ML20095J512
| ML20095J512 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 12/21/1995 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20095J511 | List: |
| References | |
| GL-88-16, NUDOCS 9512270142 | |
| Download: ML20095J512 (90) | |
Text
{{#Wiki_filter:... . - - . . _ . . _ _ _ ._. l ATTACHMENT B l l BRAIDWOOD STATION l Proposed Changes to Appendix A , Technical Specifications of facility l Operating Licenses NPF-72 and NPF-77 l 1 Revised Pages: IV V 1-4 3/4 1-1 l I 3/4 1-14 3/4 1-15 3/4 1-20 3/4 1-21 3/4 1-22 3/4 2-1 l 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-8 B 3/4 2-2 6-22 l 6 ._ _ . PDR ADOCK 05000454 P. . . ,_
,PDR l
l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l 1 1 PAGE ! SECTION 3/4 0-1 3/4.0 APPLICABILITY............................................... .i l ! 3/4.1 REACTIVITY CONTROL SYSTEMS 1 i 3/4.1.1 BORATION CONTROL Shutdown Margi n - T,,, > 200'F. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-1 Shutdown Margin - 3/4 1-3 T,,, 1 200'F........................... Moderator Temperature coef fi cient. . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-4 4 Mi nimum Temperature for Critica11ty. . . . . . . . . . . . . . . . . . . . . . 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7 i
- Flow Paths - 0perating...................................
3/4 1-8 i Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - 0perating............................... 3/4 1-10 Borated Water Source - Shutdown. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-11 Borated Water Sources - 0perati ng. . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-12 Boron Dilution Protection System. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-13a , 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN IN0PERABLE FULL-LENGTH R00. . . . . . . . . . . . . . 3/4 1-16 ! Position Indication Systems - Operating. . . . . . . . . . . . . . . . . . 3/4 1-17 Position Indication System - Shutdown. . . . . . . . . . . . . . . . . . . . 3/4 1-18 Rod Drop Time............................................ 3/4 1-19 3/4 1-20 Shutdown Rod Insertion Limit............................. Control Rod Insertion Limits............................. 3/4 1-21 FIGURE 3.1-1 "00 """ !MSE" TION LIMITS VE" SUS-THERMAE m' " TOU" LOO" 0"C"JTION.(.Twd.f.%.M.C.>.W.5.d.WN 3/4 1-22 l IV Amendment No. 30' BRAIDWOOD - UNITS 1 & 2
4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l SECTION PAGE j 3/4.2 POWER DISTRIBUTION LIMITS ,
, 3/4.2.1 AXIAL FLUX DIFFERENCE ................................... 3/4 2-1 l
, FIGURE 3.2-1 AXIAL FLUX DIFFERENCE 4IMITS AS A FUNCTION-04 RATED !"E"".".L "Oh'ER. . ( .IW '.>. E r.w.9 . .'.>. h'.T. .d; D/) . 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R............................. 3/4 2-4
..u,, e= w c ac m FIGURE 3.2-2 K(Z,)-NORMALIZED-F f0NCT-I 3/4 2-5 l q (Z)-A
' OF-GORE-HE-IGHT. . .
3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL
! FACT 0R................................................. 3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-10 i
3/4.2.5 DNB PARAMETERS........................................... 3/4 2-13 TABLE 3.2-1 DNB PARAMETERS........................................ 3/4 2-14 3/4.3 INSTRUMENTATION i 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TARE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 TABLE 3.3-2 (THIS TABLE IS NOT USED).............................. 3/4 3-7 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE l , REQUIREMENTS........................................ 3/4 3-9 I 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATI0ii SYSTEM INSTRUMENTATION........................................ 3/4 3-13 j TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM ! INSTRUMENTATICN..................................... 3/4 3-15
- TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i INSTRUMENTATION TRIP SETP0lNTS...................... 3/4 3-23 l TABLE 3. M, (THIS TABLE IS NOT USED).............................. 3/4 3-30 l TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM J
- INSTRUMENTATION SURVEILLANCE REQUIREMENTS........... 3/4 3-34
!', ; J.. .. W u _ : _ ;c BRAIDWOOD - UNITS 1 & 2 V
1 DEFINITIONS OFFSITE DOSE CALCULATION MANUAL l t 1.18 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alare/ trip setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. The 00CM shall also contain (1) the i Radioactive Effluent Controls and Radiological Environmental Monitoring
- Programs required by Sections 6.8.4.e and f, and (2) descriptions of the infomation that should be included in the Annual Radiological Environmental
- Operating and Radioactive Effluent Release Reports required by Specification ~
i 6.9.1.6 and 6.9.1.7. ( OPERABLE - OPERABILITY $ 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of perfoming its specified function (s), and , when all necessary attendant instrumentation, controls, electrical power, i cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s). (. , < - OPERATING LIMITS REPORT
%(gg{
1.19.a The OPERATING LIMITS REPORT'is the unit-specific document that provides l 2 operating limits for the current operating reload cycle. These cycle-specific operating limits shall be detemined for each reload cycle in accordance with 4 Specification 6.9.1.9. Plant Operation within these operating limits is addressed in individual specifications. l OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive l combination of core reactivity condition, power level, and average reactor
- coolant temperature specified in Table 1.2.
i PHYSICS TESTS l.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR
- 50.59, or (3) otherwise approved by the Commission.
4 PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall. i 1
) BRAIDWOOD UNITS 1 & 2 1-4 AMENDMENT NO. +9-
.- . - _ - - - -._ - _..-_. .=. ._ . - . . _ - . - - - - _ . _ .
i
- REACTIVITY CONTROL SYSTEMS ,
I l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES i l GROUP HEIGHT l l LIMITING CONDITION FOR OPERATION 3.1. 3.1 All full-length shutdown and control rods shall be OPERABLE and 1 positioned within i 12 steps (indicated position) of their group step counter j demand position.
- APPLICABILITY
- MODES 1* and 2*.
i ACTION:
- a. With one or more full-length rods inoperable due to being immovable l
as a result of excessive friction or mechanical interference or i known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in 4 HOT STANDBY within 6 hours. l b. With one full-length rod trippable but inoperable due to causes 4 - other than addressed by ACTION a. above, or misaligned from its i group step counter demand height by more than i 12 steps (indicated i
- position), POWER OPERATION may continue provided that within 1 j hour
- 1. The rod is restored to OPERABLE status within the above alignment requirements, or i 2. The rod is declared inoperable and the remainder of the rods in i the group with the inoperable rod are aligned to within
- t 12 steps of the inoperable rod while maintaining the rod j i sequence and insertion limits of fi p rc 3. M . The THERMAL POWER i level shall be restricted pursuant tn_Jpecification 3.1.3.6 during subsequent operation, or ( f4 u.us, g ,
- 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER j OPERATION may then continue provided that
5 a) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. l 3 b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours; i
! *See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
i i i BRAIDWOOD - UNITS 1 & 2 3/4 1-14 Amendment No. 22 4 #
--. u._.
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incoredetectorsandF(Z)andFhareverifiedtobe q within their limits within 72 hours; and d) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions;
- c. With more than one full-length rod trippable but inoperable due to e causes other than addressed by ACTION a. above, or misaligned from its group step counter demand height by more than + 12 steps (
(indicated position), POWER OPERATION may continue provided that: ,
\
- 1. Within 1 hour, the remainder of the rods in the group (s) with the inoperable rods are aligned to within + 12 steps of the inoperable rods while maintaining the rod sequence and
- insertion limits of fiiin 3.1-1 4The THERMAL POWER level l shall be restricted pursuan f (
subsequent operation, o and (gto,_Sgpcificationl1Jgurin L, sq.g v- % % _ y , ,, , g
- 2. The inoperable rods shall be restored to OPERABLE status within ;
72 hours. Otherwise, be in HOT STANOBY within 6 hours. i i SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be . I within the group demand limit by verifying the individual rod positions at l least once per 12 hours except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least onco , per 4 hours. 1 . 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined OPERABLE by movement of at least 10 steps in any one direction at i least once per 31 days. BRAIDWOOD UNITS 1 & 2 3/4 1-15 Amendment No. 3I l
4 4 REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT i E J LIMITING CONDITION FOR OPERATION
'\,y, % sn pnw :ma +.w- -. e m en .x. s , - 4 6 +_
j; 3 gmew s, _t_sta g. WEkhT . I 3.1.3.5 All sfiutdown rods 7 aTT'Eii'f;11, .ithi e.n T "~ l i 1 APPLICA8ILITY: MODES 1* and 2*#.
"' '~'
ACTION ** e-
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except for surveillance
- With a maximum of one shutdown rod not fully withe l testing pursuant to Specification 4.1.3.1.2, within 1 hour either
- a. .pfully 4thdraw-the-reW- l
! b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
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- 4.1.3.5 Each shutdown rod shall be determined 9ully withdeawn:
l a. Within 15 minutes prior to withdrawal of any rods in Control l Bank A, B, C, or 0 during an approach to reactor criticality, and j b. At least once per 12 hours thereafter.
"See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
! #With K,ff greater than or equal to 1. b 4 i BRAIDWOOD - UNITS 1 & 2 3/4 1-20
-x.,_s_ -
s
REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1. 3. 6 The control. banks shall be lij!tited in phys.ical insertion,as eer- '^ rigere 3.1-1. ~; 'm. .'n u s ' m dw.. r&s3b . I--wd rs ' d.IIV c n 7 7
.. - ~ w - :-
APPLICABILITY: MODES 1* and 2*#. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:
- a. Restore the control banks to within the limits within 2 hours, or
- b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the using the abec; 'igure, o^ [T,Ee.1 Twig f{-3 68, ., bank,ESposition L ,e v e n r % M .4 t-,,w .:. r:c m s- , ;,'
;. C . . . . . '. I W a i MCT ",IANCCY w i Uf'e'Us~o Niu D '" ~"
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Alarm is inoperable, then verify the individual rod positions at least once per 4 hours.
- See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
- With K,7f greater than or equal to 1.
