ML19224C977
| ML19224C977 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/25/1979 |
| From: | Stewart W FLORIDA POWER CORP. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 3--3-A-3, 3-0-3-A-3, NUDOCS 7907100393 | |
| Download: ML19224C977 (28) | |
Text
S Florida Power COHenHAT JN June 25, 1979 File: 3-0-3-a-3 Director Office of Nuclear Reactor Regulation Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Technical Specification Change Request No. 39
Dear Sir:
On May 25, 1979, Florida Power Corporation submitted revisions to the Cy-cle 2 reload Report which was filed on February 28, 1979 because FPC decided to license Crystal River Unit #3 at its present licensed power level of 2452 MWt instead of the Cycle 2 proposed power level of 2544 MWt.
This letter is to advise you that the Reactor Coolant Power Pump Mcnitors will not be installed prior to the initial Cycle 2 startup. As a result of this, the Technical Specifications submitted in support of Cycle 2 licens-ing, as amended on May 25, 1979, have been reviewed. The revisions to the proposed Cycle 2 Technical Specification, as submitted in TSCRN39 (Febru-ary 28, 1979), have been revised, deleting reference to the Reactor Coolant Power Pump Monitors and those specifications resulting from their use, de-leting the power upgrade level of 2544 MWt, and including the lowered RCS Pressure High setpoint as required by IE Bulletin 79-05B. The proposed Technical Specification pages attached are the actual cha.iges from our present Technical Specifications as a result of the work actually performed during the refueling outage.
If you or members of your staff require any further discussion of this sub-mittal, please contact us as soon as possible.
Sincerely, FLORIDA POWER CORPORATION
_ l U . h Dn0LO d W. P. Stewart rs 3E Manager, Nuclear Operations ,,3 ,[ r-O d1" TSCRN39(WPShewR02)D27 Attachment 7907100 3 9. 3 j d2 General Office 3201 Tn netounn sneet sou:n . P O Box 14042. St Petersburg Fiorda 33733 813 806 5151
a n STATE OF FLORIDA COUNTY OF PINELLAS W. P., Stewart states that he is the Manager, Nuclear Operations, of Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclecr Regulatory Conmission the information attached hereto; and that all such statements made and matters set forth tnerein are true and correct to the best of his knowledge, information and belief.
}, .
Q&GLL' W. P. Stewart Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 25th day of June,1579
~ Notary Public Notary Public, State of Florida at Large, My Commission Expires: August 24, 1979 (CrockettNotary 2 D12)
,<t- <, .
l 'I J
. s UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF )
) DOCKET No. 50-302 FLORIDA POWER CORPORATION )
C_ERTIFICATE OF SERVICE W. P. Stewart deposes and says that the following has been served on the Chief Executive of Citrus County, Florida by deposit in the United States mail, addressed as follows:
Chairman, Board of County Commissioners of Citrus County Citrus County Courthouse Inverness, Florida 32650 An original copy of revisions to Technical Specification Change Request No. 39 requesting amendment to Appendix A of Operating License No. DPR-72.
FLORIDA POWER CORPORATION f-11S.WoiLL W. P. Stewart Manager, Nuclear Operations SWORN TO AND SUBSCRIBED BEFORE ME THIS 25th DAY OF JUNE, 1979.
