View source for L-76-172, Proposed Amendment to Facility Operating License DPR-67. Allow FPL to Perform as Much of the Power Ascension Test Program as Possible

{{Adams
| number = ML18096B519
| issue date = 04/27/1976
| title = Proposed Amendment to Facility Operating License DPR-67. Allow FPL to Perform as Much of the Power Ascension Test Program as Possible
| author name = Uhrig R
| author affiliation = Florida Power &amp; Light Co
| addressee name = Stello V
| addressee affiliation = NRC/NRR
| docket = 05000335
| license number = DPR-067
| contact person = 
| case reference number = L-76-172
| document type = Letter type:L, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Amendment
| page count = 27
}}

=Text=
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMI                  ON     DOCKET NUMOE+
NRC FORM 195 I2.76)                                                                                                             5o-33 FILE NUMEEB NRC DISTRIBUTION FDR PART 50 DOCI(ET MATERIAL TO:                                                                                                           DATE OF DOCUMENT
                                                        'FLORIDA POWggg LIGHT CO MR    V Swum                                     MIAMI,   mJI.,~i, R E UHRIGo  ~ ~ ~ ~ ~ ~                                  ~    l4 DATE RECEIVED QNOTORIZEO               PROP                     INPUT FORM                          NUhlBEB OF COPIES RECEIVED
)GALETTE R I "3'Si'qn ORIGINAL CI  COPY JUNC LASS I F I E D                                                                                          incL
       "'"   'LTR NOTORIZED II-28-76................              ENCLOSURE
  =
LTR FURN INFO ON PROPOSED AMDT TO THE                         FURNISHING CHARTS, TBLS, AND FIGS IN REF TO OPERATING LICENSE...... TRANS THE FOLLOWING... :THE 'INTERIM CHANGE TO. THE. TECH SPEC AS REQUESTEDo       ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~
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                                                                                 ~ RETAINED'Y ORB-2.............
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NS ASLB                             CONSULTANTS NllC I OBM 196 (2 76I

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P     OX  013100'IAMI'LORIDA 33101
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FLORIDA POWER  5 LIGHT'OMPANY April 27,         1976 L-76-172 I
Director of Nuclear Reactor Regulation                               8 Mr. Victor Stello, Director    'ttention:
Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555 LP~

==Dear   Mr.'tello:==

Re:    St. Lucie Unit      1 Docket No. 50-335 Proposed Amendment      to Facilit     0  eratin License DPR-67 In accordance with 10 CFR 50.30, Florida Power .and Light Company (FPL) submits herewith three (3) signed originals and forty (40) conformed. copies of a request to amend Appendix A of Facility Operating License DPR-67.
Current     FSAR analyses, setpoint analyses, and Technical Specifica-tions are based on a reactor coolant flow rate of 370,000 gpm.
However, post-coie hot functional measurements have not demonstrated at this time that this flow rate exists. As a result, an interim Technical Specification amendment is needed to allow FPL to perform as much of the power ascension test program as possible.                              This amendment.,is    based   on  a.reactor  coolant           flow rate     of   354,000       gpm (109.0% of design flow). A reanalysis is being performed to support full power, operation at the lower flow rate. The re-analysis is expected to result in a proposed permanent change to the Technical Specifications but until then the change being proposed by         this letter is it   is emphasized that considered temporary.
The proposed interim changes are as described below and as shown on the accompanying Technical Specification pages bearing the date of this letter in the lower right hand corner.
Page 2-2                                                                           A VESSEL FLOW LESS MEASUREMENT UNCERTAINTIES on                     Figure 2.1-1 is  changed from 370,000      GPM to 354,000           GPM.      (Interim change).
gg'p HELPING BUILD FLORIDA

