ML20069D510
| ML20069D510 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/17/1982 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20069D508 | List: |
| References | |
| NUDOCS 8209210251 | |
| Download: ML20069D510 (31) | |
Text
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TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i il
? TRIP SETPOINT ALLOWABLE VALUES 7 FUNCTIONAL UNIT 5 10% Turbine Impulse 5^11% Turbine Impulse
$ 21. Turbine Impulse Chamber Pressure - Pressure Equivalent a (P-13) Input to Low Power Reactor Trips Pressure Equivalent
' - Block P-7
< 35% of RATED < 36% of RATED l
- 22. Power Range Neutron Flux - (P-8) Low THERMAL POWER Reactor Coolant Loop Flow, and Reactor THERMAL POWER Trip Power Range Neutron Flux - (P-10) - 210% of RATED 2 9% of RATED
- 23. THERMAL POWER Enable Block of Source, Intermediate, THERMAL POWER and Power Range (low setpoint) Reactor l
Trips Not Applicable Not Applicable
- 24. Reactor Trip P-4
" < 50% of RATED < 51% of RATED
- 25. Power Range Neutron Flux - (P-9) - THERMAL POWER Blocks Reactor Trip for Turbine THERMAL POWER Trip Below 50% Rated Power NOTATION
+ 5 I NOTE 1: Overtemperature AT (
I
) < AT,{Kj -K 2 (1 + T2 5 )[T( )-T'] + K3 (P-P') - f j(AI)}
l+tSj 1+TS4 3 where: jf = Lag compensator on measured AT
= Time constants utilized in the lag compensator for AT T3l = 2 secs.
T j AT, = Indicated AT at RATED THERMAL POWER Ky $ 1.15
. K = 0.011 -
2
TABLE 2.2-1 (Continued) - i
; 0. vi i E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i
8 NOTATION (Continued)
]> ; fi NOTE 1: (Continued)
. z .
' Z 1+152
= The function generated by the lead-lag controller for lavg dynamic compensation
" l+T53 ~
33 secs., r,&r p
= Time constants utilized in the lead-lag controller for T,yg,12 3
1 3 = 4 secs.- T = ' Average temperature "F . I
= Lag compensator on measured T,yg 1+I54 .
I = Time constant utilized in the measured T,yg lag compensator, 14= 2 secs.
] 4 at RATED THERMAL POWER)
T' < 578.2 F (Nominal T avg
,K3 = 0.00055 .
i P = Pressurizer pressure, psig P' = 2235 psig (Nominal RCS operating pressure) . Laplace transform operator (sec"I) ~/
*S =
i
' and f (al) is a function of the indicated difference between top and bottom detectors ,
l ofthdpower-rangenuclearionchambers;withgainstobeselectedbasedonmeasured ' instrument response during plant startup tests such that: and q g between - 29 percent and + 5 percent f (i) arepkrcenkRATEDTHERMALPOWERinthetopandbotlo(AI)=0(whereqborereIpectivel for q m halves of the is t tal THERMAL POWER in pert.ent of RATED THERMAL POWER).
.' and qt*9b
!. s. =
- i
t m TABLE 2.2-1 (Continued) i' E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS h E
' NOTATION (Continued)
E 4
- NOTE 1: (Continued)
-q exceeds -29 percent, the AT trip set-(ii) for each percent that the magnitude of (apointshallbeautomaticallyreducedbyIf50pS)centofi r
q exceeds +5 percent, the AT trip set-(iii) forpoint eachshallpercent that the magnitude be automatically reduced by Oof (qb86 pN) r cent of its value at RATED THERMAL POWER. I I NOTE 2: Overpower AT ( I
) 5 AT,(K4 -K S S
)( ) T -K6[T( ) - T9 -,2 f f0I)I 1+tSj 5 (1 + 15 b I*Ib4 I*Ib4 Where: j ,
= as defined in Note 1 3
I j
= as defined in Note 1 j AT,
= as defined in Note 1 lc t.