BRAIDWOOD - UNITS 1 & 2 3/4 1-21 r-n _L - . r * ~
- M G v is. S.\ -\
LT. w v., Fsemet a u v O -e s N 228' r ";i _;"/(79%, 228)--/ 220
- (29%,228) +"~~Z
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' = ;i 200 --
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t' RELATIVE POWER (Percent) R GURE 3.1-1 ROD-BAN K--I NSE RTION-LI MI-TS-VER SUS _THE RMA L_POWE R-mien Pt/Un auvv nno. a'rcJn r* D. , AT. T. A. M. BRAIDWOOD - UNITS 1 & 2 3/4 1-22 .. 3 .; , ,, 7.. - (
I 3/4.2 POWER O!STRIBUTION LIMITS _, _ , _ . . , - <f- 73 70% N MM ^ W " 3/4.2.1 AXIAL FLUX DIFFERENCE in Q W%W % L.\ M st S LIMITING CONDITION FOR OPERATION W s.J j l I 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the . , 4o4 lowing target band (flux difference units) about the target flux difference: P
- a. - 5% 'er Cycle-4-core-aver-age-accumulated-burnup of less-them equ:1 te 5000-MWO/MT%--and -
- b. ^ M, -M for Cyc44-1-core-average accumulated-burnup-of-greater then
-5000 "h'0/MTU, :nd
- c. -
3', -12* for c:ch-+ubsequent cyc!c. The indicated AFD may deviate outside the about required target band at greater ! l than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limitsyf ri;;ure 2.T 1 and the l cumulative 24 hours. penalty deviation
- p %,D timehdoes_not d C exceed 1)our tkT2hTgb M4NMduring_the_previo[us iihP T3, a ~--v v o m ;
The indicated AFD may deviate outside the abous-required target band at greater > than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed 1 hour durino the orevious 24 hours. APPLICABILITY: MODE 1 above 15% of RATED THERMAL POWER *. ACTION:
- a. With the indicated AFD outside of the-above required target band and l with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes, either:
- 1. Restore the indicated AFD to within the above required target band limits, or
- 2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.
I
- b. With the indicated AFD outside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits e4 4 l Figure 3.2J. and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce:
- 1. THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and l
- 2. The Power Range Neutron Flux - High Setpoints to less than or equal ,to ,55% of RA ,TED TH.ERMAL
- -, WER ne within the,nex,t_4 ~chours. , _
,C G
'gted wd m M UIt?ATWO "^ LinWG
- ~ w l'E\TQ_;
l "See Special Test Exceptions Specificat'on 3.10.2. M Surveillance testing of the Power Rang e Neutron Flux channel may be performed l pursuant to Specification 4.3.1.1 prov ided the indicated AFD is maintained witnin the Acce: table Operation LimitsV0 f A gure 3.2-1. A total of 16 hours operation may be accumulateo with the AFD outside of the asoaa required target bar.c during testing without penalty deviation.
- v - N ! L: 3/4 2-1 p .j ggg - (.7
l [ LIMITING CONDITION FOR OPERATION i l ACTION (Continued) , c. With the indicated AFD outside of the above-required target band for l more than 1~ hour of cumulative penalty deviation time during the 4 previous 24 hours and with' THERMAL POWER less than 50% but greater 4 than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be
- increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the above. required target band. l 1
SURVEILLANCE REQUIREMENTS [ 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel
! 1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, l and
- 2) At least once per hour for the first 24 hours after restoring
, the AFD Monitor Alarm to OPERABLE status.
- b. Monitoring and logging the indicated AFD for each OPERABLE excore i channel at least once per hour for the first 24 hours and at least once per 30 minutes.thereafter, when the AFD Monitor Alarm is
- . inoperable. The logged values of the indicated AFD shall be f assumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band l shall be accumulated on a time basis of:
- a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and .
- b. One-half minute penalty deviation for each 1 minute of POWER OPERATION '
outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER. 4.2.1.3 The initial determination of target flux difference following a refuel- . . ing outage shall be based on design predictions. Otherwise, the target flux ~ difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. g 4.2.1.4 The target flux difference shall be updated at least once per ) 31 Effective Full Power Days by either determining the target flux difference ; pursuant to Specification 4.2.1.3 above or by linear interpolation between the I most recently measured value and the predicted value at the end of the cycle
' life. . .
BRAIDWOOD - UNITS 1 & 2 3/4 2-2 AMENDMENT NO. g
\ . I l . l 1
5 t i FIGURE 3.2-1 l (.3 w sS i: s t4 u n.s vs W C' T u%Q
. AXIAL-stuX O!FrEREMCE '_IMITS a.S ^
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8' '*" 100 -- t 4 i i UNACCEPTAKE5611,90)._ ACCEPTABLE l WERATION \ ~ ;. - - :EiEE(11,90)5pOPERATION C -- s W - Z -i --
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1 80 - . i ACCEPTABLf (OPERATION,k -.- a -- - i -
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'- 631,50)i
;(31,50) 40 l ._._-
i y . - 6
/ 80 40 30, 20 10 0 10 20 30 40 M FLUX DIFFERENCE (All % \
BRAIDWOOD-UNITSli2 3/4 2-3 huevt> gg - (, I
D. :
. I I l I
I 3/4.2.2 HEATFLUXH0TCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 Fq (Z) shall be limited by the following relationships:
, - , . -3 (
2.021 Fq (Z) '< (T [X(2)] for P > 0.5, and (Unit-1 Cycle-2-- f {pS] ^ *nd-Unit 2 Cyc4e-L)- , Ch .
% '~! - / F (Z)q 5 [4 443 [K(Z)] for P < 0.5.
f' ;LE l
' h80MX(1)-) for * ' O.5, end4 Unit-1-Gycle-3-end 4f-ten-Uni-t-2-Cycle 2-i; l
L IS,~c.? _l F9 (Z) ~' [T ' odd
'A -
J g(I).t{.5,00h[X(Z)3-for-P--f4r5, : Where: , p , THERMAL POWER RATED THERMAL POWER
*nd-K(Z) is the function obtained #= eigure 3.2-2 for a given ,--- - ,-~~
core height location. '/g p o- . . . . .
- -. 7- _
APPLICABILITY: MODE 1. TD C ~ N' ' "
- NM'" _"m_O C" .-~
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)
ACTION: : With F (Z) exceeding its limit: q
- a. Reduce THERMAL POWER atqleast 1% for each 1% F f
within 15 minutes and similarly reduce the Power ' may proceed for up to a total of 72ower hours;AT Tripsubsequent 5etpoints POWER l j OPERATION may proceed provided the Overp% F (Z) exceeds the limit; q < have been reduced at least 1% for each 1 ) and
- I l
' b.
Identify and correct the cause of the out-of-limit condition prior l to increasing THERMAL POWER above the reduced limit q required by ; f ACTION a., above; THERMAL POWER may then be increased p l is demonstrated through incere mapping to be within its limit. I t (-,-~,--~,.,,.,.___,,, _ AC Ah ~IN Y l%D!. p. L 'f; *. . t?T t '
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I I i Amendment No. 23 3/4 2-4 BRAIDWOOD - UNITS 1 & 2
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e e o w n m e w e e w w if I (210d camVWWON -(ZlM 4 4 1 I t i 3/4 2-5 AmendmentNo.%
- BRAIDWOOD - UNITS 1 & 2
POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR
-LIMITING CONDITION FOR OPERATION 3.2.3 Indicated Reactor Coolant System (RCS) total flow rate and Fj, shall be
' maintained as follows for four loop operation.
~
- a. 1) *RCS Total Flowrate 2 390,400 gpm, and
,- m , 2) **RCS Total Flowrate 2 371,400 gpm, and m -- - f ,,._ -
En ,
- b. FI, s +q,r66 [1.0 + E, S (1.0-P)] f;r OfA fue1- -1 (L u F", f 1 AE [1 n + n.3_(LQ4).} for VANTAGE-5-fue'l-
, - . , ~ , - where:
( 1:G1T h' > h-w are obtained by using the movable incore Measured detectors. Anvaluesapproprof FI,iate uncertainty of 4% (nominal) or greater shall then be applied to the measured value of F", before it is compared to the requirements.,and-P- T"EP"al P0"EP
--RATE 0 Tii R".AL POWER APPLICABILITY: MODE 1.
ACTION: With RCS total flow rate or Fl, outside the region of acceptable operation: Within 2 hours either: a. 4
- 1. Restore RCS total flow rate and Fl, to within the above limits, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
i i
' Applicable to Unit I and Unit 2 until completion of cycle 5. 2
** Applicable to Unit I and Unit 2 starting with cycle 6. -
Unit 1 - Amendment No. 56 , BRAIDWOOD - UNITS 1 & 2 3/4 2-8 Unit 2 - Amendment No. 55
I i INSERT A P= IHERMAL POWER - , RATED THERMAL POWER FL" = the FL limit (s) at RATED THERMAL POWER (RTP) specified in the ! OPERATING LIMITS REPORT, and i
)
PF, = the Power Factor Multiplier (s) for FL specified in the OPERATING LIMITS ! REPORT. i l 1 i l i 1 1 l
l l POWER DISTRIBUTION LIMITS
,: , /- e r- / <- /e<n. j j , %s m 4/e CVenAmt."
~ ~
ta iww Wr1N. 6T '
" " ~ ~ " l !- BASES j.
AXIAL FLUX DIFFERENCE (Continued) i . Although it is intended that the plant will be operated with the AFD j within the target band required by Specification 3.2.1 about the target flux i difference, during rapid plant THERMAL POWER reductions, control rod motion l j will cause the AFD to deviate outside of the target band at reduced THERMAL ; 1
; POWER levels. This deviation will not affect the xenon redistribution :
j sufficiently to change the envelope of peaking factors which may be reached on ! a subsequent return to RATED THERMAL POWER (with the AFD within the target
,- band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours is
- provided for operation outside of the target band but within the limits 4 < l 3
f f;;erM2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL j POWER, deviations of the AFD outside of the target band are less significant. l The penalty of 2 hours actual time reflects this reduced significance.
- Provisions for monitoring the AFD on an automatic basis are derived from i the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE ey. core detector outputs ar.d j provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer
- outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours, respectively.