0 0 II

Director of Nuclear Reactor Regulation Attention: Mr. Victor Stello, Director Division of Operating Reactors Page Two April 27,    1976 Pacae   2-4 The    note at the bottom of Table 2.2-1 is changed to read "Design reactor coolant flow with 4 pumps operating is 354 000 gpm-" Xn Item 2, 1 06. 5% is changed to 96 5% . (Xnterim)
F                                               ~            ~
Pacae 2-8 On  Figure 2.2-3, the last factor in the PVTRIP equation is changed from 6250 to 6150. (Interim changers.
Pacae   2-9 TRXP On  Figure 2.2-4, the last factor in the Pun       equation is changed from 6250 to 6150.       (Interim change Pa e    B2-2 The    values of ROD RADIAL PEAK are changed to those indicated on attached     Figure B2.1-1. (Interim change).
Page    3/4 2-2 Interim specification 4.2.1.4.b.7 is added, to read as   follows:
            "An interim flow adjustment factor of 1.10 is to be implemented  if the reactor coolant flow rate is less than 370,000 gpm and greater than 354,000 gpm."
Pa e    3/4 2-4 Figure 3.2-2 is revised to reflect a power limitation of 90 per cent pending reanalysis at the 354,000 gpm flow rate. (Interim change).

Director of Nuclear Reactor Regulation Attention: Mr. Victor Stello, Director Division of Operating Reactors Page Three April 27,     1976 Page 3/4 2-13 In Table 3.2-1, the reactor coolant flow rate is changed from 370,000 gpm to 354,000 gpm. (Interim change);
Page  3/4 2-14
    . Figure 3.2-4 is revised to reflect a power limitation of 90%
pending reanalysis. at the 354,000 gpm flow rate.    (Interim change)  .
Page B3/4 2-1 Bases   3/4.2.1 is revised to include mention of the 1.10 interim flow factor.. (Interim change).
The proposed interim amendment has been reviewed and the con-clusion reached that      it does not involve a significant hazards consideration, therefore, prenoticing pursuant to 10 CFR 2.105 should not be required. A written safety evaluation is attached.
Very           yours, ob   rt E.
Vice President Uhri.g"        (
REU/MAS/cpc Attachment cc:    Mr. Norman C. Moseley Jack R. Newman, Esquire

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STATE OP PXORXDA     )
                     )  SS COUNTY OP DADE       )
Robext E. Uhrig', being f3.rst duly sworn, deposes                        and says'hat he  is  a Vice President of Plorida Power 6 Light Company,   the JiCenSee herein; That he has executed <he forogoing Bocumont~ that the statements made in this said document are true and coxrect to the best of his J:novledge, information and belief, and, that he'is authorized to execute the document on behalf of said xicensee.
Subscribed and sworn    to before         me this        7E,'ay          of,                                     1976.
Motar    Pub). q, n an d'or the County of Dado,  State of Plorida QCITAAY PUdl.lC SlAlC Gt A.QAIOA 0l LAP(6 "gY  CCMNllSS10N   fXPIBES APllll. 2,  le Ny commission Q.spires,     >pHosg   vlwu.vAwAAo eweNG          AdPNo"v

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TABLE  2.2-1 REACTOR PROTECTIVE INSTRUMENTATION             TRIP SETPOINT LIMITS, FUNCTIONAL UNIT                                    TRIP SETPOINT                             ALLOWABLE VALUES 1 . Manual Reactor  Trip                           Not Applicable                            Not Applicable
                                                                                                                                    .Oe
: 2. Power Level  - High (1)
Four Reactor Coolant    Pumps                  <  9.61K above   THERMAL POMER,           <  9.61W.above  THERMAL. POWER,    andO Operating                                      with a minimum setpoint of 155        . a minimum setpoint of 155 of RATED of RATED THERMAL POttlER, and a         . THERiSL POWER and a maximum of maximum of < 96 . 5g 'of RATED            < 96, 5$ of RATED THERMAL POWER.
THERMAL POllER.
: 3. Reactor Coolant Flow    -  Low   (1)
Four Reactor Coolant Pumps      .         ..   > 95% of. design reactor coolant         > 95'5 of design reactor coolant Operating                                      flow with    4 pumps operating*           flow with 4 pumps operating*
: 4. Pressurizer 'Pressure - High         .';.    .
                                                     < 2400   psia                             <-2400 psia
: 5. Containment Pressure    -  High                < 3.9 psig                                < 3.9 osig
: 6. Steam Generator Pressure     -  Low     (2)    > 485   psig                              > 485  psig
: 7. Steam Generator Water Level -Low               > 36.3'A Water Level   - each         '; > 36.3C Mater Level    -
generator each'team steam generator
: 8. Local Power Density   -   High (3)             Trip 'setpoint adjusted to not            Trio set ooint adjusted to not exceed the limit lines of                 exceed the limit lines of Figures 2.2-1 and 2.2-2                   Figures 2.2-1 and 2.2-2.
~
  *Design reactor coolant flow with        4 pumps     ooerating is 354,000**    . ."'..-,
  **Interim value pending reanalysis.