Notary Public Notary Public State of Florida at Large My Commission expires: August 24, 1979 (NOTARIAL SEAL)
(Cert. Serv.D12)r 0 /> T
Revisions to Technical Specification Change Request No. 39 for (1) 2452 MWt versus 2544 MWt; (2) deletion of the Reactor Coolant Power Pump Monitors; and (3) Item 5 of IE Bulletin 79-05B. Page 1-1: In Definition 1.3, RATED THERMAL POWER should be defined as 2452 MWt (as it is in the present Technical Specification) instead of 2544 MWt. Page 2-2: In Figure 2.1-1, the RCS Pressure-High Trip should be redrawn as 2300 psig instead of 2355 psig. Page 2-3: In Figure 2.1-2, the limit line for acceptable 3 & 4 pump operation should be lowered by 2.0% of RATED THERMAL POWER. Page 2-5: In Table 2.2-1, in Item 2, the Trip Setpoint and Allowable Value with three pumps operating should be changed to 173% of RATED THERMAL POWER (as it is in the present Technical Speci-fication) instead of 179.9%. In Table 2.2-1, in Item 6, the Trip Setpoint and Allowable Value should be changed to 12300 psig instead of 12355 psig. Page 2-6: In Table 2.2-1, delete Item 8 referring to the Reactor Coolant Pump Power Monitors and i evise Item 9 to Item 8 (as it is in the present Technical Specifications). Page 2-7: In Figure 2.2-1, change the Trip Setpoint envelopes fer accep-table 4 and acceptable 3 & 4 pump operation to the envelopes that are in the present Technical Specifications. Page B2-5: In the Bases for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance, change Items 1 and 2 of the second paragraph to the Items 1 and 2 that are in the present Technical Specifications. Page B2-6: In the first paragraph the flux-to-flow ratio should be changed to 1.043% (as it is in the present Technical Specifi-cations) instead of 1.07%. In the Bases for RCS Pressure-Low, High and Variable Low, in the second paragraph, change the trip setpoint for RCS Pressure-High to 2300 psig instead of 2355 psig. Page B2-8: For Curves 1 and 2, change the Power (% RTP) to 117.3% and 90.5%, respectively, instead of 113.1% and 87.2%, respectively. P? ' TSCRN39(WPShewR02)D27 ') Li g ]F d 'r v
Page 3/4 1-27: In Figure 3.1-1, change the Power Level Cutoff to 92% of Rated Thermal Power instead of 90%. Change the length of mid-Lvcle 2 to 233fl0 EFPD instead of 225110 EFPD. Page 3/4 1-28: In Figure 3.1-2, change the Power Level Cutoff to 92% of Rated Thermal Power instead of 90%. Change the length of mid-Cycle 2 to 233110 EFPD instead of 225110 EFPD. Page 3/4 1-29: Change the length of mid-Cycle 2 to 233110 EFPD instead of 225110 EFPD. Page 3/4 1-30: Change the length of mid-Cyucle 2 to 233110 EFPD instead of 225110 EFPD. Page 3/4 1-38: C.;ange the length of mid-Cycle 2 to 233110 EFPD instead of 225110 EFPD. Page 3/4 1-39: Change the length of mid-Cycle 2 to 233110 EFPD instead of 225110 EFPD. Page 3/4 2-2: Change the length of mid-Cycle 2 to 233+10 EFPD instead of 225110 EFPD. Page 3/4 2-3: In Figure 3.2-2 the point defined as (18, 92) should be changed to (18, 90). Change the length of mid-Cycle 2 to 233+1C EFPD instead of
~~
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CURVE REACTOR COOLANT FLOW (lb/hr) 1 139.86 x 106 2 104.47 x 106 FIGURE 2.1-2 REACTOR CORE SAFETY LIMIT
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2.1 SAFETY LIMITS BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel clad-ding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling rgime wb're the heat transfer coefficient is large and t% cladding surface temperature is slightly above the coolant saturation tem;a 'ature. Operation above the upper boundary of the nucleate boiling regime would result in excessive cladding temperatures because of the onset of departure f rom nucleate boiling (DNB) and the resultant sharp reduction in heat trans-fer coefficient. DNB is not a directly measurable parameter during opera-tion and therefore THERMAL POWER and Reactor Coolant Temperature and Pres-sura have been related to DNB through the BAW-2 DNB correlation. The DNB correlation has been developed to predict the DNB flux and the location of DNB for axially unifonn and non-uniform heat flux diatributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indica-tive of the margin to DNB. The minimum value of the DNBR during steady state operation, normal opera-tional transients, and anticipated transients is limited to 1.30. Thi s value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum gossible thermal power, 112%, when the reactor coolant flow is 139.86 x 10 lbs/hr, which is 106.5% of the design flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors with potential fuel densification effects: FN = 2.57; FN = 1.71-, FN = 1.50 Q AH Z The design limit power peaking factors are the most restrictive calculated at full power for the range from all control mds fully withdrawn to minimum allowable control md withdrawal, and form the core DNBR design basis. CRYSTAL RIVER - UNIT 3 B 2-1 bJ
LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in onder to prevent re-actor thermal limits from being exceeded. These thermal limits are either power peaking kw/ft limits or DNBR limits. The AXIAL POWER IMBALANCE reduc-es the power level trip produced by the flux-tc~ flow ratio such that the boundaries of Figure 2.2-1 are produced. The flux-to-flow ratio mduces the power lev 0 trip and associded mactor power-reactor power-imbalance bound-aries by 1.043% for a 1% flow ret.ction. RCS Pressure - Low, High and Variable Low The High and Low trips are provided to limit the pressure range in which re-actor operation is pennitted. During a slow reactivity insertion startup accident from icw power or a slow reactivity insertion from high power, the RCS Pressure-High setpoint is reached before the Nuclear Overpower Trip Setpoint. The trip setpoint for RCS Pressure-High, 2300 psig, has been established to maintain the system pressure below the safety 1:mit, 2750 psig, for any design transient. The RCS Pressure-High trip is backed up by the pressurized code safety valves for RCS over pressure protection, and is therefore set lower than the set pressure for these valves, 2500 psig. The RCS Pressure-High trip also backs up the Nuclear Overpower trip. The RCS Pressure-Low,1800 psig, and RCS Presssure-Variable l'w. (11.80 Tout F-5209.2) psig, Trip Setpoints have been established to maintain the DNB ratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction. It also prevents reactor operation at pres-sures below the valid range of DNB correlation limits, prote"ing against DNB. Due to the calibration and instrumentation errors, the safety analysis used a RCS Pressure-Variable Low Trip Setpoint of (11.80 Tout F-5249.2) psig. Reactor Containment Vessel Pressure - High The Reactor Containment Vessel Pressure-High Trip Setpoint 14 psig, provides positive assurance that a reactor trio will occur in the unlikely event of a steam line failure in the containmer vessel or a loss-of-coolant accident, even in the absence of a RCS Pressure-Low trip. CRYSTAL RIVER - UNIT 3 B 2-6 n b' 7r a w J U .r) ' ,
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gs rMv.84 v. e > t: Sb!pf a REACTOR C00LAtiT FLOW PUMPS OPERATING CURVE, FLOW (lb/hr) POWER (%RTP) (TYPE OF LIMIT) 1 139.86 x 106 (106./%) 117.3% 4 Pumps (DriBR) 2 104.47 x 106 (79.7%) 90.5% 3 Pumps (DriBR) PRESSURE / TEMPERATURE LIMITS AT MAXIMUM ALLOWABLE POWER FOR MINIMUM DriBR BASES FIGURE 2.1 CRYSTAL RIVER - UtlIT 3 B 2-8 3'{- nu;i J
I10 ~' (180,102) (225,102) 100 Power Level Cutoff (92% UNACCEPTABLE of Rated Thermal Power) OPERATION _ (lao,90) i , (225,90)
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I10 (270,102) 100 _ Power Level Cutoff (92% : 90 _ of Rated Thermal Power) 7 0 - u (255,80) O 70 - is E i; UNACCEPTABLE g 60 - OPERATION T . a: 50 ACCEPTABLE (175,50) OPERATION o
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c' e 30 - 20 - 10 - (0,0) 0 , , . . i , i , ' ' ' 0 50 !00 150 200 250 300 Rod Index, % Withdrawn 0 25 60 75 100 0 25 50 75 1 1 ! I f f 1 I f f I Ig0 Group 5 Group 7 0 25 50 75 100 1 I i I t ' I Group 6 FIGURE 3.1-2 REGULATING ROD GRu INSERTION LIMITS FOR 4 PUMP OPERATION ,,, ER 233 + 10 EFPD
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100 90 - (247.O.76.5) 80 - (158.3.76.5) u 70 - y UNACCEPTABLE E OPERATION - 60 - E 5 (300,50 0) ' r 50 - (100,50) u $ ACCEPTABLE 0 - [ OPERATION i 30 - 2 20 20 - (0,0) 0 ' ' ' ' ' ' ' ' ' ' ' 0 25 50 75 I00 125 150 175 200 225 250 275 300 Rod index, % Withdrawn 0 75 100 0 25 50 75 100
, 2, 5 5,0 i i i e i i Group 5 Group 7 0 25 50 75 100 f i ! ! 1 Group 6 FIGURE 3.1-3 REGULATING R0D GROUP INSERTION LIMITS FOR 3 PUMP OPERATION FROM 0 EFPD TO 233 + 10 EFPD 32> 5 ra.,vr: .