Figure B 3/4 2-1 shows a typical monthly target band. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOWRATE AND l NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flowrate, and nuclear
- enthalpy rise hot channel factor ensure that: (1) the design limits on peak
- local power density and minimum DNBR are not exceeded, and (2) in the event of
! a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided: i i ,
- a. Control rods in a single group move together with no individual rod
, position differing by more than + 12 steps, indicated, from the group demand position,
- b. Control rod groups are sequenced with overlapping groups as described
- in Specification 3.1.3.6, BRAIDWOOD - UNITS 1 & 2 B 3/4 2-2 mvp ms t U .
l ! l 4
.-. -.- . --- . .~ -. - - - - - - . - .... - - - -- - - - . _ - .
i ADMINISTRATIVE CONTROLS I REPORTING REQUIREMENTS (Continued)
! ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- i j 6.9.1.6 The Annual Radiological Environmental Operating Report coverino the operation of the facility during the previous calendar year shall be su5mitted prior to May 1 of each year. The report shall include summaries interpreta-i
' tions, and analysis of trends of the results of the Radiological, Environmental 4 Monitoring Program for the reporting period The material provided shall be consistentwiththeobjectivesoutlinedinil)FRPart50. the ODCM and (2) Sections i IV.8.2 IV.B.3, and IV.C of Appendix I to 10 C j SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT" 1 ; 6.9.1.7 A Radioactive Effluent Release Report covering the operation of the - >
1 facility during the previous year shall be submitted prior to May 1 of each - -
- year. The report shall include a sunnary of the quantities of radioactive 1
liquid and gaseous effluents and solid waste released from the facility. The f". j material provided shall be J1) consistent with the objectives outlined in the 1
- ODCM and PCP and (2) in condormance with 10 CFR 50.36a and Section IV.B.1 of i i Appendix ! to 10 CFR Part 50. !
d I MONTHLY OPERATING REPORT i j 6.9.1.8 Routine reports of operating statistics and shutdown experience, !
- including documentation of all challenges to the PORVs or RCS safety valves, l i shall be submitted on a monthly basis to the Director, Office of Resource i Washington D i Management, U.S. Nuclear Regulatory Commissioncopy to the
, no later Regional than Administrato!
3 the 15th of each month following the calendar month covered by the report. I v m 0PERATING LIMITS REPORT Luu 6 : L - 5.a.:.a 0 r: ting 1i:it: :h:115: ::t:bl1:h:d ::d d:: mat:d in th: OPERAT4NG-l 4-lMITS ar.a5=:T befer: :::t ::1::e :y:1: :r ::1.
- m i 4:g p:rt 25 releed :
t5:< :Lcr:t1,g 1icit: :h:11 b-
- g L"ci:.
Th:
...oiou.io:::1yt10:1 ::t5:d:
..m4- ~ . a ------ a um a uor :::d t: d:t: 4-- a--m-. n
! WU cDely i M" E '"9Df'p "C'"T=3"5 1:hti".6"scr: 5 i5:iEEEi3:E Di:tributi:: 5515 tie:3EtEES:TdM"recedures--- C$ntr:1 ::d L :d10:r Fellei-Ign-Methods 2- ! t D:: dated-SeY.yomber-4974 dated Je 4082 and 3) NF4R-0015 "":::h::rk Of ""." " : i i -Using-the-PHdENlbP -d MC l/er Cc--"terO MFSR-9081 Cede:" d:ted"" n h: Jel"rklaa0. of "RTh: ""cle:r eserat4ag-- D::tgn-Methods-44mits-sha11--be determined :: that--all : plic:ble limechanicaL 14mits, cere theraab hy It: '- g ! auc!es* li="s ! such-as-shutdown-margin r and-trans4ent and accident anal"ri: li=it:) of the safety-analys4s-are-me t. The OPERAT.ING-LIMLTS-REP 0"T, 61eding-anI.cr mid 10 sha14-be- 0 -
- =:6": ;
i revis4ons-or.-supplemer,ts-theretoCContr:1 ^ pre"id-d " pen
-:k with-copies-te th:issuance,Regionab
. reload-cyc14 te the-SRC 00cumen i
-Administrator -d "-M dent I : ::ter.
E 9 l "A single submittal may be made for a multi-unit station. I "A single submittal may be made for a multi-unit station. The submittal should 'j l combine those sections that are common to all units at the station; however, ; i for units with separate radwaste systems, the submittal shall specify the l releases of radioactive material from each unit. i l BRAIDWOOD - UNITS 1 & 2 6-22 AMENDMENT NO. f9' ! 1
f j INSERT B i OPERATING LIMITS REPORT 6.9.1.9 Operating limits shall be established and documented in the OPERATING j LIMITS REPORT (OLR) before each reload cycle or any remaining part of a reload 1 cycle for the following: i l 1. Moderator Temperature Coefficient for Specification 3.1.1.3, i 2. Shutdown Bank Insertion Limit for Specification 3.1.3.5, ) 3. Control Bank Insertion Limit for Specification 3.1.3.6,
- 4. Axial Flux Difference Limits, Target Band for Specification 3.2.1, l
S. Heat Flux Hot Channel Factor and K(Z) for Specification 3.2.2,
- 6. Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for
, Specification 3.2.3, and j 7. F, Radial Peaking factor for Specification 4.2.2.2.
l! The analytical methods used to determine the operating limits shall be those previously reviewed and approved by the NRC in the latest approved version of the following documents:
- 1. WCAP 9272-P-A, " Westinghouse Reload Safety Evaluations Methodology" i (Westinghouse Proprietary). (Methodology for Specification: Shutdown Bank Insertion Limit, Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor)
) 2. WCAP-8385," Power Distribution Control and Load Following Procedures-Topical j- Report"(Westinghouse Proprietary). (Methodology for Specification: Axial Flux 4 Difference, Constant Control Offset Control) i l 3. WCAP 9220-P-A, " Westinghouse ECCS Evaluation Model-1981 Version" (Westinghouse Proprietary). (Methodology for Specification: Heat Flux Hot i Channel Factor)
- 4. WCAP 9581-P-A, Add. 3, "BART A-1: A Computer Code for Best Estimate Analysis of Reflood Transients - Special Report: Thimble Modeling Westinghouse i ECCS Evaluation Model"(Westinghouse Proprietary). (Methodology for Specification: Heat Flux Hot Channel Factor)
- 5. WCAP 10266-P-A, "The 1981 Version of Westinghouse Evaluation Model using BASH Code"(Westinghouse Proprietary). (Methodology for Specification: Heat l Flux Hot Channel Factor) f i
- , - - >n,- ---- - - - - - , .
j i
- 6. NFSR-0016, " Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods", (Methodology for Specification: Shutdown Bank insertion Limit, Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot
- Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor)
- 7. NFSR-0081, " Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoen!x-P and ANC Computer Codes",
i (Methodology for Specification: Shutdown Bank Insertion Limit, Control Bank 1 Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear j Enthalpy Rise Hot Channel Factor) l L (: 8. WCAP 10079-P-A, "NOTRUMP, A Nodal transient Small Break and General ; Network Code" (Westinghouse Proprietary). (Methodology for Specification: Heat ; Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor) ! j 9. WCAP 10054-P-A, " Westinghouse Small Break ECCS Evaluation Model using NOTRUMP Code,"(Westinghouse Proprietary). (Methodology for Specification: '
- Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise l Hot Channel Factor)
'l 10. Comed letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21,1994, transmitting an attachment that documents applicable ' sections of WCAP-11992/11993 and Comed application of the UET methodology - addressed in " Additional Information Regarding Application for Amendment to l Facility Operating Licenses-Reactivity Controls Systems" l The operating limits shall be determined so that all applicable limits (e.g. fuel thermal- ' mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety j analysis are met. i The OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements j thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector, i l l 3 t i_ _ _ _ - _ _ _ - - _ _ - - _ _ _ _ - _ _ _ __ _ . . ,
ATTACHMENT B BYRON STATION l Proposed Changes to Appendix A Technical Specifications of facility i Operating Licenses NPF-37 and NPF-66 l l i Revised Pages: IV ; V 1-4 3/4 1-14 3/4 1-15 3/4 1-20 3/4 1-21 3/4 1-22 3/4 2-1 . 3/4 2-2 ! 3/4 2-3 ! 3/4 2-4 3/4 2-5 3/4 2-8 l B 3/4 2-2 6-22 I l 1 l
I l j LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3 SECTION PAGE 3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1- BORATION CONTROL Shutdown Ma rgi n - T,yg > 200*F. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-1 Shutdown Margin - T,yg i 200'F........................... 3/4 1-3 Moderator. Temperature Coefficient........................ 3/4 1-4 Minimum Temperature for Criticality. . . . . . . . . . . . . . . . . . . . . . 3/4 1-6
'3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown.....................................