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IP ST. LUCIE  -   UiiIT 1 4/27/76

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PO~(ER  DISTRIBUTION L If)ITS SURVEILLANCE     RENDU IREHENTS        Continued
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17.85     xM where:
L  is the    maximum   allowable linear heat rate as determined from Figure      3.2-1  and is based on the core average burnup at the time of the latest incore flux map.
: 2.       M  is the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination.
4.2.1.4      Incore Detector Honitorin S stem - The incore detector moni-toring system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:
: a. Are adjusted to          satisfy the requirements of the core   power distribution         map  which shall be updated at least once per    31 days.
: b. Have     their alarm setpoint adjusted to less than or equal to the limits shown on Figure 3.2-1 when the following factors are appropriately included in,the setting of these alarms:
Flux peaking augmentation factors as shown in Figure 4.2-1,
: 2.       A  measurement-calculational       uncertainty factor of 1.10,
: 3.       An    engineering uncertainty factor of 1.03, 4~       A  linear heat rate uncertainty factor of 1.01       due  to axial fuel densification and thermal expansion, 5..      A THERMAL POWER measurement        uncertainty factor of 1.02,   and
: 6.       A  rod bow penalty factor       of 1.05.
7        An   interim flow ad]ustment factor of 1.10 is to be implemented.
if the
               ~
reactor coolant flow rate is less than 370,000 gpm and greater than 354,000 gpm.*
ST. LUG IE  UNIT        1                       3/4 2-2
                      *Interim specification pending reanalysis.
4/Z7/76

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FLOW    IS <370,000 Ci                                                                 BUT    >354,000 GPH
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0.4 a8              W.6            Q.4                 W.2                    0            0.2                0.4                     0.6                 0.8 PERlPHERAL AXIALSHAPE tNOEX Figure 3.2.2 AXlALSHAPE fNOEX vs Fraction of Maximum Allowable Power Level per Specification 4.2.1.3 ST. LUCI E       -  UNIT  1                                 3/4 2-4 4/27/76

                               'A8LE 3.2-1 DNB MARGIN LIHITS Four Reactor Coolant Pumps Parameter                              ~0era tin Cold Leg Temperature                       < 542'F Pressurizer Pressure                       ~ 2225   psia*
Reactor Coolant Flow Rate                                > 354,000 gpm~*
AXIAL SHAPE INDEX                          Figure 3.2-4
 *Limit not applicable during either a THERMAL PO!tER ramp increase in excess of 5~ of RATED THERfiAL POMER or a THERtQL POlJER step increase of greater than 101. of RATED THERf1AL POWER.
**Interim va~ue pending reanalysis.
ST. LUCIE  - UNIT 1                3/4 2-13

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0.8    .             0.0            OA                   0.2             .          0                    W2 YI FIGURf 3.24 AXIALSHAP f INDEX Operating Limits v ith 4 Reactor Cooiant Pvmps Operating
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ST. l.UCIE             -    UNIT l.                                                  3/4 2-14