K 4 5 1.087 g
= 0.02/*F for increasing average temperature and 0 for decreasing average K
5 temperature y
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3 73 5 compensation , g , i i' i - e
TABLE 2.2-1 (Continued) 5 f y' REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8 5I NOTATION (Continued) l s. [ NOTE 2: (Continued)
' 1 5
= Time constant utilized in the rate-lag controller for Tavg' T5 = 10 secs.
j g
= as defined in Note 1 1
4
= as defined in Note 1 K =
6 0.0011 for T > T" and K6 = 0 for T i T" T = as defined in Note 1
" =
3 T" Indicated T,yg at RATED THERHAL POWER (Calibration temperature for AT instrumentation, 5 578.2 F) S = as defined in Note 1 f p(AI) = 0 for all AI f
,' NOTE 3: The channel's maximum trip setpoint shall not exceed its computed trip point by more than i
,- 2 percent.
! I ~
l' 3 i-
2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE
- The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fissio'n products to the reactor coolant. Overheating of the fuel cladding is prevented -
by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant snarp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been 4 developed to predict the DNB flux and the location of DNB for axially uniform a and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, ! defined as the ratio of the heat flux that would cause DNS at a partict.lar core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR during steady state operation, normal operational traasients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. The curves are based on an enthalpy hot channel factor, Fh, of 1.55 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is
; included for an increase in Fh at reduced power based on the expression:
Fh=1.55[l+0.3(1-P)] where P is the fraction of RATED THERMAL POWER i SEQUOYAH - UNIT 1 B 2-1 n ----n , ,-- , - ~ .,-n --
8 e POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOWRATE AND R LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R j , R 2shall be maintained within the regions of allowable operation shown on Figure 3.2-3 for 4 loop operation: Where: N Fgg
- a. =
R) 1.49 [1.0 + 0.3 (1.0 - P)]
- b. R
- 2 [1 - RBP (Bu)] ,
= THERMAL POWER c* P l RATED THERMAL POWER '
- d. F H
= MeasuredvaluesofFhobtainedbyusingthemovable i incore detectors to cbtain a power distribution map.
ThemeasuredvaluesofFhshallbeusedtocalculate R since Figure 3.2-3 includes measurement uncertainties of 3.5% for flow and 4% for incore measurement of F and
- e. RBP (Bu) = Rod Bow Penalty as a function of region average burnup as shown in Figure 3.2-4, where a region is defined as those assemblies with the same loading date (reloads) or enrichment (first core).
. APPLICABILITY: MODE 1 ACTION:
With the combination of RCS total flow rate and R), R2 utside the re.gions of acceptable operation shown on Figure 3.2-3:
- a. Within 2 hours:
- 1. Either restore the combination of RCS total flow rate and R), R 2 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.
SEQUOYAH - UNIT 1 3/4 2-10
~ "~
+ i-i e f . I * - 19 . o 48 t
! l
$ [. uuinion. unit. int **uUNCERTAINTIES ME ASUREMENT nusinininius.inneinuieesi, OF .
j
, 3.5% FOR FLOW AND 4% FOR INCORE f -
f N MEASUREMENT OF F3H ^ < E 46 IN THIS FIGURE. l ACCEPTABLE. q 4 .;n. t OPERATION t_. g ! . REGION FOR l R2ONLY
' ..] . U N ACCE PTAB'L .
44 + .
"+i '
% ! OPERATION g f! ,
j REGION Q m 42 ACCEPTABLE - j g g OPERATION . j e REGION FOR . 7 % R, & R, j! , (1.029,40.68) O ,
" 40 i , /
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1 i
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~
UNACCEPTABLE OPERATION - REGION NIIIIIIIIIIINIIIIIIIIIINI 36 0.90 0.92 0.94 0.96 0.98 1.00 1.02 1.04 1.06 ,
- R, = F5u/1.49[1.0 + 0.30 0 - P)]
!, R2= R,/[1 - RBP(Bu)]
f FIGURE 32-3 RCS Total Flowrate Versus R, and R2 - Four Loops in Operation
3/4.2 POWER DISTRIBUTION LIIIITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) LI!!ITING CONDITION FOR OPERATION
. 3. 2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the allowed operational space defined by Figure 3.2-1.
APPLICASILITY: MODE 1 ABOVE 50 PERCENT RATED THERMAL POUER ACTION: _
- a. With the indicated AXIAL FLUX DIFFERENCE outside of the Figure 3.2-1 limf *d, 1.) Either restore the indicated AFD to within the Figure 3.2-1 limits within 15 minutes, or 2.) Reduce THER!tAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux - High Trip setpoints to less than or equal to 55 percent of RATED THERitAL POWER within the next 4 hours.