CRYSTAL RIVER - UNIT 3 3/4 1-29
100 90 - 80 - UNACCEPTABLE OPERATION (246,76.5) u i 70 - ? E 60 - E 50 - (175,50) 3 g LIO _ r.< J 30 . y ACCEPTABLE E OPERATION 20 - 10 - 0 i i i . e i i ' ' ' e 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod index, f. Withdrawn 0 2,5 5,0 7,5 100 0 25 50 75 100 Group 5 Group 7 0 25 50 75 100 l I I I Group 6 FIGURE 3.1-4 REGULATING R0D GROUP INSERTION LIMITS FOR 3 PUMP OPERATION AFTER 233 + 10 EFPD d l.- ?S [ G r; CRYSTAL RIVER - UNIT 3 3/4 1-30 ~
REACTIVITY CONTROL SYSTEMS ROD PROGRAM LIMITING CONDITION FOR OPERATION 3.1.3.7 Each control rod (safety, regulating and APSR) shall be programmed to operate in the core position and rod group specified in Figure 3.1-7. APPLICABILITY: MODES 1* and 2*. ACTION: With any control rod not programmed to operate as specified above, be in HOT STANDBY within 1 hour. SURVEILLANCE REQUIREMENTS 4.1.3.7
- a. Each control rod shall be demonstrated to be programmed to operate in the specified core position and rod group by:
- 1. Selection and actuation from the control room and verifica-tion of movement of the proper rod as indicated by both the absolute and relative position indicators:
a) For all control rods, after the control rod drive patches are locked subsequent to test, reprogramming or maintenance within the panels. b) For specifically affected individual rod s , fol l owi ng maintenance, test, reconnection or nodification of power or instrumentation cables from the control rod drive control system to the control rod drive.
- 2. Verifying that each cable that has been disconnected has been properly matched and reconnected to the specified control rod d ri ve.
- b. At least once each 7 days, verify that the control rod drive patch panels are locked.
*See Special Test Exceptions 3.10.1 and 3.10.2.
CRYSTAL RIVER - UNIT 3 3/4 1-33 325 e :, o a s;
Q @ @ @
@ @ Q GROUP NUMBER OF RODS FUNCTION l 8 SAFETY 2 8 SAFETY 3 12 SAFETY 4 9 SAFETY 5 8 CONTROL 6 8 CONTROL 7 8 CONTROL 8 8 APSRs TOTAL 69 FIGURE 3.1-7 CONTROL R00 LOCATIONS AND GROUP ASSIGNMENTS CRYSTAL RIVER - UNIT 3 3/4 1-34
D @ DELETED CRYSTAL RIVER - UNIT 3 3/4 1-35 jc>:c> p a ,,.