C 3/4.2    POllER DISTRIBUTION  LIMITS BASES 3/4.2.1     LINEAR HEAT RATE The   limitation. on linear heat rate ensures that in the event of a
  . LOCA,   the peak temperature of the fuel cladding will not exceed 2200'F.
Either of the two core power distribution monitoring systems, the
  'xcore Detector Monitoring        System and the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2.
In conjunction with the use of the excore monitoring system and in estab-lishing the AXIAL SHAPE 'INDEX limits, the following assumptions are made:
: 1) the CEA insertion limits of pecifications 3.1.3.2, 3.1.3.5 and 3.1.3.6 are satisfied, 2) the flux peaking augmentation factors are as shown in Figure 4.2-1, 3) the AZIMUTHAL POMER TILT restrictions of Specification 3.2.3 are satisfied,'. and 4) the TOTAL RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.2.
The  Incore Detector Monitoring System continuously provides a direct measure    of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that, the peak linear heat rates will be maintained within the allowable limits of Figure 3.2-1. The setpoints for these alarms include allowances, set in the conservative directions, f'r 1) flux peaking augmentation factors as shown in Figure 4.2-1, 2) a measurement-calculational     uncertainty factor of 1.10, 3) an engineering uncertai'nty factor of 1.03, 4) an allowance of 1.01 for axial fuel densification and thermal expansion, 5)'a THERMAL POHER measurement uncertainty factor of 1.02,      ,   6) a rod bow penalty factor of 1.05, and 7) an interim flow adjustment factor is to be implemented if the reactor coolant flow rate is (370,000 but >354,000 gpm.
~
3/4.2.2 and 3/4.2.3      TOTAL RADIAL PEAKING FACTOR - F  AND AZIMUTHAL POWER TILT - T The limitations on F and T are provided to ensure that the assump-tions   used in the analysiF for ektablishing the ONB Margin, Linear Heat Rate, Thermal Margin/Low Pressure and Local Power Density - High LCOs and LSSS  .setpoints remain valid during o~eration at the various allowable CEA group insertion limits. If either F or T exceed their basic limitations, operation may continue under the addftiona7 restrictions imposed by the ST. LUCIE    - UNIT  1                 8 3/4 2-1 4/27/76

SAFETY EVALUATION DNB  Safety Limit Figure 2.1-1 (Page 2-2)
Change 370,000     GPM  to 354,000 GPM  as indicated on Figure 2.1-1.
: b. Figure B2. 1-1 (Page B2-2)
Change the values of the rod radial peaks to those indicated on Figure B2.1-1.
Basis The  present Reactor Core Thermal Margin,.Safety Limit shown by Figure 2.1-1, page 2-2, represents a, locus of points corres-ponding to a limit of 1.3 DNBR (W-3) for a reactor coolant flow of 370,000 GPM and for the rod radial peaks and axial power distributions shown in Figure B2.1-1, page B2-2. If reactor coolant flow is decreased, there is a corresponding increase in the enthalpy rise across. the core. This results in an increase in the quality in the hot channels, and the limit of 1.3 DNBR is reached at a lower value of rod radial peak. The values of rod radial peak indicated on the revised Figure B2.1-1 have been reduced to a factor of 0.95 to accommodate a 5% reduction in the reactor coolant flow. Thus for the revised road radial peaks, and the axial power distributions in the revised Figure B2.1-1 and for the reduced reactor coolant flow of 354,000 GPM, the Thermal Margin Safety Limit in Figure 2.1-1, page 2-2,continues to represent a locus of points corresponding to a limit of 1.3 DNBR (W-3) .
: 2. Limiting Safety    System   Settings
: a. Table 2.2-1 (Page 2-4)
Change note on bottom     of page to read:
          "Design   reactor  coolant   flow with 4 pumps operating is 354/000 GPM."
: b. Table 2.2-1 (Page 2-4)
Item 2, Power Level-High, under Trip Setpoint and Allowable Values, change the maximum, limit from <106.5%
to  <96.5%  of Rated Thermal Power.
Basis Item (a) is modified to be consistent with the reduced reactor coolant flow rate. Item (b), the variable portion of the high power trip, remains at a maximum allowable settingof 10% above indicated power since this parameter is not sensitive to reactor coolant flow rate. The upper limit of the high power trip has been reduced to a value of 96.5% of Rated Thermal Power. This is consistent with an operating limit power level of 90% of Rated Power.