- b. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the Figure 3.2-1 limits.
W SEQUOYAH UNIT 1 3/4 2-1 . 4 e
~*
PC'.lER DISTRIBUTION LI!1ITS SURVE!LLANCE REQUIREMENTS
- 4. 2.1.1 The indicated AXIAL FLUX OIFFERENCE shall be determined to be within its limits during POWER OPERATION above 50 percent of RATED THERMAL POWER by;
- a. Monitoring the indicated AFD for each OPERABLE excore channel:
- 1. At least once per. 7 d5ys when the AFD Moritor Alarm is OPERABLE, and
- 2. At least once per hour for the first 24 hours after re-storing the AFD Monitor Alarm to OPERABLE status.
- b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.
The logged values of the indicated AXIAL FLUX OIFFERENCE shall be assumed to exist during the interval preceding each log-ging.
- 4. 2.1. 2 The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be
- outside the limits.
4 SEQUOYAH UNIT 1 3/4 2-2 o N
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-10 0 10 20 30 40 50
-50 -40 -30 -20 Flux Difference (aI)"
FIGURE 3,2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (TYPICAL EXAMPLE) SEQUOYAH UNIT 1 3/4 2-4 s
a . POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-F0(Z) LIMITING CONDITION FCR OPERATION 3.2.2 Fg (z) shall be limited by the following relationships: Fq (z) 1 L 2.237 ] [K(i)] for P > 0.5 e
. Fg (z) 1 L 2.237 ] [K(Z)] for P 10.5 0.o H MA([0WER where P = RAic-J ancRMAL POWER and K(z) is the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY: MODE 1 ACTION: With Fg (z) exceeding its limit: 1.- Recuce THERMAL POWER at least 1 percent for each I percent F0 (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 (value of K4) have been reduced at least 1 percent (in AT span) for each I percent Fg(z) exceeds the limit.
- b. Identify and correct the cause of the out of limit condi-tion prior to. increasing THERMAL POWER; THERMAL POWER may then be increased provided Fg(z) is demonstrated through incore mapping to be within its limit. .
h
. SEQUOYAH UNIT 1 3/4 2-5 i .
\ . .
e . PCWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS
~
4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 FQ (z) shall be evaluated to cetemine if FQ (z) is within its limit by:
- a. Using the moveable incore detectors to obtain a power distri-
'bution map at any THERMAL POWER greater than 5 percent of RATED THERMAL POWER.
- b. Increasing the measured Fg(z) component.of the power distri-bution map by 3 percent to account for manuf acturing tolerances and further increasing the value by 5 percent to account for measurement uncertainties.
- c. Satisfying the following relationship:
N 2 F g (z) i p.23, x Uz) for P > 0.5 N Fq (z)-1 x* f, I for P 10.5 where F' (z) is the measured Fg(z) increased by the allow-ances for manufacturing tolerances and measurement uncertainty, limit F g is the Fg limit, X(z) .is given in Figure 3.2-2, P is the relative THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.14. N
- d. Measuring Fg (z) according to the following schedule:
- 1. Upon achieving equilibrium conditions after exceeding by ,
10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which Fg(z) was last determined,* or
- 2. At least once per 31 effective full power days, whichever occurs first.
s
*0uring power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.
SEQUOYAH UNIT 1 3/4 2-6 i 4
e . PC'4ER DISTRIBUTIO.'l LIMITS ,,' SURVEILLANCE REQUIREMENTS (Cont)
- e. With measurements indicating ,
maximum !F (z) over z gg{ has increased since the previous determination of FqM (z) either of the following actions shall be taken:
- 1. Fg M(z) shall be increased by 2 percent over that specified in 4.2.2.2.c, or
- 2. F0 M(z) shall be measured at least once per 7 effec-t1ve full power days until 2 successive maps indii: ate that maximum F (z) is not increasing.
over z ggj
- f. With the relationships specified in 4.2.2.2.c above not being satisfied:
-1. Calculate the percent F Q (z) exceeds its limit by the following expression:
. . t
,f -
M f F0 (z) x W(z) mum -1 , x 100 for P .3, 0.5
< g ,
2 237 x g(z)
' M
)
maximum F0 (z) x W(z) -1 ( x 100 for P < 0.5
< over z j 2.237 .
