 2 3 a 12/9/76 i$

Unit 4

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Unit 3 An upper bound envelope of, 2.32 times the normalized peaking factor axial dependence of Figure 3.2>>3 has been determined (from extensive analyses at.

design power considering all operating maneuvers) to be consistent with the technical specifications on power distribution control as given in Section 3.2. The, results of the loss of coolant accident analyses based on this upper bound envelope indicate a peak clad temperature of 2150'F at design power; corresponding to. a 50'F margin to the 2200 F FAG limit.

When an F measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate experimen'tal uncertainty allowance for a full core map taken with the movable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.

N In the, specified limit of F , there is an 8 percent allowance for uncertain-ties which means that normal operation of the core is expected to result. in F (1.55/1.08. The logic behind the 1'arger uncertainty in this case is that AH-(a) normal perturbations in the radial power shape (e.g., rod misalign-N ment) affect P H, in most cases without necessarily affecting F ,(b) the has a direct influence on F through moveme'nt of rods, and can limit AH'perator it to the desired value, he has no direct control over F<H N and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests can be compensated for in P by tighter axial control, but compensation for F is less readily available. When a measurement of F

AH is taken, experimental error must be allowed for and 4% is the appro-priate allowance for a full core map taken with the movable incore detector flux mapping system.

Measurements of the hot channel factors are required as part of start-up physics tests, at least once each full rated power month or operation~ and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map taken following initial goading provides confirmation of the basic nuclear Unit 3 a3.2-4 12/9/76

Unit 4 An upper bound envelope of 2. 24 times the normalized peaking factor axial dependence of Figure 3.2-3 has been determined to be consistent with the technical specifications on power distrib'ution control as given in Section 3.2.

When an F measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate experimental uncertainty allowance for a full core map taken with the movable incore detector flux mapping syst'm and three percent is the appropriate allowance for manufacturing tolerance.

In thh specified limit of N F>H, there is an 8 percent allowance for uncertain-ties which means that normal operation of the core is expected to result in F <1.55/1.08. The logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power shape (e.g., rod misalign-ment) affect F , in most cases without necessarily affecting F ,(b) the has a direct influence on F through movement of rods, and can limit q'perator it to the desired value, he has no direct control over F>H N and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests can be compensated for in F by tighter axial control, but compensation for F is less readily available. When a measurement of F

AH is taken, experimental error must be allowed for and 4% is the appro-priate allowance for a full core map taken with the movable incore detector flux mapping system.

1feasurements of the hot channel factors are required as part of start-up physics tests, at least once each full rated power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map taken following initial Roading provides confirmation of the basic nuclear Unit 4 B3.2-4 a

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Unit 3 Flux Difference (4$ ) and a reference value which corresponds to the full design power equilibrium value of Axial Offset (Axial Offset 44/fractional power) . The ref erence value of flux difference varies with power leve1 and burnup but expressed as axial offset it varies only with burnup.

The technical specifications on power distribution- contro1 assure that th' F upper bound envelope of 2.32 times Figure 3.2-3 is not exceeded and xenon q

distributions are not developed which at. a later time, would cause greater local power peaking even though the flux difference is then within the 1imits specified by the procedure.

The target (or reference) value of flux difference is determined as follows.

At any time that equilibrium xenon conditions have been established, the in-dicated flux difference is noted with part length rods withdrawn from the core and with the full length rod control rod bank more than 190 steps withdrawn (i.e., normal rated power operating position appropriate for the time in life.

Control rods are usually withdrawn farther as burnup proceeds).- This value, divided by the fraction of design power at which the core was operating is the design power value of the target flux difference. Values f'r all other core power levels are obtained by multiplying tne design power value by the fractional .power. Since the indicated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviation of

+5% 4I are permitted from the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring the target flux difference every rated power month. For this reason, methods are permitted by Item 6c of Section 3,2 fo'r updating the target flux differences. Figure B3.2>>1 shows a typical construction of the target flux difference band.. at BOL and Figure B3.2-2 shows the typical variation of the full power value with burnup.