0 ~,
: c. Figure 2. 2-3 (Page 2-8)
Change p      P equation to" read:
var P var    =1388 x        +  12.5 x T.in  6150 DNB Figure 2.2-4 (Page 2-9)
Change p      p equation to read as specified    in item 2.c   above.
var Basis The Thermal Margin/Low Pressure        trip is  provided to prevent operation .when the DNBR (N-3) is       less than 1.30. The available margin to the limit of 1.30 DNBR        previously calculated for a wide range of core conditions .has been reduced to a factor of 0.95 to
'ccommodate a 5S reduction in the reactor coolant flow. The revised LSSS implement this reduction by modifying the computed value of pressure at which a reactor trip is initiated. This modification is shown in the ptrip equation in the revised Figures 2.2-3 and 2.2-4.             var The revised thermal margin LSSS provide at least the same margin of safety as that provided with the present LSSS.
: 3. Limiting Conditions for Operation
: a. Table 3.2-1 (Page 3/4 2-13)
Change  reactor coolant flow rate to read:       354,000  GPM
: b. Figure 3.2-4 (Page 3/4 2-14)
Change to implement the revised Figure 3.2-4.
: c. Figure 3.2-2 (Page 3/4'2-4)
Change to implement the revised Figure 3.2-2.
Basis Item 3.a. is modified to be consistent with the reduced reactor coolant flow rate. Item 3.b. refers to the Limiting Condition for Operation (LCO) on the axial shape index to maintain the required steady state operating margin to DNB. The available margin to a 1.3 DNBR calculated for a wide range of core con-ditions was reduced to a factor of 0.95 to accommodate a 5%
reduction in reactor coolant flow. The revised LCO was estab-lished consistent with this reduction. The transient analyses of the design basis LCO and the initial assumptions in the PSAR have been reviewed, and       it  was concluded that the revised LCO in Figure 3.2-4 provides at least the same margin of safety as that provided with the present LCO.

Item 3.c. refers to the LCO on axial shape index'to maintain the require steady-state operating margin on linear heat rate when the ex-. core detectors are used to monitor this limit. Since the limiting value of kw/ft has at present been established by the LOCA analysis, Figure 3.2-2 has been reduced to a factor of 1.10 to accommodate the 5% reduction in reactor coolant flow. The basis for this reduction is provided after item 3.e below.
: d. Specification 4.2.'1.4 (Page 3/4 2-2)
Add  item 7 under part b to read:
: 7. A flow adjustment factor of 1.10 is to be implemented if  the reactor coolant flow rate. is less than 370,000 GPM and greater tQan 354,000 GPM.
: e. Basis 3/4.2.1 (Page B3/4 2-1)
Add  to the paragraph on incore detectors the statement in modification 3.d. above.
Basis              St. Lucie 1 Reduced Flow LCOS Performance Results The results of a St. Lucie 1 break spectrum analysis, using the approved Combustion Engineering large break evaluation model are reported in Referencg 1. This analysis, which employed a system flow rate of 1)) 44 x 10 lbs/hr, demonstrated that the LOCA Accept-ance Criteria     'ere (PLHGR) of 15.8 kw/fta met at a peak linear heat. generation rate Subsequent to the analysis reported in Reference 1, a conserva-tive re-analysis has been performed to determine the allowable PLHGR when the system flow rate is reduced to 134.06 x 10         lbs/hr which is 4% less than the flow rate used in Reference 1. The allowable PLHGR at the reduced flow was found to be 15.6 kw/ft.
In order to determine sensitivity to linear heat rate, a second reduced flow case was examined, at a PLHGR of 14.2 kw/ft. The results of this case,     as   well as those discussed above, are summarized. below:
Worst Break*    LOCA Results 6
System Flow (X10      lb/hr)       139.44     134.06     134.06 PLHGR   (kw/ft)                     15.8       15.6       14.2 Peak Clad Temperature ( F)          2192   "
2189       1956 Peak Local Clad Oxidation(%) 10.42             9.71        5. 32 Peak Core-Wide Clad Oxidation <. 787 These results show that      all
(>)            '3)
                                                        <.787 LOCA acceptance criteria
                                                                    <.787 are met at. a PLHGR of 15.6 kw/ft, and there is substantial margin, at 14.2 kw/ft.
Discharge (0. 8   DEG/PD)

Based on these   considerations, (1) the proposed change does not increase the probability or consequences of accidents or mal-functions 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 signi-ficant hazards'onsideration, (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.
REFERENCES
: 1. Safety Xnjection System Analysis of St. Lucie 1 for FSA'R, Section 6.3.3.6 as amended by Revision. 55, 2-9-76.
: 2. CENPD-132, "Calculative Methods for the C-E Large Break LOCA Evaluation Model", August, 1974 (Proprietary).
CENPD-132, Supplement 1, "Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model," December, 1974 (Proprietary).
CENPD-132, Supplement   2, "Calculational Methods for the C-E Large Break LOCA Evaluation  Model',  July, 1975.
: 3. Acceptance Criteria for Emergency Core Cooling Systems for    .
Light-Water-Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3  Friday, January 4, 1974.

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