2/4 1-7 Flow Paths - Operating................................... 3/4 1-8 Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - Operating............................... 3/4 1-10 Borated Water Source - Shutdown.......................... 3/4 1-11 ; Borated Water Sources - Operating. . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-12 ' Boron Dilution Protection 3/4.1.3 System......................... 3/4 1-13a Il[, ' MOVABLE CONTROL ASSEMBLIES ! Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00. . . . . . . . . .3/4 . . .1-16 Position Indication Systems - Operating.................. 3/4 1-17 ' Position Indication System - Shutdown.................... 3/4 1-18 Rod Drop Time............................................ 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Red Insertion Limits............................. 3/4 1-21 FIGURE 3.1-1 ROD-BAliK-INS ERTI ON - LIMITS -VERSU S -THERMA L POWER-FOUR-LOOP-OPERATION........................... n .- -
~ 3/4 1-22 ,
s '.('Tib3 faae6 IE Ucr us$ )
/
BYRON - UNITS 1 & 2 IV Amendment No. 44-1
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 1 3/4.2 POWER DISTRIBUTION LIMITS l 3/4.2.1 AXIAL FLUX DIFFERENCE...............:.................... 3/4 2-1 FIGURE 3.2-1 4XI AL-FLUX-DI FFERENCE--1:IMIT6,AS= AtEU@J10E0J ,
-RATED 4HERMAL40WER. . ./.TW.C. 9.4%W. .G. .W.'#C3/4 2-3 -
3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR... 3/4 2-4 FIGURE 3.2-2 -K(Z)-NORMALIZED-F q (2)-AS-A-FUNCTION-0F-EORE-HEIGHT. . . 3/4 2-5 I' 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL , FACT 0R................................................. 3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-10 3/4.2.5 DNB PARAMETERS........................................... 3/4 2-13 TABLE 3.2-1 DNB PARAMETERS........................................ 3/4 2-14 3/4.3 INSTRUMENTATION 1 1 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 l l TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 l TABLE 3.3-2 (THIS TABLE IS NOT USED).............................. 3/4 3-7 d. TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................................ 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM ' INSTRUMENTATION........................................ 3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..................................... 3/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS...................... 3/4 3-23 TABLE 3.3-5 (THIS TABLE IS NOT USED).............................. 3/4 3-30 d. TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM ! l INSTRUMENTATION SURVEILLANCE REQUIREMENTS........... 3/4 3-34 l l 1 BYRON - UNITS 1 & 2 V AMEN 0 MENT NO. 4 DEFINITIONS
- 0FFSITE DOSE CALCULATION MANUAL 1 ;;
1.18 The DFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology l and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid 3 ; l effluent monitoring Alam/ Trip Setpoints, and in the conduct of the Environ-i mental Radiological Monitoring Program. The CDCM shall also contain (1) the Z' E. i* Radioactive Effluent Controls and Radiological Environmental Monitoring Programs : ; i required by Sections 6.8.4e and f, and (2) descriptions of the information ! ; j that should be included in the Annual Radiological Environmental Operating and : =~ t Semiannual Radioactive Effluent Release Reports required by Specifications ; l 6.9.1.6 and 6.9.1.7. . - - I OPERABLE - OPERASILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, j cooling or seal water, lubrication or other auxiliary equipment that are i required for the system, subsystem, train, component, or device to perfom its j function (s) are also capable of performing their related support function (s). OPERATING LIMITS REPORT i 1.19.a The OPERATING LIMITS REPORT is the unit-specific document that provides l t operating limits for the current operating reload cycle. These cycle-specific ! operating limits shall be determined for each reload cycle in accordance with ! Specification 6.9.1.9. Plant operation within these operating limits is addressed in individual specifications. 1 OPERATIONAL MODE - MDDE i 1.20 An OPERATIONAL M DE (i.e., MODE) shall correspond to any one inclusive i combination of core reactivity condition, power level, and average reactor i coolant temperature specified in Table 1.2. PHYSICS TESTS . I 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental ! nuclear characteristics of the core and related instrumentation: (1) described l in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, l or (3) othemise approved by the Commission. l l PRESSURE BOUNDARY LEAKAGE . I j 1.22 PRESSURE B0UNDARY LEAKAGE shall be leakage (except steam generator tube
- leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall. -
BYRON - UNITS 1 & 2 1-4 AMENDMENT NO. 46-
,, ,e-. - - + - - - , , , - ,, - - - - . -
4 i 4 REACTIVITY CONTROL SYSTEMS
- 3/4.1.3 MDVA8LE CONTROL ASSBOLIES 1
- GROUP HEIGHT
!i .
LIMITING CONDITION POR OPERATION i,
! 3.1.3.1 All full-length shutdoun and control rods shall be OPERA 8LE and positioned within
- 22 steps (indicated position) of their group step counter 1 demand position.
1 l APPLICA8!LITY: MDOES 1* and 2*. j ACTION: ] a. With one or more full-length rods. inoperable due to being iemovable j as a result of excessive friction or mechanical interference or known to be untrippable, determine that the S WTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in
- HDT stale 8Y within 6 hours.
j
- b. With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a. above, or misaligned from its k
i group stap counter demand height by more than
- 12 steps (indicated j position), POWER OPERATION may continue provided that within 1 i
hour: = ! 1. j The rod is restored to OPERA 8LE status within the above alignment requirements, or I 2. The rod is declared inoperable and the remainder of the rods in i j the group with the inoperable rod are aligned to within
- 12 steps of the inoperable rod while maintaining the rod i t sequence and insertion limits of T; ,..,. 3.1 1. The THERMAL POWER !
level shall be restricted pursuant toj op4.1.-3.% ) during subsequent operation, or gd(ced, 3 a . 5, G, , , 7
- 3. The rod is declared inoperable and the N' i requirement of Specification 3.1.1.1 is satisfied. POWER j OPERATION mmy then continue provided that:
j i a) The THERMAL POWER level is reduced to less than or equal i ' to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron Flux Trip ! Satpoint is reduced to less than or equal to 85% of RATED < l THERMAL POWER. l { b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 l t is determined at least once per 12 hours; I i l *See Special Test Exceptions Specifications 3.10.2 and 3.10.3. j 8YRON - UNITS 1 & 2 3/4 1-14 AMENDMENT NO. 6 l
~
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable
, incore detectors and Fq (Z) knd F" g are verified to be within their limits within 72 hours; and
, d) A raevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confire that the previously analyzed results of these accidents ,
remain valid for the duration of operation under these conditions;
- c. With more than one full-length rod trippable but inoperable due tn :
causes other than addressed by ACTION a. above, or misaligned from : its group step counter demand height by more than + 12 steps _ (indicated position), POWER OPERATION may continue provided that: 3
- 1. Within 1 hour, the remainder of the rods in the group (s) with }
the inoperable rods are aligned to within + 12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Fig.de 1 1-1. The THERMAL POWER level , shall be restricted pursuanthto Specification 3.1.3.6 during - subsequent operation, and v pui Qc e -. . - . .y -
- 2. The inoperable rods shall be [ rest 5r~ed To' OPERABLE status
~
2 72 hours. , othentise, be in NOT STANDBY within 6 hours. l l l i SURVEILLANCE REQUIREMENTS I 4.1.3.1.1 The position of each full-length rod shall be determined to be ! within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the rod position , deviation monitor is inoperable, then verify the group positions at least once l per 4 hours. ! 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days. t BYRON - UNITS 1 & 2 3/4 1-15 AMENDMENT NO. +3-i
l REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION _ - x i
- i 3.1.3.5 All shutdown rods shall be felly "ithdr.awa. 0 5 O f </.1/h,;t ; /M //Yf APPLICABILITY: MODES la and 2*#.
actyeZk/is Nr M /C0 DMID* A2nh* th;er/ca'bryod Ne $sce,4k fib ' Withamaximumofoneshutdownrodnot-fuNP:WTEhdriwn N ept for veillance testing pursuant to Specification 4.1.3.1.2, within 1 hour either:
- a. ful-ly-wi thdraw-the-rodr-o#
b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
,m~
L { k a /i{t real 2/b'S . It / Ele' S /hki co'ri$a ' N- h: 4is C NE//~W& L.'M G $R or
~
J SURVEILLANCE REQUIREMENTS N [4 de .tvh Ale 4.1.3.5 Each shutdown rod shall be determined fu14y-withdrawm fh,4 ,f, , j i
- a. ,j.
Within 15 minutes prior to withdrawal of any rods in Control ^ Bank A, B, C, or D during an approach to reactor criticality, and
- b. At least once per 12 hours thereafter.
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3. #With K,ff greater than or equal to 1. '
1 i l 4 i BYRON - UNITS 1 & 2 3/4 1-20 gg.png.?;- t/}. 4
i REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as down-in. '
-Figure-3+1; -
.)
APPLICABILITY: MODES 1* and 2*#. (cpu /h*te' *in h't t C&#Made \ ACTION: l-Mir.' t?fE-g,7 ,J With the control banks inserted beyond the above insertion limits, except for ! surveillance testing pursuant to Specification 4.1.3.1.2: ' a. Restore the control banks to within the limits within 2 hours, or i l b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which 1,s. allowed by. g ba g positi.o,n using the -above-figure',--on A ' i incur'r'on /im,Vs speejnl/f): " I-
- c. Be in at least HOT STANDBY within 6 ho Me C&d/7#, ' 'd.C 8Md' ,,-
%w . -
.- i l
i SURVEILLANCE REQUIREMENTS ' 4.1.3.6 The position of each control bank shall be determined to be within l the insertion limits at least once per 12 hours except during' time intervals ! when the Rod Insertion Limit Alarm is inoperable, then verify the individual rod positions at least once per 4 hours. ~ i l 1 l
"See Special Test Exceptions Specifications 3.10.2 and 3.10.3. #With K,ff greater than or equal to 1.
BYRON - UNITS 1 & 2 3/4 1-21
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-FIGU RE--3-1 -
-R00 -B AN K-I N S EllTI ON ~ t-lMI TS-VERS US-THtiRMA L-POWER -
-FOUR -LOOP-OfERATION-BYRON - UNITS 1 & 2 3/4 1-22 /,91T.u s;r t. .
3/4.2 POWER DISTRIBUTION LIMITS
'm //G /* . ,?,
3/4.2.1 AXIAL FLUX DIFFERENCE N( T" ##d '# 3/'"" m
~
h /At C Ylt X ~ M L M / C LIMITING CONDITION FOR OPERATION dMdM '
\
3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the i
-following target band (flux difference units) about the target flux differencef */
- e. a 5% for Cycle 1- core average accumulated burnup of less than on
-equal-to-5000 MWD /MTU,-and-tr 3%,--9%-for-Cycle-1-core-average accumulated burnup of-greater-than
-5000 ";!D/HTur -and- ,
-c. 3%,-12%-foreach subsequent-cycle.-
The indicated AFD may deviate outside the above required target band at greater . than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits f-Figure-322-1 and the cumulative penalty deviation time does not exceed ! our,_during th' 4 previous ^ ' 24 hours.
's reil' u., n, /Ac c7EFA',%
The indicated AFD may deviate outside the obove- require TiFget band at greater ////c than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed 1 hour during the previous 24 hours. (dQg-s APPLICABILITY: MODE I above 15% of RATED THERMAL POWER *. ACTION: a. With the indicated AFD outside of the above-required target band and with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes, either:
- 1. Restore the indicated AFD to within the ebove. required target band limits, or ,
2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER. ; b. With the indicated AFD outside of the above required target band for 8 ' more than I hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits +6 . figere 0.2-1 and with THERMAL POWER less than 90% but equal to or a greater than 50% of RATED THERMAL POWER, reduce: 1. THERMAL 30 minutes,POWER and to less than 50% of RATED THERMAL POWER within
- 2. The Power Range Neutron Flux - High #
Se'tpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. m "See Special Test Exceptions Specification 3.10.2. y
"# # #' ~ '
nt ., . . . . Surveillance testing of the Power Range Neutron Flux channe may e performed' pursuant to Specification 4.3.1.1 provided the indi ted AFD is maintained ' within the Acceptable Operation Limits af "igun 0. 14 A total of 16 hours operation may be accumulated with the AFD outside of the ebeve required target band during testing without penalty devia *on.