6 \ , 0.5 , d
- 2. Either of the following actions shall be taken:
- a. Place the core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied. Power level may then be increased provided the AFD limits of Figure
; 3.2-1 are reduced 1% AFD for each percent FQ (z) exceeded its limit, or
- b. Ccmply with the requirements of Specification 3.2.2 for FQ(z) exceeding its limit by the percent cal-culated above SEQUOYAH UNIT 1 3/4 2-7
POWER DISTRIBUTION LIMITS SURVE'ILLANCE REQUIREMENTS (continued)
- g. The limits specifiec in 4.2.2.2.c, 4.2.2.2.e, and 4.2.2.2.f above are not applicable in the following core plane regions:
- l. Lower core region 0 to 15 percent inclusive.
~
- 2. Upper core region 85 to 100 percent inclusive.
4.2.2.3 When gF (z) is measured for. reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured Fo(z) shall be obtained from a power distribution map and increased by 3 percent to account for manufacturing tolerances and further increased by 5 percent to account for measurement uncertainty. e 4. v s
. SEQUOYAH UNIT 1 3/4 2-8
= -- _. . - - - )
i I CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION
- 3. 6.1. 5 Primary containment average ain temperature shall be maintained:
- a. between 85*F* and 105*F in the containment upper compartment, and
- b. between 100*F* and 125'F in the containment lower compartment.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the containment average air temperature not conforming to the above limits, restore the air temperature to within the limits within 8 hours or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.5.1 The primary containment upper compartment average air temperature shall be the weighted average ** of all ambient air temperature monitoring stations located in the upper compartment. As a minimum, temperature readings will be obtained at least once per 24 hours from the following locations: Location
- a. Elev. 743 ft.
- b. Elev. 786 ft.
- c. Elev. 786 or 845 ft.
4.6.1.5.2 The primary containment lower compartment average air temperature shall be the weighted average ** of all ambient air temperature monitoring stations located in the lower compartment. As a minimum, temperature readings will be obtained at least once per 24 hours from the following locations: Location
- a. Elev. 722 ft. -
- b. Elev. 700 ft.
- c. Elev. 685 or 703 ft.
" Lower limit may be reduced to 60*F in MODES 2, 3 and 4.
** The weighted average is the sum of each temperature multiplied by its respective containment volume fraction. In the event of inoperable temperature sensor (s), the weighted average shall be taken as the reduced total divided by one minus the volume fraction represented by the sensor (s) out of service.
~^
SEQUOYAH - UNIT 1 3/4 6-10
)
- CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM
.- LIMITING CONDITION FOR OPERATION 3.6.1.9 Three pairs (three purge supply lines and three purge exhaust lines) of con-tainment purge system lines may be open; the cpntainment purge supply and exhaust isolation valves in all other containment purge lines shall be closed. Operation with purge supply or exhaust isolation valves open for either purging or venting shall be limited to less than or equal to 1000 hours per 365 days. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With a purge supply or exhaust isolation valve open in excess of the above cumulative limit, or with more than one pair of containment purge system lines op n, close the isolation valve (s) in the purge line(s) within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within th2 following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.9.1 The position of the containment purge supply and exhaust isolation valves shall be determined at least once per 31 days. 4.6.1.9.2 The cumulative time that the purge supply and exhaust isolation valves are open during the past 365 days shall be determined at least once per . 7 days. SEQUOYAH - UNIT 1 3/4 6-15 wL . . . . - - ,
s . 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the calculated DNSR in the core at or above design during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded. The definitions of certain hot channel and peaking f actors as used in these specifications are as follows: Fq (z) Heat flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manu-facturing tolerances on fuel pellets and rods. F N Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of H the integral of linear power along the rod with the highest integrated power to the average rod power. 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) The limits on AXIAL FLUX DIFFERENCE assure that the Fg (z) upper bound envelope of 2.237 times the normalized axial peaking factor is not ex-ceeded during either normal operation or in the event of xenon redis-tribution following power change!.. Provisions.for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detec-tor outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed AI-Power operating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, RCS FLOWRATE AND NUCLtAR ENTHALPY RI5E h0i CHANNEL FACTOR The limits on heat flux hot channel f actor, 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. SEQUOYAH UNIT 1 B3/4 2-1 a e
' ,PONER' DISTRIBUTION LIMITS y.