Strict:control of the flux difference (and rod*position) ih not as necessary during part power operation. This is because xenon distribution contro1 at part power is not as significant as the control at full power and allowance t

has been made in predicting the heat flux peaking factors for less strict co"..

trol at part power. Strict control of the flux difference i.s not possible during certain physics tests or during the required, periodic excore calibra-3 Unit 3 B3.2-6

,,12/9/76

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Unit 4, Flux Difference (hP) and a reference value which corresponds to the full design power equilibrium value of Axial Offset (Axial Offset ~ ~4/fractional power). The reference value of flux difference varies with power level and burnup but expressed as axial offset it varies only with burnup.

The technical specifications on power distribution contro1 assure 'that the F upper bound env'elope of 2. 24 times Figure 3.2-3 is not exceeded and xenon distributions are not developed which at a later time, would cause greater local power peaking even though the flux difference is then within the limits specified by the procedure.

The target (or reference) value of flux difference is determined as follows.

At any time that equilibrium xenon conditions have been established, the in-V dicated flux difference is noted with part length rods witndrawn from the core and with the full length rod control rod bank more than 190 steps withdrawn (i.e., normal rated power operating position appropriate for the time in life.

Control rods are usually withdrawn farth r as burnup proceeds).'his value, divided by the fraction of design power at wnich the core was operating is the design power value of the target flux difference. Values for all other core power level@ are obtained by multiplying tne design power value by the fractional power. Since the indicated equilibrium valu was noted, no allowances for excore detector error are necessary and indicated deviation of

+5% GI are permitted from the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring the target flux difference every rated power month. For this reason, methods are permitted by Item 6c of Section 3.2 for updating the target flux differences. Figure B3.2-1 shows a typical construction of the target flux difference band at BOL and Figure B3.2-2 shows the typical variation of the full power value with burnup.

Stiict control of the flux difference (and rod position) is not as necessary II during part power operation. This is because xenon distribution control at part power is not as significant as the control at full power and allowance has been made in predicting the heat flux peaking factors for less strict co..

trol at part power. Strict control of the flux difference is not possible during certain physics'tests or during the required, periodic excore calibra-Unit 4 B3.2-6 a 12/9/76

UNIT 3 3.4 ENGINEERED SAFETY FEATURES Applies to the operating status of the Engineered Safety Features.

~Ob ective: To define those limiting conditions for operation that are necessary: (1) to remove decay heat from the core in emergency or normal shutdown situations, (2) to re-move heat from containment in normal operating and emergency situations, and (3) to remove airborne iodine from the containment atmosphere in the event of a Maximun Hypothetical Accident.

a. The reactor shall not be 1

made critical, except for low power physics tests, unless the following conditions are met:

1. The refueling water tank shall contain not less than 320,000 gal. of water with a boron con-II centratlon of at least. 1950 ppm

2. The boron injection tank shall contain not less than 900 gal. of a 20,000 to 22,500 ppm boron solution. The solution in the tank, and in isolated portions. of the inlet and outlet piping, shall be maintained at a temperature of at least 145F. TMO channels of heat tracing shall be operable f'r the flow path.

3. Each accumulator* shall be pressurized to at least 600 psig and contain 825-841 ft3 of water with a boron concentration of at least 1950 ppm, and shall not be isolated.

4. FOUR safety injection pumps shall be operable.

3.4-1 UNIT 3

'i2/9/76

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UNIT 4 3.4 ENGINEERED SAFETY FEATURES Features.

~Ob ective: To define those limiting conditions for operation that are necessary: (1) to remove decay heat from the core in emergency or normal shutdown situations, (2) to re-move heat from containment in normal operating and emergency situations, and (3) to remove airborne iodine from the containment atmosphere in the event. of a Maximum Hypothetical Accident.

a. The reactor shall not be made critical, except for low power physics tests, unless the following conditions are met:

1. The refueling water tank shall contain not less than 320,000 gal. of water with a boron con-centration of at least 1950 ppm.

2. The boron injection tank shall contain not less than 900 gal. of a 20,000 to 22,500 ppm boron, solution. The solution in the tank, and in isolated portions of the inlet and outlet piping, shall be maintained at a temperature of at least 145F. TWO channels of heat tracing shall be operable for the flow path.

