View source for L-76-429, Letter Transmitting Proposed Amendment to Facility Operating Licenses, Based on Result of re-evaluation of ECCS Cooling Performance Calculated in Accordance with an Approved Westinghouse Evaluation Model

{{Adams
| number = ML18227C907
| issue date = 12/21/1976
| title = Letter Transmitting Proposed Amendment to Facility Operating Licenses, Based on Result of re-evaluation of ECCS Cooling Performance Calculated in Accordance with an Approved Westinghouse Evaluation Model
| author name = Uhrig R
| author affiliation = Florida Power &amp; Light Co
| addressee name = Stello V
| addressee affiliation = NRC/NRR
| docket = 05000250, 05000251
| license number = DPR-031, DPR-041
| contact person = 
| case reference number = L-76-429
| document type = Legal-Affidavit, Letter type:L, License-Operating (New/Renewal/Amendments) DKT 50
| page count = 23
}}

=Text=
{{#Wiki_filter:MRC FORM 195                                              U.S. NUCLEAR REGULATORY       CO'ION    DOCKET NUMBER (2. 76)                                                                                                50-250 FILE NRC DISTRIBUTION FQR PART 50 DOCKET MATERIAL                                            NUMBER'O:
Mr>>-V>>   Stell o                         FROM:    FPL                                     DATE OF DOCUMENT 12;21>>76
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Miami, Fla,       33101 R>>E,   Uhrig                            DATE F C 6 I V6 D RLETTER                  RlNOTORIZED             PROP                     INPUT FORM              NUMBER OF COPIES RECEIVED CI ORIGINAL          K UNC LASS IF I ED Q COPY
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3          sigped 37     CC DEBCBIr TIDN    Ltr notarized 12-21 76 requesing for                            Revised & addi pages to OL/Tech Specs Amdt    to OL/Tech Specs ~ >>,trans the following                  pertaining to       ECCS cooling performance ~ ~ ~
I (40 cys encl rec'd)
.PLANT NAME:         Turkey Pt, Unit 3 SAFETY                                FOR 'ACTION/INFORMATION                                   DHL           12>>
ASSXGNED AD:
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CONS    TANTS fI NRC FORM 196 {2 76)

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                                   .5'LORIDA ppCK54                                                                    1'4
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                           .ag j                                           POWER & LIGHT COMPANY ylr                                  December 21, 1976 L-76-429
                               ,.'Hegulati>ji Oo(:kej Fjf>8'ffice of Nucleai,Reactor Regulation Attention: Mr.'ictor Stello, Director                                                  @COfto Division of Operating Reactors                                         SSc       SN U. S. Nuclear Regulatory Commis'sion Washington, D. C. 20555                                                             O,S.~~IA'       ~

==Dear Mr. Stello:==

Re:   Turkey Point Unit 3 Docket No. 50-250 Proposed Amendment to Facility 0 eratin License DPR-31 In accordance with 10 CFR 50.30 Florida Power 6 Light Company hereby submits three (3) signed originals and forty (40) copies of a request. to amend Appendix A of Facility Operating License DPR-31.
This proposal is being submitted as a result of a re-evaluation of ECCS cooling performance calculated in accordance with an approved Westinghouse Evaluation Model. The ECCS re-evaluation was forwarded to you on December 9, 1976 under our cover letter L-76-419. The proposed changes are described below and shown on the accompanying Technical Specification pages bearing the date of this letter in the lower right hand corner. NRC approval of the proposed amendment is requested prior to the end of the current refueling outage, however, the amendment is not necessary for the conduct of the'efueling, or return to operation following completion of refueling.
Page  3.2-3 Page  3.2-3 is designated applicable to Unit 3 only. The page contains a revision to Specification 3.2.6.a such that the limit on Heat Flux Hot Channel Factor (Fq) for Unit 3 is reduced from 2.32 to 2.24.
Page  3.2-3a Page  3.2-3a is designated applicable to Unit        4           only.
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Office of Nuclear Reactor Regulation Attention: Mr. Victor Stello, Director Page Two Pa es  B3.2-4 and B3.2-6 Pages  B3.2-4 and B3.2-6 are designated applicable to Unit   3 only. These pages present the basis for the revised Unit   3 F<  limit.
Pa es  B3.2-4a and B3.2-6a Pages  B3.2-4a and B3.2-6a are designated applicable to Unit   4 only.
Page  3.4-1 Page  3.4-1 is designated applicable to Unit 3 only. The accumulator water volume in Specification 3.4.1.3 is revised from 825-841 ft~ to 875-891 ft.
Page  3.4-1a Page  3.4-la is designated applicable to Unit  4  only.
The proposed amendment has been reviewed by the Turkey Point Plant Nuclear Safety Committee and the Florida Power 6 Light Company Nuclear Review Board. They have concluded that, it does not involve an unreviewed safety question. A safety evaluation is attached.
Very truly yours, Robert E. Uhrig Vice President REU/MAS/cpc Attachments cc:   Mr. Norman C. Moseley Robert Lowenstein, Esquire