4 BASES (Cont) Each of these is measurable but will normally only be detemined period-This periodic ically as specified in Specifications 4.2.2 and 4.2.3. surveillance is sufficient to insure that the limits are maintained provided:
- a. Control rods in a single group move together with no individual rod insertion differing by more than + 13 steps from the group demand position.
- b. Control rod groups are sequenced with overlapping groups as described in Specifiction 3.1.3.6.
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
Fyg will be maintained within its limits provided conditions a. tnrough d. above are maintained. As noted on Figures 3.2-3 and 3.2-4, RCS flow and FyH may be " traded off" against one another to ensure that the calculated DNBR will not be below the design DNBR value. The relaxation of FfH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. When RCS flow rate and Fh are measured, no additional allowances are necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4. Measur ent errors of 3.5 percent for RCS total flow rate and 4 percent for F have been allowed for in detemination of the design DNBR v ue. R1 , as calculated in Specification 3.2.3 and used in Figure 3.2-3, accountsforF[g less than or equal to 1.49. This value is the value used j inthevarioussafetyanalyseswhereF[,q influences parameters other than DNBR, e.g. peck clad temperature, and thus is the maximum "as measured" -l value allowed. R , as defined, allows for the inclusion of a penalty for forRodBowonDNbRonly. Thus, knowing the "as measured" values of FAH i and RCS flow allow for " trade off" in excess of R equal to 1.0 for the purpose of offsetting the Rod Bow DNBR penalty. ThepenaltiesappliedtoFfH t account for Rod Bow (Figure 3.2-4) as a function of burnup are consistent with those described in,Mr. John F. Stolz's (NRC) letter to T. M. Anderson (Westinghouse) dated April 5,1979 and
' W 8691 Rev. 1 (partial rod bow test data).
When an Fq measurement is taken, both experimental error and manu-facturing tolerances must be allowed for. 5 percent is the appropriate allowance for a full core map taken with the incore detector flux map-ping system and 3 percent is the appropriate allowance for manufacturing tolerance. The' hot channel factor F0 M(z) is measured periodically and in-creased by a cycle and height dependent power factor, W(z), to provide i assurance that the ifmit on the hot channel factor FQ (z), is met. W(z) accounts for the effects of nomal operation transients and was detemined from expected power control maneuvers over the full range of burnup conditions in the core. The W( ) function for nomal operation is provided in the Peaking Factor Limit Report per Specification 6.9.1.14. SEQUOYAH UNIT 1 B 3/4 2-2
= .
THIS FIGURE DELETED i r 4
Figure S 3/4 21 TYP1 CAL INCICATED AXfAL FLUX CIFFERENCO VERSU3 THERMAL PCWER SEQUOYAll UNIT 1 B 3/4 2-3
PCWER DISTRIBUTICH LIMITS BASES (Cont)
- 3/4.2.4 QUADRANT POWER TILT RATIO _
The quadrant power tilt ratio limit assures that the radial power dis-tribution satisfies the design values used in the power capability anal-ysis. Radial power distribution measurements are made during startup testing and periodically during power operation. The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does'not correct the tilt, the margin for uncertainty of Fo is rein-stated by reducing the power by 3 percent from RATED THER!tAL POWER for each percent of tilt in excess of 1.0.
. 3 /4.2.5 ONS PARN4ETERS The limits on the DNS related parameters assure that each of the para-meters are maintained within the normal steady state envelope of opera-tion assu.:ed in tne transient and accident analyses. The limits are consistent with the intial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.3 throughout each analyzed transient.
The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient opera-tion. , e e SEQUOYAH UNIT 1 B 3/4 2-3 4
ADMINISTRATkVECONTE0[S
- e. An unplanned offsite-release of 1) more than I curie of radioactive materfal in liquid effluents, 2) more than 150 curies of noble gas
- ,in gaseous effluents, or 3) more than 0.05 curies of radioiodine in
'~ gasecas effluents.' 'The report of an unplanned offsite release of radioactive material shall include the following information:
- 1. A description of the event and equipment involved.
- 2. Cause(s)' for the unplanned release.
3? Actions taken to prevent recurrence.
- 4. Consequences of the unplanned release.
f. Measured levels of radioactivity in'an environmental sampling medium
~
dete.rmined to exceed the reporting level values of Table-3.12-2 when averaged over any calendar quarter sampling period.