REACTIVITY CONTROL SYSTEMS XENON REACTIVITY LIMITING CONDITION FOR OPERATION 3.1.3.8 THERMAL POWER shall not be increased above the power level cutoff specified in Figures 3.1-1 and 3.1-2 unless xenon reactivity is within 10 percent of the equilibrium value for RATED THERMAL POWER and is approaching stability. APPLICABILITY: MODE 1. ACTION: With the requirements of the above specification not satisfied, reduce THERMAL POWER to less than or equal to the power level cutoff within 15 minutes. SURVEILLANCE REQUIREMENTS 4.1.3.8 Xenon reactivity shall be determined to be within 10% of the equilibrium value for RATED THERMAL POWER and to be approaching stability prior to increasing THERMAL POWER above the power level cutoff. CRYSTAL RIVER - UNIT 3 3/4 1-36 b ' N !; f ' av.
REACTIVITY CONTROL SYSTEMS AXIAL POWER SHAPING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.9 The axial power shaping rod group shall be limited in physical insertion as shown on Figures 3.1-9 and 3.1-10. APPLICABILITY: MODES 1 and 2*. ACTION: With the axial power shaping rod group outside the above insertion limits, either:
- a. Restore the axial power shaping rod group to within the limits within 2 hours, or
- b. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figure within 2 hours, or
- c. Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.9 The position of the axial power shaping rod group shall be deter-mined to be within the insertion limits at least once every 12 hours.
*With keff >1.0.
CRYSTAL RIVER - UNIT 3 3/4 1-37
- n r-( ')
.!! ,I iz g
T I10 (7,102) (31,102) 100 - 90 -
! UNACCEPTABLE 5 OPERATION 3 (39,80) o_ go , (o,so)
~
m 2 2 70 - y 60 - O
" 50 - (100,50) t ACCEPTABLE
) 40 - OPERATION 30 -
20 - 10 - 0 i i e i i e i i 1 30 40 50 60 70 80 90 100 0 10 20 Rod Position , % Withdrawn FIGURE 3.1-9 AXIAL POWER SHAPING ROD GROUP INSERTION LIMITS FROM 0 EFPD TO 233 + 10 EFPD CRYSTAL RIVER - UNIT 3 3/4 1-38 -) [ h t o-
110 (8,102) (34,102) 100 - 00
~
'(8,90) ,
(34,90) UNACCEPTABLE
.0PERATION 80 , (0,80) Hs,80) b 8
- o. 70 -
%e }
60 - E g 0 - I ,50 ' ACCEPTABLE q OPERATION wt 40 - 5 [ 30 - 20 10 0 i i i i t t t i t 0 10 20 30 40 50 60 70 80 90 100 Rod Position, 7. Withdrawn FIGURE 3.1-10 AXIAL POWER SHAPING ROD GROUP INSERTION LIMITS AFTER 233 + 10 EFPD CRYSTAL RIVER - UNIT 3 3/4 1-39 ' bb(
Power, % of Rated Thermal Power (-10.2.102) (10.2,102)
~
--- 100
(-15.3,90)
-- 90 ' ' (' * ' }
(-25.4,80) - 80 (12,80) 70 ACCEPTABLE UNACCEPTABLE OPERATION UNACCEPTABLE OPERATION - 60 OPERATION __ 50 E __ 40 y 2 o 30 $ t __ 20 0 W _- 10 i f I t ! I l 1
-40 -30 -20 10 0 10 20 30 40 Axial Power imbalance, %
FIGURE 3.2-1 AXIAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 EFPD TO 233 + 10 EFPD CRYSTAL RIVER - UNIT 3 3/4 2-2 b2 5 e,e ,a-
Power, 7. of Rated Thermal Power (-19.3,lu2) (l8.3,102)
--100
(-20.7,90) -- 90
* (18,90)
(-29,80) 4
- 80 > (19.6,80)
ACCEPTABLE -- 70 OPEr.4 TION
~~
UNACCEPTABLE OPERATION 50 __ 'O 5 __30 Cp UNACCEPTABLE i5 OPERATION 82
- 20
-- 10
, i , i , i i i i i
-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, 7
' IGURE 3.2-2 AXIAL POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 233 + 10 EFPD CRYSTAL RIVER - UNIT 3 3/4 2-3 m ,'j 1; L U '>
POWER DISTRIBUTION LIMITS NUCLEAR HEAT FLUX HOT CHANNEL FACTOR - Fn LIMITING CONDITION FOR OPERATION 3.2.2 Fg shall be li nited by the following relationships: Fg i 3.08 P TH M M where P = RATED THERMAL POWER a nd P < 1. 0. APPLICABILITY: MODE 1. ACTION: With Fg exceeding its limit:
- a. Reduce THERMAL POWER at least 1% for each 1% Fg exceeds the limit within 15 minutes and similarly reduce the Nuclear Overpower Trip Setpoint and Nuclear Overpower based on RCS Flow and AXIAL POWER IMBALANCE Trip Setpoint within 4 hours.