STATE OF FLORIDA      )
                      )     ss  ~
COUNTY OF DADE        )
Robert E. Uhrig, being   first     duly sworn, deposes and says:
That he is a Vice President of Florida Power                       6  Light Company, the Licensee 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 Licensee.
Robert E. Uhrig Subscribed and sworn to before         me     this day  of NOTARY PUB   IC, in and for the County of Dade, State of lorida                >PY Pll RV STATE OP Fl.ORIDA AV l ~n~e lAY CONhtlSS(OlII EXPIRES JAM.
26, l970 VPO@THKU GE

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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 CRDH, or (b) the rod    is misaligned from    its  5@ggy~more         O>IL$0 than 15   inches, or                 ~~!!!Ml-'> t Wgl1 (c) the rod    drop time de oot met. P~".g,gp~yot<~~@
: 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 xod.
: 5. CONTROL ROD POSITION INDXCATXON 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 axe inoperable for two hours or more, the nuclear overpower trip shall be reset to 93% of rated power.
: 6. ONER 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)   <  (2.24/P) x K(Z) for P > .5 Fq(Z) < (4.48) x K(Z) for P-< .5 FgH < 1.55 [1 + 0.2 (1-P) 3 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.oE Fq.
: b. Following    initial loading   before the reactor is .
operated above /5% of rated power and at regulax-effective full rated power monthly intexvals thereafter, powex distribution maps, using the movable detector system shall be made, to conform that the hot channel factor limits of the specifi.ca-tion are satisfied. For the purpose of this comparison, 3~2 3                                   12/21/76

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UNIT 4 reactivity insertion   upon  ejection greater than 0.3% 6  k/k at rated  power. Inoperable rod worth shall be dete ined within 4 weeks.
: b. A control rod shall be considered inoperable       if (a) the rod can..ot be moved by the CRD~f, 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 nargin, 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. POLAR DISTRIBUTION LLitITS
: 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) < (2.32/P) x K(Z) for P ) .5 Fq(Z) < (4.64) x K(Z) for P < .5 FN < 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 oiven in Figure 3.2-3 and Z is the core height location of'q.
b.. Following initial loading before the reactor is operated above 75% of rated powex and at regular
            .effective full rated power monthly intervals thereafter, power distribution maps~ using the movable detec ox 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 3a 12/21/76

0 UNIT 3 3.4         ENGINEERED SAFETY FEATURES Applies to the operating sta-us of the Engineered Safety Features.
Ob jeer='=         To define-those      limiti"g co ditions for operation that are necessary:       (1) to re=ove decay heat from'he core in emergency or normal shutdown situations, (2) to re-move heat from conta n=ent in normal operating and emergency situations, and (3) to remove a$ .rborne iodine from the containment atmosphere in the event of a Ifaximum Hypothetical Accident.
: l. SAFETY INJECTION ='~1') RES~MUAL HEAT RENOVAL SYSTPiiS
: a. The" reactor shall not be made crier.ca1, except for low power physics tests, unless the following 1., The refueling water tank shall. contain not less than 320,000 gal. of water with a boron con-centration of at least 1950 ppa.
: 2. The   boro"   i jection   tank shall contain not less than 900 ga". o= a   20,000 to 22,500 ppm boron solution. The solution in the          tank, and in isolated portio"s of the inlet         and    outlet piping, sha' be -aintained at          a temperature of at lees" 145:-. T40 channels          of heat tracing shall be opera"'e for the flow         path.
: 3. E'ach  accu=ulator shall be pressurized to at least 600 psig and contain 875-891 Ift      I   3 of water witn a bc=on concentration of at least 1950 pp..., and   j-.=-ll not be  iso3.ated.
: 4. FOUR   safety in-.ection    pumps  shall   be   operabl
: 3. '-1 UNIT 3 12/21/7b