3. Each accumulator shall be pressurized to at least 600 psig and contain 875-891 ft of water with a boron concentration of at least 1950 ppm, and shall not be isolated.

FOUR safety injection pumps shal3. be operable.

UNIT 4 3e4-1 a

. 12/9/76

SAFETY EVALUATION I. Introduction This safety evaluation supports the following proposed changes to the Unit 4 Technical Specifications:

(1) The maximum allowable nuclear peaking factor (F )

is decreased from 2.32 to 2.24.

(2) The limits on Safety Injection accumulator water volume are increased from 825-841 ft to 875-891 ft II. Discussion A. Core C cle 4 A re-evaluation of ECCS cooling performance calculated in accordance with an approved Westinghouse Evaluation Model has been performed. The re-evaluation shows that for breaks up to and including the double ended severence of a reactor coolant pipe, the ECCS will meet the Acceptance Criteria presented in 10 CFR 50.46.

The detailed re-evaluation is contained in FPL letter L-76-419 of December 9 , 1976, and shows that, at a core power level of 102% of 2192 Mwt and a minimum accumulator water volume of 875 ft, per accumulator, the maximum allowable nuclear peaking factor is 2.25.

However, since the Technical Specifications allow a maximum core power level of 2200 Mwt, the re-evaluation is being revised using the higher power level. The revised calculation is expected to yield a maximum Fq of 2.24.

B.

The following analysis shows that, by operating with F<

<2.08 for the remainder of Cycle 3, we can provide conservative assurance of safe operation.

The Westinghouse ECCS re-evaluation assumed:

1. 10% steam generator tube plugging ft2.25

2. F =

3. 895 accumulator minimum water volume

4. 2192 Mwt core power level However, for the remainder of Cycle 3, the actual parameters will be:

l. 7% steam generator tube plugging

=

2. F ft2.08

3. 895 accumulator minimum water volume

4. 2200 Mwt core power level

SAFETY EVALUATION (Continued)

Therefore, the following adjustments are needed to adapt the ECCS re-evaluation to the remainder of Cycle 3:

(1) Tube Plu ing Ad'ustment From the Westinghouse ECCS re-evaluation, Peak Clad Temperature (PCT) is:

(a) PCT =

2192 Mwt 2198'F for 10% plugging, 875 ft (b) PCT =

Mwt 2162'F for 5% plugging, 875 ft , 2192 As shown, the PCT varies by 36'F as tube plugging increases from 5% to 10%. Therefore, for 7% tube pluggingF PCT = 2198 F [ (. 60) (36'F) ]

2198DF 21.6DF 2176.4DF (2) F Adjustment PCT is further reduced by lowering F from 2.25 to 2e08 (BFq = e17) ~

using the relationship 10 F By it can be shown that . 17F q 01F

(

10 F

.01F ) = 170'F, i.e., lowering F f=om 2.25 to 2.08 results in a reduction of 190'F in PCT. Therefore, for Fq=2.08, PCT = 2176.4'F 170'F 2006.4DF

SAFETY EVALUATION (Continued)

(3) Accumulator Wa ter Volume Ad'ustment Westinghouse ECCS analyses show that a decrease in accumulator water volume from 875 ft to 825 ft Therefore, corresponds to a for 825 ft3, 36'F increase in PCT.

PCT = 2006.4 F + 36 F 2042.4oF (4) Core Thermal Power Adjustment Increasing the core power from 2192 Mwt to 2200 Mwt will increase the PCT as follows:

(a) hFq = ) (2.25) = .008 2192 (b) (. 008Pq) . 01F 01Fq 8 P Therefore, for 2200 Mwt, PCT = 2042.4 F + 8 F 2050.4 F Thus, operation with F <2.08 for the remainder of Cycle 3 will result in a PCT of 2050.4'F, which is well below the ECCS acceptance criteria of 2200'F.

III.Con'clusions s

Based on these considerations, (1) the proposed change does not increase the probability or consequences of accidents or malfunctions of equipment important to safety and does not reduce the margin of safety as defined in the basis for any technical specification, therefore, the change does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and

'security or to the health and safety of the public.

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