- b. Demonstrate through ir are mapping that Fg is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours.
- c. Identify and ccrrect the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that Fg is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours after attaining 95% or greater RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.2.1 Fg shall be determined to be within its limit by using the in-core detectors to obtain a power distribution map: CRYSTAL RIVER - UNIT 3 3/4 2-4
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TABLE 3.2-2 QUADRAT!T POWER TILT LIMITS STEADY STATE TRANSIENT MAXIMUM LIMIT LIMIT LIMIT QUADRANT POWER TILT as Measured by: Symmetrical Incore Detector System 3.46 8.96 20.0 Power Range Channels 1.96 6.96 20.0 Minimum Incore Detector System 1.90 4.40 20.0 CRYSTAL RIVER - UNIT 3 3/4 2-11 j p r, r ,; t,i ,v
TABLE 3.2-1 DNB MARGIN LIMITS Four Reactor Three Reactor Coolant Pumps Coolant Pumps Parameter Operating Operating Reactor Coolant Hot Leg Temperature, TH 'F < 604.6 < 604.6(1) Reactor Coolant Pressure, psig.(2) > 2061.6
> 2057.2(1)
Reactor Coolanc Flow Rate, lb/hr > 139.86x106 > 104.47x106 (1) Appplicable to the loop with 2 Reactor Coolant Pumps Operating. (2) Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step increase greater than 10% of RATED THERMAL POWER. 325 0.': CRYSTAL RIVER - UNIT 3 3/4 2-13
POWER DISTRIBUTION LIMIT 5 BASES FN Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio AH of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power. It has been determined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking and on minimum DNBR at full power are met, providea: Fg ~< 3.08; FN < 1.71 l AH ~ Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking. It has been determined that the above hot channel factor limits will be met provided the following conditions are maintained.
- 1. Control rods in a single group move together with no individual rod insertion differing by more than + 6.5% (indicated position) from the group average height.
- 2. Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6.
- 3. The regulating rod insertion limits of Specification 3.1.3.6 and the axial power shaping rod insertion limits of Specificaticn 3.1.3.9 are maintained.
- 4. AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE is a measure of the difference in power between the ton and bottom halves of the core. Calculations of core average axial peaking factors for many plants and measurements from operating plants under a variety of operating conditions have been correlated with AXIAL POWER IMBALANCE.
. The correlation shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCE is maintained within the limits of Figures 3.2-1 and 3.2-2.
The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod insertion and are the core DNBR design basis. Therefore, for operation at a fraction of RATED THERMAL POWER, the design limits are met. When using incore detectors to make power distribution maps to determine Fg and F N , AH
- a. The measurement of total peaking factor, Fheas , shall be increased by 1.4 percent to account for manufacturing tolerances and further increased by 7.5 percent to account for measurement error.
CRYSTAL RIVER - UNIT 3 8 3/4 2-2 s r n, 3b3 o. ..}}