0 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.
: 1. SAFETY INJECTION AND RESIDUAL HEAT         ~i OVAL  SYSTEMS'.
The   reactor shall not be   made    critical,           except  for-low power physics tests, unless the following conditions are met:
l.~ 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. TtlO channels of heaL tracing shall be operable for 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 operablee 3e4-1 a 12/21/76 I

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 distribution 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 experimenta1-uncertainty allowance for a full'ore map taken with the movable incore detector flux mapping system and three percent is the appropriate allowance for manufactuzing tolerance.
In thh 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
         <1.55/1.08. The logic behind the larger uncertainty zn this case ia that, F hEE (a) " normal perturbations in the radial power shape (e.g., zod misalign-ment) affect F>H, in most cases without necessarily affecting F ,(b) the operator has a direct influence on F through movement of rods, and can limit it to the desired value, he has no direct control over F8 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 v
F bH is taken, experimental error must be allowed for and 4/ is the appro-priate allowance for a full coze map taken with the movaole 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 of'peration, and whenever abnormal power distribution conditions require a r duction of core power to a level based on measured hot channel factors.       The incore map taken following initial Roading provides confirmation o the basic nuc1ear UNIT 3
: 83. 2-4 12/21/76

UNIT 4 M upper       bound envelope  of 2.32 ti es tna nox alized peak.ing factor axiaL c'apendanca of Figure 3.2-3 has been Eat                  'nad (from extansive analyses at design power considering all opera'g ma"auvers) to ba consistent with tna technical specifications on power distr bution control as given in Section 3.2. The results of the loss of coo>>~t accident analys~ based on this upper 'bound envelope indicate a peak cl-d. temperature of                 2150't    design power, corresponding to a 50 P marg~ to the 2200 P FAC 3.imit 0
xnan an F measuremant is taken, bo~ experimental error and manufacturing tolerance must be allowed for. Piva percent is the apprapriate experimentaL
   'ncertainty allowance for a full cora map taken with the. znovable incore detector flux mapping system and tare pe-cent is the ap~ropriate allowance for manufacturing tolerance.
La the     specified    limit of P,         there   is  an 8  percent allo~ance for uncertain-wnich means that normal oparat on of tha core is a:~ected to result in bH'ies
           <1.55/1.08. Tne logic behind tne larger uncertainty an this case is that (a)      normal perturbatiors in the red al ".ower.supe (e.~., rod misalign-                ~
affect   P,    in most cases without necessarily affecting P,(b) the                 .'ant) operator has a direct influence on = 6 tnrough movement oE rods, and can linjt H
it    to the desired value, ha has no c=ract control. over F H and-(c) an erro-in tha predictions for radial powa" shape, which may b cRetectad during stzrtuo physics tests can be compensa ad =or in Fq by tip>ter'xial control bu" conpensation for P> is less raa ily available. E'en a measurement of
       ~~H is taken, experimental error rust ba allowed for and l~% is. the appro-"-
pria"a allowance 'for a full core r-p take" A.th the movable"incore detector                          ~
flux mapping      system.                                                    '
Zaasuremants      of the hot channel facto s a"a required as part of start-up physics tests, at least orce each fuU. ra-ed power month of operation, and v,.",anavar abnormal power distribution co;.=itiors require a rad ction of core po=-ar to a level based on measured ~ot channel factors                    Th rcore map taken following in "ial 3:oading pro .=des confirmation o                th basic nuclear
                                                      ')  / a UNIT 4 12/21)7r,

UNIT 3 Flux Difference (Llg) and a reference value which corresponds to the full design power equilibrium value of Axis~ Of set (Axial Offset = A4/fractional power). The reference value of flux difference varies with power level and burnup but expressed as axi 1 offset   it varies only with burnup.
The   technical specifications on power distribution control. assure that tne F upper bound envelope of 2.24 times Figur 3.2-3 is not exceed d and xenon distributions are not developed which at a later time, would cause greater local power peaking even though the flux difference is then witnin the limits speciried 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 tne 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 fart". r as burnup proceeds). This value, divided by th- fraction of design power at which the core 'was operating is tne design power value of the target flux difference. Values for all other core power levels are obtained oy multiply=-ng tne design power value by the fractional power. Since the indicated equilibrium valu was noted, no allowances for ezcore detector error ar necessary and indicated deviation of
   +5/ 61 are permitted from the indicated reference value. During peri'ods where extensive load following is. required,     it may be impractical to establish the required core conditions for measuring tne 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. Fipuxe B3.2-1 shows a typical construction of the target flux dif=erence band at 'BOL*and Figure B'3.2-2 shows the typical variation of the fuM> power value with buxnup.
Strict control of   the flux difference (and rod position) is not as necessary during part power operation. This is because xenon distribution control at part power is not as significant as tne control at full power and allowance has been made in pr dieting the heat "lux peaking factors for less strict co=
trol at part power. Strict control of the ux difference is not possible
                                                 =.
during certain physics tests or during tne        quired, periodic excore calibra-UNIT 3                                '3.2-6 12/21/76