. . - i .
RADIAL PEAKING FACTOR LIMIT REPORT 6;9.1.14 The W(z) function for normal operation shall be provided to - the "irector, Nuclear Kehetor-Regulaticns, Attention: Chief of the Core )- Perforaance Branchy U. S. Nuclea'r Regulatory Commission, Washington, D.C.-20555 at least 60 days prior to cycle initial critic _ality. . In the, eventrehat these values would be submitted at some other time during core life, it will be rubmitted 60 days prior to .the date the values would become effective unless otherwise exempted by tle Commission. 4 Any information needed tf o support W(z) will be by request from the NRC and need not be included in this report. ) SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of tha Office of Inspection and Enforcement Regional Office within the time period specified , for each report. 6.10 RECORD RETENTION In addition to the applicable record reten, tion requirements of Title 10, Code l of Federal Regulations, the followir.g records shall be retained for at least j the minimum period indicated. The follcwing. records shall be retained for at least five years: l 6.10.1
- a. Records and logs 6f unit operation covering tim ~o interval at each power _ lev'el.
- b. Records and logs of principal mainten'ance activities inspections, repair and replacement of principal items of equipment related to
~ ~
I nuclear safety.
- c. All PEPORiABLE OCCURRENCES submitted to the Commission.
, d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
- e. Racords ofLchanges nade'to the procedures required by Specification
~6.8.1 and 6;S.4.
- f. Records of radioactive shipments.
- g. Records of sealed source and fission detector leak tests and results.
- h. Records of annual physical inventory of all sealed source material
, of record.
- SEduGYAH-UNIT 1 6-22
9
- APPENDIX B PEAKING FACTOR LIMIT REPORT s
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Sequoyah Nuclear Plant Unit 1 - Peaking Factor Limit Report This peaking factor limit report is provided in accordance with paragraph 6.9 1.14 of the Sequoyah unit 1 technical specifications. The cycle 2 W(z) function for RAOC operation is shown in figure 1. W(z) was calculated using the method described in NS-EPR-269, letter from E. P. Rahe (Westinghouse) to C. H. Burlinger (NRC) August 31, 1982. This W(z) function is used to confirm that the heat flux hot channel factor, Fg (z), will be limited to the technical specification values of: Fg (z) S
~K(E)_ h 9 > On 5 $
Md p _ Pg (z) 1 4M74 .K (a)3
~
io r P 1 05 This W(z) function, when applied to a power distribution measured under equilibrium conditions, demonstrates that the initial conditions assumed in the LOCA are met along with the ECCS acceptance criteria of 10 CFR 50.46. 4 l l {
*(a,c) i .e . g-HEIGHT FAX +r i .e (FEET) w(Z) (a.C) i... .15 1.000
- 45 1.000
- 1.000
- i... 75 1.05 1.000
- i.at 1.35 1.000
- 1.65 1.399 ie 1.95 1.365 2.25 1.328 i.io s
aai 2.55 1.289 8 2.85 1.250 8.i* , , 3.15 1.221 3.45 1.220
' 8' 8
> 3.75 1.216 4.05 1.219 i.3 ,
4.35 1.219 4.65 1.254
'l ,
4.95 1.284
'#* 5.25 1.309 l 1.329
, 5.55 i.e. 8 ' 5.85 1.352
- : 6.15 1.368 i.3, 6.45 1.376 2 6.75 1.376
; i. ,
,,s 8 7.05 1.367 c ,
7.35 1.350 i.to 7.65 1.323 - r 7.95 1.289 i.ie I 8.25 1.248 8.55 1.219 t s* 8.85 1.203 9.15 1.203 i 9.45 1.206 1.209 8" 9.75
] 10.05 1.217
< 10.35 1.233 i } 10.65 1.000 *
,,,, 10.95 1.000
- 11.25 1.000 *
' 11.55 1.000
- 11.85 1.000
- i.e. .
l I i.ee s.ee e.se it.w it.ee la.ee
..se e.se e.oo 3.ee
- e. i.so s.ee 3.co C::stE kEIGHT (FEET 3
- Top and bottom 15". excluded as per Technical Specification 4.2.2.2.g FIGURE 1 TYPICAL WESTINGHOUSE RELOAD CORE RAOC W(Z)
B-21
# g
- e
- G APPENDIX C RELOAD TEST PROGRAM e
G e
c* , The reload core design will be verified by performance of the following tests:
- 1. Control rod drop times,
- 2. Critical boron concentration measurements, 3 Control rod bank worth measurements using rod swap method,
- 4. Moderator temperature coefficient measurements, and
- 5. Flux distribution measurements using the incore flux mapping system.