- BYRON - UNITS 1 & 2 3/4 2-1 yy.f,-ns r u.
i
~
,QMITINGCONDITIONFOROPERATION ACTION (Crmtinued)
- c. With the indicated AFD outside of the ebeve required target band for i more than I hour of cumulative penalty deviation time during the previous 24 hours and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the ebove required target band.
SURVEILLANCE REQUIREMENTS 4.2.1.1 TheindicatedAFDshallbedeI.orminedtobewithinitslimitsduring POWER OPERATION above 15% of RATED THERMAL POWER by: 1.
- a. Monitoring the indicated AFD for each OPERABLE excore channel:
- 1) At least once per 7 days when the AFD Monitor Alare is OPERABLE, and 1
- 2) At least once per hour for the first 24 hours after restoring
- the AFD Monitor Alarm to OPERA 8LE status.
- b. Monitoring and logging the indicated AFD for each OPERABLE excore 4
channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is ! inoperable. The logged values of the indicated AFD shall be
- e sumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its target band i when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the 4bove required target band l shall be accumulated on a time basis of: ' a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above j 50% of RATED THERMAL POWER, and - l b. One-half minute penalty deviation for each 1 minute of POWER OPERATION i outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.
- 4. 2.1. 3 Pie initial determination.of target flux difference following a refuel-ing outage shall be based on design predictions. Othentise, the target flux difference of each OPERABLE excore channel shall be determined by measurement n
j at least once per 92 Effective Full Power Days. {. 4.2.1.4 The target flux difference shall be updated at least once per 31 Effe n Re Full Power Days by either determining the target flux difference pursuan % Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the predicted value at the and of the cycle life. j
.$ l l
6 BYRON - UNITS 1 & 2 3/4 2-2 AMENDMENT NO.,49' t
FIGURE 3.2-1 AXI AL-F LUX - DIFFERENCE-L-IMITS- AS -A FUNCTION --OF-RATED-THERMAL-POWER. W- s (TNr A6ucEisnor u:ED) ', A s' -
/
IEm
- w. ,
bi ~
~O!i2 Ni :.J
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zi:i1E
'm: c OFW 100 UNACCEPTABLE E( 11,90).- EEE(11,90)EUNACCEPTABLE
' ~~
VPERA, TION \
,.s .y
--s.; ~ ',OP. ERA. TIO. N ,
80 [- '.
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g fhEACCEPTABLE!: OPERATION!'. i '."
.i '.
J' '. ( 31,50) (31,50) 40 20 0 7 50 40 30, 20 10 0 10 20 30 40 50 fEUX DIFFE*E-NCE-MI) W BYRON - UNITS 1 & 2 3/4 2-3 Ans:irc us.:4 r s2.
POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2 (Z shall be ited by the following relationships: Fg (Z) 5 -[ 2-}- [K(Z)] for P > 0.53, and ({g w crP]j, C ' N # q(Z) F 4-64} [K(Z)] for P 5 0.5X. IEY] \ / / g] gfZ-)-$-E2-50]-{K(Z-)3-for P > 0-588, e i g(2-)-1-{h003-EK(Z4-}-for ; % 53 4 f[x //, # /s//m9 a/ . Where: N'T #M p ~_ THERMAL POWER , Stoff O?rP):puZis' RATED THERMAL POWER A /4, 6FE4r,v6 4afr_-
.and K(Z) is the function eb arined-from-Mrgure-3-2 fE82/* .sndw' f5F'a given core height location. ~
l APPLICABILITY: MODE 1. sp<ceN/ar /4r CF6iM7W3 23//,0 TGN ACTION: #' With F (Z) exceeding its limit: 9
- a. Reduce THERMAL POWER at least 1% for each 1% q F (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit; and 0
- b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit. 9
. *U n i t Cyc l e a nd - U n i t Cyc l e ** Unit-1-Cycle- 4 and at ter;-Unit-2 Cycle-3-and-after BYRON - UNITS 1 & 2 3/4 2-4 AMENDMENT NO.,36'
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w x, ImOdonznyswon-unz.. 1 < I \ I SYR0N - UNITS 1 & 2 , 3/4 2-5 AMsNOMENT NOS. 34,.34 .
. POWER DISTRIBUTION LIMITS l 3 /4. 2. 3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR i
I LIMITING CONDITION FOR OPERATION 3.2.3 Indicated Reactor. Coolant System (RCS) total flow rate and Fl, shall be maintained as follows for four loop operation.
- a. 1)* RCS Total Flowrate 2371,400 gpm, f
~^ 2)" RCS Total Flowrate 2390,400 gpm, and
- m. , ;-
~
M b. Fl, sb (1.0 + (1.0-P)] -for-OFA-fuel fl,---s h 65 -[ h 0 + 0. 3 - ( 1. 0-P) ]-for-VANTAGE-5-fuel-
,_. ~ ~ - ~ .
where: Me are obtained by using the movable incore '
- - ~ ured values detectors. An approprof Fj,iate uncertainty of 4% (nominal) or greater shal then be applied to the measured value of Fl, before it is compared to '
the requirementsy- g g f- THERMAL-POWER
-RATED-THERMAL-POWER-APPLICABILITY: MODE 1.
ACTION: With RCS total flow rate or Fl, outside the region of acceptable operation:
- a. Within 2 hours either:
) 1. Restore RCS total flow rate and Fl, to within the above limits, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
i 4 .4
- Applicable to Unit 1. Applicable to Unit 2 after cycle 5. 5 5
,,Not applicable to Unit 1. Applicable to Unit 2 until the completion of cycle 5. _ ,
BYRON - UNITS 1 & 2 3/4 2-8 AMENDMENT NO. ,65' 1 I
INSERT A P= THERMAL POWER , RATED THERMAL POWER FL"= the FL limit (s) at RATED THERMAL POWER (RTP) specified in the OPERATING LIMITS REPORT, and P F, = the Power Factor Multiplier (s) for FL specified in the OPERATING LIMITS REPORT. l
POWER DISTRIBUTION LIMITS , , , _ BASES - m AXIAL FLUX DIFFERENCE (Continued) - W' b'
' "2l2/C f:nd?~
b l Although it is intended that the plant will be operated with he-AFW within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL , POWER levels. This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target (- band) provided the time duration of the deviation is limited. Accordingly, a / 1-hour penalty deviation limit cumulative during the previaus 24 hours is providedforoperationoutsideofthetargetbandbutwithinthelimits-of.g ) i Fi;;r: 2.2-1 while at THERMAL POWER levels between 50% and 90% of RATED l THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours actual time reflects this reduced significance. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER'1evels 4 between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer ] outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour and 2 hours, respectively. Figure B 3/4 2-1 shows a typical monthly target band. h 2 -~ 3/4.2 and 2/4 2 3 NcAT FLUX HOT CHANNEL FACTOR, and RCS FLOWRATE AND
-~ ~
nuutEA e THALPY RISE HOT CHANNEL FACTOR
- The limits on heat flux hot channel factor, RCS flowrate, and nuclear i enthalpy rise hot channel factor ensure that
- (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of i a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance
+ criteria limit. ' Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided:
- a. Control rods in a single group move together with no individual red position differing by more than + 12 steps, indicated, from the
- gronp demand positinn,
! b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6, BYRON - UNITS 1 & 2 8 3/4 2-2 hl~hi!M7' ', . s
ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued) ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1.6 The Annual Radiological Environmental Operating Report covering the 3 operation of the facility during the previous calendar year shall be submitted -l'
~
prior to May 1 of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be ' consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT **
;), j 6.9.1.7 A Radioactive Effluent Release Report covering the operation of the i .':
facility during the previous year shall be submitted prior to May 1 of each year. ! The report shall include a summary of the quantities of radioactive liquid and The material - gaseous effluents and solid waste released from the facility.provided shall be (1) consiste and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.
)
MONTHLY OPERATING REPORT 1 l 6.9.1.8 Routine reports of operating statistics and shutdown experience, ' including documentation of all challenges to the PORVs or RCS safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report. OPERATING LIMITS REPORT m QA .g @m C.S.I.0 0 prating-44mits-shall be-established-and-documented-4n-the-OPERATING, I' l ., tIMITS-REPORT-before-each-reload-cycle-or-any-remaining-part-of-a-reload cycle. H C ,,/ -The-analytical-methods-used-to-determine-the operating-1imits-shall-be-those- !V previously reviewed-and-approved-by-the- NRC-in-Topical-Reports:-1-)-WCAP-9272-P-A " Westinghouse-Reload-Safety-Evaluations -Methodology"-dated-July-1985r, e)-WCAP-8385 " Power-Distribution-Control-and-Load Following-Procedures"-dated-September-1974,-3)-NFSR-0016 " Benchmark-of-PWR Nuclear-Design-Methods"-dated duly-1983,-and/or-4)-NFSR-0081 " Benchmark-of-PWR Nuclear-Design Methods-Using-the-PH0ENIX-P-and-ANC-Computer Codes"-dated July-1990.-The-operating 41mits-shalb be-determ i ned - so -t h a t-a l l-appl i c abl e -l i mi t s -(e . g rr-fuel-t he rmal-mec h a n i c al-timits,-core-thermal-hydraulic 4imits,-ECCS-limitsrnuclear--l-imits such-as
-shutdown margin,-and-transient-and-accident-analysis 4imits)-of-the safety-analysis-are-metr-The OPERATING 41MITS-REPORTr-including-any-mid-cycle-revisionsr or-supplements-thereto,-shall-be-provided upon-issuancerfor-each-reload-cycle, to-the NRC-Document-Control-Desk- with-copies to the-Regional-Administrator and-Resident-Inspector: *
*A single submittal may be made for a multi-unit station.
,,A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
BYRON - UNITS 1 & 2 6-22 AMENDMENT NO. 69'
' i I
___- _ - _ - _ _ - _ --_- - ------- O
INSERT B OPERATING LIMITS REPORT 6.9.1.9 Operating limits shall be established and documented in the OPERATING-LIMITS REPORT (OLR) before each reload cycle or any remaining part of a reload cycle for the following:
- 1. Moderator Temperature Coefficient for Specification 3.1.1.3,
- 2. Shutdown Bank Insertion Limit for Specification 3.1.3.5,
- 3. Control Bank Insertion Limit for Specification 3.1.3.6,
- 4. Axial Flux Difference Limits, Target Band for Specification 3.2.1,
- 5. Heat Flux Hot Channel Factor and K(Z) for Specification 3.2.2,
- 6. Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification 3.2.3, and
- 7. F Radial y Peaking factor for Specification 4.2.2.2.