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UNIT 4 Flux Dizference {Ag) and a reference -alue which corresponds to tne full desim power equilibrium value of A~~ Of=set (Axial Offset ~ <0/fractional po"er). The reference value of flux izzer nce varies with power level and burnup but expressed as azi 1 offset "t varies only with burnup.
inc technical specifications on pow r distribution control assure that tne P upper bound envelope of 2.32 times Pigur 3.2-3 is not exceed d and. xenon dzstrxbutzons are not developed whicn at a latez time, would cause greater local power peaking -even though the flux di=ference is then within the limits specified by the procedure.
The   target (or reference) value of fizz differ'ence is determined as fo1lows.
At any time that equilibrium xenon conditions have been established, the. in-dicated flux di fference is noted witn par t length rods withdrawn from. the core and with tne full length zod control =od ba"k more than 190 steps withdrawn (i.e., normal rated power operating positio" appropriate for the time in life..
Control rods are usually withdrawn faith r as burnup proceeds). This. value, divided by tha fraction of design power at -hich the core was operating is tne.
design power value of the target f1m cizfe"ence. Values Sor aU ot'hez. core power levels are obtained by multiplymg the design power value by the fract.onal power. Since the indicate: equilibrium valu was noted, no allo =-aces for excore detector error ar necessary and indicated deviation of
  +5K LI are permitted from the indicatM re= rence value.        During periods where extensive load following is required,      it may be impractical, to establish the required core conditions fo m asuring -he target flux difzerence every rated power month. Por this reason, ~ thocs are permitted by Item 6c of Section 3 2 for updating the target f~uz di=ferences. Figure B3.2-1 shows a
             ~
typical construction of the target fl w dif=erence band at EOL and Piguge B3.2-2 snows tne typical variation of the fuH~ po r value with burnup.
Strict control of    the flux difference {and od position) is not as necessary during part power operation. This is be"a se xenon distribution control. at part power is not as significant as the co=trol at full power and allowan'ce h s b=-en made in predicti.".g the heat = ux "=-aking factors for less trict co=;
trol at part power. Strict control of t..e =:ux difference xs no possible duri-..= certain physics tests or   du ig   ne =  -uir d, per odi.c excore ca1i'ora-UNIT 4                                 B3.2-5 a 12/21/76    I

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SAFETY EVALUATXQ'8 1.0       introduction This safety evaluation supports the ollowing proposed changes to the Unit 3 Technical Specifications:
The maximum allowable nuclear peaking'actor (Pq) is decreased from 2.32 to 2.24.
The  limits  on Safety ejection accumulator water volume are increased from 825-841 875-891   ft3..                     ft3 to
 ~
2.0        Discussion ECCS      Re-evaluation A     re-evaluation of ECCS cooling. performance calculated
       .-:     in accordance with an approved 'Plestinghouse Evaluation Model has been performed. The re-evaluation shows that for breaks up to and including the douhle ended
           'everence of a reactor coolant pipe, the ECCS vill.
meet, the Acceptance Criteria presented in 10 CFR 50.46.
The detailed re-evaluation is contained in PPL letter
         .. L-76-419 of December 9, 1976, and shows that, at a core power level of 102% o~ 2192 Nwt and a minimum water volume of 875 ft. per accumulator, 'ccumulator the maximum allowable nuclear peaking factor. is 2.25.
            'owever, since the Technical Specifications allow a maximum core power level o 2200 Mwt, the re-evaluation.
is being revised using the higher power level. The revised calculation is expected to yield a maximum F<
of 2.24.
: v. 'I
'I3.0        Conclusions 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       vill  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. wilJ not be inimical to t:he common      defense and security or to the health and safety of the public.

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