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$8 ,
9 e APPENDIX D REMOVAL OF ROD CO.'JTROL RESTRICTIONS l
- ce .
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= . . . , ,
400 Chestnut Street Tower II s*.i n .- July 22, 1982 Director of Nuclear Reactor Regulation Attention: Ms. E. Adensam, Chier Licensing Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC .20555
Dear Ms. Adentam:
In the Matter of ) Docket Nos. 50-327 - Tennessee Valley Authcrity ) 50-328 In 1979 Westinghouse Electric Corporation identified to the NRC, Core Performance Branch, by letters dated November 15 and November 28, 1979 (Reference letter numbers NS-TMA-2162 and US-TMA-2167), a concern with regard to certain assumptions utilised in the dropped rod accident safety analysis applicable to some Westinghouse NSSS designs. This concern was derived primarily from the potential for an unanalysed power overshoot
.while in automatic control following selected dropoed rod events which did not result in a reactor trip. The concern was applicable to all Westinghouse plants which rely upon the power range neutron flux high-negative rate reactor trip to mitigate the consequences of the drooped rod accident. Operating plants were notified of an unreviewed safefy question under 10 CFR 50 59 and nonoperating plants notified of a significant deficiency under 10 CFR 50.55(e). Westinghouse reco= mended, and NRC subsequently required, certain operational restrictio'ns above 90-percent power (either manual rod control or restricted rod insertion limits when in auto =atic rod control) to address this concern on an interin basis and to provide further evaluation.
It is our understanding that a meeting was held between mem5ers of the Core Performance Branch staff and Westinghouse to discuss the Westinghouse dropped rod evaluation process. This process demonstrated that the DNB design basis can be met for this FSAR Chaoter 15 condition II event. We have been notified by Westinghouse that this evaluation process results in conclusions that will allow removal of the interim operating requirements en rod control and insertion. It is also our understanding that an agreement has been reached betwec:a Westinghouse and members of the Core Performance Branch staff that the removal of operating requirements would take place after the NRC review of the information subsequently submitted by Westinghouse letter dated January 20,1982 (Reference letter number NS-EPR-2545). This letter
' serves as notification that the debpped rod evaluation process docu=eited by Westinghouse letter NS-EPR-2545, dated January 20, 1982 applies to*
Sequoyah unit 1, cycle 2 and Sequoyah unit 2, cycle 2. The results confirn that the DNB design basis is met for the dropped rod accident. Based upon this method, it can be concluded that the interim restrictions on rod control and insertion will no longer be necessary. l
- 7. _ . - . . . _ _ _ . - _,- - ,_ ,.
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ar. ,.W g., ,g .. .. m . Director of Nuclcse Reactor Regulation . 7917 ?2, 1@,? We forrull/ request the I;3C to review the reteri71 tuh*::itted by Westinghouse OlG-?PR-29315, Janu try 70, 1932), and cuhsequently renovo the interin operationsl restrictions effective with the ntartup of cycla 2 for both units. Appro*al is needed b-rore startup of Sequoysh unit 1, cycle 2, presently scheduled for refueling cuta:;e in Se:tes5er 193?. '
~
If you have any questions concernint; this rntter, please get in te rch trith J. S. Wills at FTS 358-2683. Very truly yeten, TDPIESSE2 VALLF.Y AUTHORITY 1.
, ,( (.:. 21, y L. M. Mill:,' tanager
. {. . ' " " " ' ' . ,2j s !!uelear Licensing
-),' '
Sworn 'o. before e.e
- 1. ! Cf.fils'ckh' E d subsca' hod.ay of b b 1932 .-
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[* Notarr Public
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ec: U.S. Nuclear Regulatory Cc :.istien 3c:; ion II Attn: Mr. Jares P. O'Reillv, Regional .toninistrator 101 Parietta Street, Guite 3100 Atlanta, Georr;ia 30303 . i r t J 9 e 9 m eme e m .. .> g g
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