The analytical methods used to determine the operating limits shall be those previously reviewed and approved by the NRC in the latest approved version of the following documents:
- 1. WCAP 9272-P A," Westinghouse Reload Safety Evaluations Methodology" (Westinghouse Proprietary). (Methodology for Specification: Shutdown Bank Insertion Limit, Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor)
- 2. WCAP-8385,' Power Distribution Control and Load Following Procedures-Topical Report" (Westinghouse Proprietary). (Methodology for Specification: Axial Flux Difference, Constant Control Offset Control) j
- 3. WCAP 9220 P-A, " Westinghouse ECCS Evaluation Model-1981 Version" (Westinghouse Proprietary). (Methodology for Specification: Heat Flux Hot Channel Factor)
- 4. WCAP 9561-P-A, Add. 3, "BART A-1: A Computer Code for Best Estimate Analysis of Reflood Transients - Special Report: Thimble Modeling Westinghouse ECCS Evaluation Model" (Westinghouse Proprietary). (Methodology for Specification: Heat Flux Hot Channel Factor)
- 5. WCAP 10266-P-A, "The 1981 Version of Westinghouse Evaluation Model using BASH Code" (Westinghouse Proprietary). (Methodology for Specification: Heat Flux Hot Channel Factor)
- 6. NFSR-0016, " Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods", (Methodology for Specification: Shutdown Bank Insertion Limit, Control Bank Inser11on Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor)
- 7. NFSR-0081, ' Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes",
(Methodology for Specification: Shutdown Bank Insertion Limit, Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor)
- 8. WCAP 10079-P-A, 'NOTRUMP, A Nodal transient Small Break and General Network Code" (Westinghouse Proprietary). (Methodology for Specification: Heat Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor)
- 9. WCAP 10054-P A, " Westinghouse Small Break ECCS Evaluation Model using NOTRUMP Code," (Westinghouse Proprietary). (Methodology for Specification: -
Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor)
- 10. Comed letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21,1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and Comed application of the UET methodology addressed in " Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Controls Systems' l The operating limits shall be determined so that all applicable limits (e.g. fuel thermal- i mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as l SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety i analysis are met. l The OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements !
thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector. I
)
i i i l l E_--_ _ _ . _ .. . _ . , _
ATTACHMENT C Evaluation of Significant Hazards Considerations for Proposed Changes to Appendix A Technical Specifications of Facility Operating Licenses NPF-37, NPF-66, NPF-72, and NPF-77 Commonwealth Edison Company (Comed) has evaluated the proposed amendment and determined that it involves no significant hazards considerations. According to ' 10CFR50.92(c), a proposed amendment to an operating license involves no significant
. hazards considerations if operation of the facility in accordance with the proposed amendment would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. involve a significant reduction in a margin of safety.
A. INTRODUCTION Generic Letter (GL) 88-16, " Removal of Cycle-Specific Parameter Limits from Technical Specification", dated October 4,1988 was issued to encourage licensees to amend Technical Specifications (TS) related to cycle-specific parameters. Several Technical Specifications contain limits associated with reactor physics parameters that generally change with every core reload, possibly requiring changes to the Technical Specifications each fuel cycle. The generic letter provided guidance for relocating certain cycle specific core operating limits from the Technical Specifications to a licensee controlled document, provided NRC approved methodologies are used to determine these cycle-specific core operating limits. Consequently, changes to these cycle dependent core operating limits would be allowed without prior NRC approval and results in a monetary and personnel resource savings for the licensees and the NRC. Comed proposes relocating these cycle specific core operating limits from the Technical Specifications to an Operating Limits Report (OLR). Currently, the OLR contains a cycle-specific limit for the radial peaking factor, F y and Moderator Temperature Coefficient (MTC). C-1
This amendment request proposes expanding the OLR to include the additional cycle-specific core operating limits from the GL. If approved, the following Technical Specification limits will also be located in the OLR:
- 1. Shutdown Bank Insertion Limit for TS 3.1.3.5,
- 2. Control Bank Insertion Limit for TS 3.1.3.6,
- 3. Axial Flux Difference Limits, Target Band for TS 3.2.1,
- 4. Heat Flux Hot Channel Factor and K(Z) for TS 3.2.2, and
- 5. Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification 3.2.3.
The cycle-specific parameters proposed for relocation to the OLR and the proposed TS markups are consistent with the guidance provided in Westinghouse Owners Group letter WOG-90-016, " Core Operating Limits Report License Amendment Submittal", dated January 19,1990. The proposed changes are also consistent with NUREG-1431, " Standard Technical Specifications for Westinghouse Plants."
- 1. The proposed Technical Specification changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The relocation of the cycle-specific core operating limits from the Technical Specifications has no influence or impact on the probability or consequences of j any accident previously evaluated. The Technical Specifications will continue to require operation within the analyzed core operating limits and the appropriate actions will be taken if the limits are exceeded. The cycle specific limits within the OLR will be implemented and controlled by plant procedures. Any needed revisions of the limit values in the OLR will be performed based on NRC approved methodology as delineated in TS 6.9.1.9. Each accident analysis addressed in the Byron and Braidwood Updated Final Safety Analysis Report (UFSAR) will be examined with respect to changes in cycle dependent parameters. These parameters are obtained from the application of NRC approved reload design methodologies, to ensure that the transient evaluation of new reloads are bounded by previously accepted analysis. This examination, which will be performed under the requirements of 10 CFR 50.59 process, ensures that future reloads will not involve a significant increase in the probability or consequences of an accident previously evaluated. 4 Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. C-2
f 4
- 2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The relocation of the cycle specific variables has no influence or impact, nor does it contribute in any way to the probability or consequences of any new or different kind of accident. No safety related equipment, safety function or plant operations will be altered as a result of this proposed change. The cycle specific variables are calculated using NRC approved methods and submitted to the NRC for their review to allow the Staff to continue to trend the values of these limits. The Technical Specifications will continue to require operation within the analyzed core operating limits and appropriate actions will be taken, when, or if, the limits are exceeded. Therefore, the proposed changes do not in any way create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. The proposed change does not involve a significant reduction in a margin of safety for the following reasons:
The margin of safety is not affected by the relocation of cycle specific core operating limits from the Technical Specifications. The margin of safety presently provided by current Technical Specifications remains unchanged. Appropriate measures exist to control the values of these cycle specific limits. The proposed amendment continues to require operation within the core limits as obtained from the NRC approved reload design and safety analysis methodologies. Appropriate actions are required to be taken, when, or if, thest. ... nits are exceeded. The development of the limits for future reloads will continue to conform to those methods described in the NRC approved documentation. In addition, each future ; reload willinvolve a 10 CFR 50.59 safety review to assure that operation of the Byron and Braidwood units within the cycle specific limits will not involve a 4 reduction in the margin of safety as defined in the basis for any Technical Specification. i Therefore, the proposed changes do not Impact operation of the plant in a manner that involves a significant reduction in the margin of safety. l Therefore, based on the above evaluation, Commonwealth Edison has concluded that l these changes do not involve significant hazards considerations. ; l I C-3
ATTACHMENT D Environmental Assessment for Proposed Changes to Appendix A Technical Specifications of Facility Operating Licenses NPF-37, NPF-66, NPF-72, and NPF-77 Commonwealth Edison has evaluated the proposed changes associated with expanding the Operating Limits Report against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has been determined that the proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types of or significant increases in the amounts of any effluent that may be released off-site nor do they affect any of the permitted release paths, or (ill) a significant increase in individual or cumulative occupational radiation exposure (10 CFR 51.22(C)(9)). The proposed changes do not involve changes in record keeping and reporting requirements (10 CFR 51.22(c)(10)). Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth 10 CFR 51.22(c)(9) and (10). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment or environmentalimpact statement of the proposed changes is not required for the changes proposed by the Technical Specification change request.
ATTACHMENT E Byron Station Unit 1 Cycle 7 Operating Limits Report i l l i
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 1, CYCLE 7 1.0 OPERATING LIMITS REPORT This Operating Limits Report (OLR) for Byron Station Unit 1 Cycle 7 has been prepared in accordance with the requirements of Technical Specification 6.9.1.9. The Technical Specifications affected by this report are listed below: 3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.3.5 Shutdown Rod insertion Limit 3/4.1.3.6 Control Rod insertion Limits 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor
# a E-1
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 1, CYCLE 7 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using the NRC approved methodologies specified in Technical Specification 6.9.1.9. 2.1 Moderator Temoerature Coefficient (Specification 3/4.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:
- a. The BOL/ARO/HZP-MTC shall be less positive than 0 Ak/k/*F.
- b. The EOL/ARO/RTP-MTC shall be less negative than -4.1 x 10" Ak/k/*F.
2.1.2 The EOL/ARO/RTP-MTC Surveillance limit is: l The 300 ppm /ARO/RTP-MTC should be less negative than or equal to d
-3.2 x 10 Ak/k/ F.
where: BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero Thermal Power EOL stands for End of Cycle Life RTP stands for RATED THERMAL POWER , l 2.2 Shutdown Rod insertion Limit (Specification 3/4.1.3.5) 2.2.1 All shutdown banks shall be withdrawn to at least 228 steps. I E-2
l OPERATING LIMITS REPORT (OLR) for BYRON UNIT 1, CYCLE 7 2.3 Control Rod Insertion Limits (Specification 3/4.1.3.6) 2.3.1 The control banks shall be limited in physical insertion as shown in Figure 1. 2.4 Axial Flux Difference (Specification 3/4.2.1) 2.4.1 The AXIAL FLUX DIFFERENCE (AFD) target band is +3, -12% of the target flux difference. 2.4.2 The AFD Acceptable Operation Limits are provided in Figure 2. j l
- e E-3
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 1, CYCLE 7 2.5 Heat Flux Hot Channel Factor (Specification 3/4.2.2) 2.5.1 Fo(Z) s [F5"/P](K(Z)] for P > 0.5 Fo(Z) s [F5"/0.5)(K(Z)] for P s 0.5 where: P = the ratio of THERMAL POWER to RATED THERMAL POWER F5" = 2.50 K(Z) is provided in Figure 3. 2.5.2 Fh = F"" [1 + 0.2(1 - P)] F7" limits within specified core planes shall be:
- a. For the lower core region from greater than or equal to 0% to less than or equal to 50%:
- 1) F7" s 1.950 for all core planes containing bank "D" control rods l
- 2) For all unrodded core planes:
F7" s 1.732 0 s Cycle Burnup s 10,000 MWD /MTU F"'" 10,000 < Cycle Bumup < 16,000 MWD /MTU ' Ff" s 1.746 s 1.716 Cycle Bumup 216,000 MWD /MTU ,
- b. For the upper core region from greater than 50% to less than or equal '
to 100%: )
- 1) F7" s 1.890 for all core planes containing bank "D" control rods
- 2) For all unrodded core planes:
F"" s 1.784 0 s Cycle Bumup s 10,000 MWD /MTU F "7 s 1.807 10,000 < Cycle Bumup < 16,000 MWD /MTU Ff" s 1.769 Cycle Bumup 216,000 MWD /MTU j 2.5.3 A plot of [Fo(z)
- Paw) vs Axial Core Height is provided in Figure 4 and . ,
Table 1 contains the data plotted in Figure 4. E-4
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 1, CYCLE 7 2.6 Nuclear Enthalov Rise Hot Channel Factor (Specification 3/4.2.3) F"w s FAI"[1.0 + PFm(1.0 - P)) where: P = the ratio of THERVAL POWER to RATED THERMAL POWER FSI" = 1.65 for VANTAGE 5 Fuel . FSIP = 1.55 for OFA Fuel PFm = 0.3 l 0 E-5 l
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1 I 1 OPERATING LIMITS REPORT (OLR) for BYRON UNIT 1, CYCLE 7 l 1 l FIGURE 2: Axlal Flux Difference Limits As A Function of Rated Thermal Power
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OPERATING LIMITS REPORT (OLR) for BYRON UNIT 1, CYCLE 7 Table 1: (Fo(z)
- Pu] vs. Axial Core Height CORE HEIGHT MAXIMUM (FEET) Fo
- P 4 BOTTOM 1 0.1252 0.41 2 0.3756 0.77 3 0.6259 1.83 4 0.8763 2.13 5 1.1267 2.37 6 1.3771 2.47 7 1.6274 2.50 8 1.8778 2.47 j 9 2.1282 2.27 i 10 2.3786 2.40 11 2.6289 2.44 4 12 2.8793 2.47 13 3.1297 2.48 4 14 3.3801 2.49 15 3.6305 2.50
< . 16 3.8808 2.30 17 4.1312 2.49 18 4.3816 2.49 19 4.6320 2.48 20 4.8823 2.46 21 5.1327 2.44 22 5.3831 2.40 23 5.6335 2.18 24 5.8838 2.35 25 6.1342 2.44 26 6.3846 2.45 4 27 6.6350 2.46 28 6.8853 2.46
- 29 7.1357 2.44 30 7.3861 2.28 31 7.6365 2.42 32 7.8868 2.39 .,
33 8.1372 2.35 34 8.3876 2.32 35 8.6380 2.32 36 8.8883 2.27 37 9.1387 2.19 38 9 3891 2.31 30 9.6305 2.31 40 9.8898 2.34 41 10.1402 2.36 42 10.3006 2.36 43 10.6410 2.23 44 10.8914 2.13 45 11.1417 1.94 46 11.3921 1.66 47 11.6425 0.71 TOP 48 11.8929 0.41 1 E-10
ATTACHMENT F Byron Station Unit 2 Cycle 6 Operating Limits Report l I i l 1 l l l 1
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2, CYCLE 6 1.0 OPERATING LIMITS REPO'RT This Operating Limits Report (OLR) for Byron Station Unit 2 Cycle 6 has been prep in accordance with the requirements of Technical Specification 6.9.1.9. The Technical Specifications affected by this report are listed below: 3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.3.5 Shutdown Rod insertion Limit 3/4.1.3.6 Control Rod Insertion Limits 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor 4 'k F-1
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2, CYCLE 6 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using the NRC approved methodologies specified in Technical Specification 6.9.1.9. 2.1 Moderator Temperature Coefficient (Specdication 3/4.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:
- a. The BOL/ARO/HZP-MTC shall be less positive than 0 Ak/k/ F.
- b. The EOUARO/RTP-MTC shall be less negative than -4.1 x 10 Ak/k/*F.
2.1.2 The EOL/ARO/RTP-MTC Surveillance limit is: The 300 ppm /ARO/RTP-MTC should be less negative than or equal to
-3.2 x 10" Ak/k/ F.
where: BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero Thermal Power l l EOL stands for End of Cycle Life RTP stands for RATED THERMAL POWER ,, 2.2 Shutdown Rod Insertion Limit (Specification 3/4.1.3.5) 2.2.1 All shutdown banks shall be withdrawn to at least 228 steps. k F-2 l
~ ' - . - .
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2, CYCLE 6 2.3 Control Rod in$ertion Limits (Specification 3/4.1.3.6) 2.3.1 The control banks shall be limited in physical insertion as shown in Figure 1. 2.4 Axial Flux Difference (Specification 3/4.2.1) 2.4.1 The AX!AL FLUX DIFFERENCE (AFD) target band is +3, -12% of the target flux difference. 2.4.2 The AFD Acceptable Operation Limits are provided in Figure 2. ) i 4 E e 'l f
'e F-3
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2, CYCLE 6 1 2.5 Heat Flux Hot Channel Factor (Specification 3/4.2.2)
! 2.5.1 Fo(Z) s [F5"/P][K(Z)] for P > 0.5
)
Fo(Z) s [F5"/0.5][K(Z)] for P s 0.5 ~ f where: P = the ratio of THERMAL POWER to RATED THERMAL POWER l
)
FE" = 2.50 .
< K(Z)is provided in Figure 3.
i 2.5.2 Fh = F7"[1 + 0.2(1 - P)] ! 1 F7" limits within specified core planes shall be: l
- a. For the lower core region from greater than or equal to 0% to less than or equal to 50%:
- 1) F7" s 2.052 for all core planes containing bank "D" control rods j
! 2) For all unrodded core planes: i 0 s Cycle Burnup s 12,000 MWD /MTU F7" s 1.765 Cycle Burnup > 12,000 MWD /MTU ,, F7" s 1.774 t t ' b. For the upper core region from greater than 50% to less than or equal ' ! to 100%:
- 1) F7" s 1.994 for all core planes containing bank "D" control rods
- 2) For all unrodded core planes:
i 0 s Cycle Burnup s 12,000 MWD /MTU F{" F, s s1.786 1.750 Cycle Burnup > 12,000 MWD /MTU s 2.5.3 A plot of [Fo(z)
- Pu] vs Axial Core Height is provided in Figures 4 and 5 and Tables 1 and 2 contain the data plotted in Figures 4 and 5. .
I a F-4
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2, CYCLE 6 1 2.6 Nuclear Enthalov Rise Hot Channel Factor (Specification 3/4.2.3) FL s FEI"[1.0 + PFm(1.0 - P)] where: P = the ratio of THERMAL POWER to RATED THERMAL POWER FAI" = 1.65 for VANTAGE 5 Fuel i FAI" = 1.55 for OFA Fuel PFa = 0.3 a i
)
J 1 1 ) i 4 i. J F-5
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1 1 I OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2, CYCLE 6 1 I 1 1 FIGURE 2: Axial Flux Difference Limits As A Function of Rated j
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1 F-7 l
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OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2. CYCLE 6 Figura 4: [Fo(z)
- P ,1 va. Axial Cora Height Cycle Burnup of 0 to 12,000 MWDIMTU (6.0, 2.50) 2.600
_ It l l _ - - _ a" - - -
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, a -
- - - Top Surveillance Limit '
0.600 , - - Bottom Surveillance Limit : - -- 0.400 - ' _ _ _ O.200 ---- '
.O.000 i '
0.0000 2.0000 4.0000 6.0000 8.0000 10.0000 12.0000 BOTTOM Core Height (Feet) TOP F-9
OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2. CYCLE 6 Figure 5: [Fo(z)
- Pol vs. Axial Core Height Cycle Burnup of > 12.000 MWD /MTU (6.0, 2.50) 2.600 I _ _ In_ I
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, - - - Top Surveillance Limit
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.~
. F-10 i
1 1 OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2, CYCLE 6 l 1 . TABLE 1: [Fo(z)
- Pna) vs Axial Core Height i I
Cycle Bumup of 0 to 12,000 MWD /MTU CORE HEIGHT MAXIMUM (FEET) Fo
- P BOTTOM 1 0.1252 0.424 2 0.3756 0.852 3 0.6259 1.908 4 0.8763 2.192 5 1.1267 2.405 6 1.3771 2.487 7 1.6274 2.409 8 1.8778 2.400 9 2.1262 2.310 10 2.3786 2.408 11 2.6290 2.414 12 2.8793 2.444 13 3.12E7 2.406 j 14 3.3801 2.480 15 3 6305 2.495 16 3.8808 2.304 17 4.1312 2.497 18 4.3816 2.499 19 4 6320 2.492 .
20 4.8823 2.478 I 21 5.1327 2.461 22 5.3831 2.435 23 5.6335 2.227 24 5.8839 2.404 25 6.1342 2.419 26 6.3846 2.419 27 8.6350 2.437 28 6.8854 2.445 ,* 29 7.1357 2.435 30 7.3861 2.283 31 7.6365 2.419 - 32 7.8869 2.398 33 8.1372 2.376 34 8.3876 2.344 35 8.6380 2.334 36 8.8864 2.289 37 9.1388 2.217 38 9.3891 2.296 39 9.6395 2.296 40 9.8899 2.327 41 10.1400 2.358 42 10.3910 2.361 43 10.6410 2.240 44 10.8910 2.171 45 11.1420 1.994 46 11.3920 1.735 47 11.6430 0.807 TOP 48 11.8933 0.436 l F-11 l l
l OPERATING LIMITS REPORT (OLR) for BYRON UNIT 2, CYCLE 6 TABLE 2: [FO(z)
- PRw) vs Axial Core Height Cycle Bumup of > 12,000 MWD /MTU CORE HElGHT MAXIMUM (FEET) Fo
- P BOTTOM 1 0.1252 0.425 2 0.3756 0.853 3 0.6250 1.912 4 0.8763 2.196 5 1.1267 2.407 6 1.3771 2.480 7 1.6274 2A00 8 1.8778 2.400 9 2.1282 2.308 10 2.3786 2.405 11 2.6290 2A00 12 2.8793 2.430 13 3.1297 2.460 14 3.3801 2.472 15 3.6305 2.484 16 3.8808 2.295 17 4.1312 2.484 1 18 4.3816 2.487 19 4.6320 2.477 20 4.8823 2.463 21 5.1327 2.444 22 5.3831 2.417 23 5.6335 2.205 24 5.8830 2.380 25 6.1342 2A07 26 6.3846 2A06 27 6.6350 2A22 28 6.8854 2.430 '
29 7.1357 2.423 30 7.3861 2.271 31 7.6385 2.422 *- 32 7.8809 2A02 33 8.1372 2.377 34 8.3676 2.345 ) . . 35 8.6380 2.337 36 8.8884 2.293 37 9.1388 2.218 j 38 9.3801 2.297 l 30 9.6395 2.295 j 40 9.8800 2.326 " 41 10.1400 2.358 ] 42 10.3910 2.380 " 43 10.6410 2.240 44 10.8010 2.108 1 45 11.1420 2.007 46 11.3920 1.743 47 11.6430 0.784 , TOP 48 11.8930 OA23 -
)
I F-12 I
TTACHMENT G Braidwood Station Unit 1 Cycle SA -l Operating Limits Report l l l i l a. 4 I 1 l
)
OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 1, CYCLE 5A 1.0 OPERATING LIMITS REPORT This Operating Limits Report (OLR) for Braidwood Station Unit 1 Cycle SA has been prepared in accordance with the requirements of Technical Specification 6.9.1.9. The Technical Specifications affected by this report are listed below-3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.3.5 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod Insertion Limits 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor 't 1 l G-1
i l l OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 1, CYCLE 5A 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using the NRC approved methodologies specified in Technical Specification 6.9.1.9. 2.1 Moderator Temperature Coefficient (Specification 3/4.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:
- a. The BOUARO/HZP-MTC shall be less positive than 0 Ak/k/ F.
- b. The EOUARO/RTP-MTC shall be less negative than -4.1 x 10" Ak/k/*F.
2.1.2 The EOUARO/RTP-MTC Surveillance limit is: The 300 ppm /ARO/RTP-MTC should be less negative than or equal to
-3.2 x 10" Ak/k/ F.
where: BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero Thermal Power EOL stands for End of Cycle Life RTP stands for RATED THERMAL POWER , 2.2 Shutdown Rod insertion Limit (Specification 3/4.1.3.5) 2.2.1 All shutdown banks shall be withdrawn to at least 231 steps. G-2
1 OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 1, CYCLE 5A 4 . Control Rod insertion Limits (Specification 3/4.1.3.6) 2.3 i t, 2.3.1 The control banks shall be limited in physical insertion as shown in Figure 1. 2.4 Axial Flux Difference (Specification 3/4.2.1) - 2.4.1 The AXIAL FLUX DIFFERENCE (AFD) target band is +3, -12% of the
- target flux difference.
2.4.2 The AFD Acceptable Operation Limits are provided in Figure 2. 4 2 i I 4 I l l i I
\ ..
G-3
' OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 1, CYCLE SA i 2.5 Heat Flux Hot Channel Factor (Specification 3/4.2.2) 2.5.1 Fo(Z) s [F5"/P][K(Z)] for P > 0.5 Fo(Z) s [F5"/0.5][K(Z)] for P s 0.5 where: P = the ratio of THERMAL POWER to RATED THERMAL POWER F5" = 2.50 4 K(Z)is provided in Figure 3. 2.5.2 FL = F""[1 + 0.2(1 - P)] FS" limits within specified core planes shall be:
- a. For the lower core region from greater than or equal to 0% to less than or equal to 50%:
- 1) F7" s 2.700 for all core planes containing bank "D" control rods
- 2) F7" s 1.755 for all unrodded core planes
- b. For the upper core region from greater than 50% to less than or equal to 100%: .,
- 1) F7" s 2.052 for all core planes containing bank "D" control rods
- 2) FS" s 1.772 for all unrodded core planes 2.5.3 A plot of (Fa(z)
- Paw] vs Axial Core Height is provided in Figure 4 and Table 1 contains the data plotted in Figure 4.
G-4
OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 1, CYCLE 5A 2.6 Nuclear Enthalov Rise Hot Channel Factor (Specification 3/4.2.3) FL s; F7(1.0 + PFm(1.0 - P)] where: P = the ratio of THERMAL POWER to RATED THERMAL POWER F7 = 1.65 for VANTAGE 5 Fuel - F7 = 1.55 for OFA Fuel PF, = 0.3 G-5
OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 1, CYCLE SA Figure 1: i Control Bank Insertion Limits Versus Percent Rated Thermal Power (29 %, 228)
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OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 1, CYCLE SA
. Table 1: [Fo(z)
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U A T'ACHMENT H Braidwood Station Unit 2 Cycle 5 Operating Limits Report e l I 9 t _ _ ___ J
l OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 2, CYCLE 5 l 1.0 OPERATING LIMITS REPORT d
- This Operating Limits Report (OLR) for Braidwood Station Unit 2 Cycle 5 has been prepared in accordance with the requirements of Technical Specification 6.9.1.9.
The Technical Specifications affected by this report are listed below- l 3/4.1.1.3 Moderator Temperature Coefficient ! 3/4.1.3.5 Shutdown Rod insertion Limit ! j 3/4.1.3.6 Control Rod Insertion Limits ; 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor i H-1
OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 2, CYCLE 5 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using the NRC approved methodologies specified in Technical Specification 6.9.1.9. 2.1 Moderator Temoerature Coefficient (Specification 3/4.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:
- a. The BOUARO/HZP-MTC shall be less positive than 0 Ak/k/*F.
- b. The EOUARO/RTP-MTC shall be less negative than -4.1 x 10" Ak/k/*F.
2.1.2 The EOUARO/RTP-MTC Surveillance limit is: The 300 ppm /ARO/RTP-MTC should be less negative than or equal to
-3.2 x 10" Ak/k/ F.
where: BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero Thermal Power EOL stands for End of Cycle Life RTP stands for RATED THERMAL POWER 2.2 Shutdown Rod Insertion Limit (Specification 3/4.1.3.5) 2.2.1 All shutdown banks shall be withdrawn to at least 228 steps. I H-2 i
OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 2, CYCLE 5 2.3 Control Rod insertion Limits (Specification 3/4.1.3.6) 2.3.1 The control banks shall be limited in physical insertion as shown in Figure 1. 2.4 Axial Flux Difference (Specification 3/4.2.1) 2.4.1 The AXIAL FLUX DIFFERENCE (AFD) target band is +3, -12% of the target flux difference. 2.4.2 The AFD Acceptable Operation Limits are provided in Figure 2. l H-3
OPERATING LIMITS REPORT (OLR) for.BRAIDWOOD UNIT 2, CYCLE 5 2.5 Heat Flux Hot Channel Factor (Specification 3/4.2.2) l 2.5.1 Fo(Z) s [F5"/P][K(Z)] for P > 0.5 Fo(Z) s [F5"/0.5][K(Z)] for P s 0.5 l where: P = the ratio of THERMAL POWER to RATED THERMAL POWER F5" = 2.50 K(Z)is provided in Figure 3.
- 2.5.2 Fh = F7'[1 + 0.2(1 - P)]
l F7" limits within specified core planes shall be: y
- a. For the lower core region from greater than or equal to O'A to less than !
l or equal to 50%: 3
- 1) F7" s 2.052 for all core planes containing bank "D" control rods I 2) FE" s 1.735 for all unrodded core planes
- b. For the upper core region frorn greater than 50% to less than or equal to 100%: ..
- 1) F7" s 2.052 for all core planes containing bank "D" control rods
- 2) F7" s 1.817 for all unrodded core planes 2.5.3 A plot of [Fo(z)* Pu] vs Axial Core Height is provided in Figure 4 and Table 1 contains the data plotted in Figure 4.
4 H-4
OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 2, CYCLE 5 2.6 Nuclear Enthalov Rise Hot Channel Factor (Specification 3/4.2.3) FE s; FAI'[1.0 + PFm(1.0 - P)] where: P = the ratio of THERMAL POWER to RATED THERMAL POWER FE" = 1.65 for VANTAGE 5 Fuel FE" = 1.55 for OFA Fuel PF m = 0.3 l 1 l 4 4 I i 6 H-5
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OPERATING LIMITS REPORT (OLR) for BRAIDWOOD UNIT 2, CYCLE 5
. Table 1: [Fo(z)
- Pu] vs. Axial Core Height CORE HEIGHT WOGMUM (FEET) Fa
- P 0.1262 0.424 soTToM 1 0.3758 OA44 2
3 OA200 1.882 4 0.8783 2.170 5 1.1287 2.385 6 1.3771 2.488 7 1A274 2A00 8 1J778 2.489 9 2.1232 2.205 10 2.3788 2.354 11 2.8289 2.385 12 2A793 2.378 13 3.1297 2.389 14 3.3801 2.414 15 3.6305 2.427 16 3.8808 2.24 17 4.1312 2.430 18 4.3816 2.433 19 4 8320 2.423 20 4.8823 2 400 5.1327 2.391 21 22 5.3831 2.386 23 5.8335 2.156 24 5.8838 2.323 25 6.1342 2.442 26 6.3648 2.466 27 6.6380 2.464 28 6.8883 2.467 I 29 7.1367 2.447 30 7.3881 2.289 31 7.8385 2.433 32 7.8868 2.408 33 8_1372 2.382 .. ; l j 34 8.3876 2.300 l 35 8.8380 2.371 36 8.8864 2.315 37 9.1387 2.246 38 9.3891 2.347 30 9.6305 2.368 40 9.8800 2.377 41 10.1400 2.362 42 10.3910 2.300 4 10.6410 2.241 44 10.8910 2.162 4 11.1420 2.010 7 4 11.3820 1.747 47 ' 11.6420 0.805 , . 4 11A830 0.441 TOP H-10 __ ______________-._____________________________-._______a}}