ML20101D415
| ML20101D415 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 03/05/1996 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20101D410 | List: |
| References | |
| RTR-NUREG-1431 NUDOCS 9603200253 | |
| Download: ML20101D415 (14) | |
Text
{{#Wiki_filter:' P QPTR 3.2.4 3.2 PO'.'ER DISTRIBUTION LIMITS 41Z limih TWei4r. pow ..___ wo7g....
%38/, bc/aw 49 b pdmances of 3.2.4 QUADRANT POWER TILT RATIO (QPla)
-for uc.h I % a guired Ac4 lea A.1.2., ,
4h. Cepk6n W 1 I,o o LCO 3.2.4 is m u d red b He. The QPTR shall be s 1.02. compklisr c4' 5ft 3.7. 4.i._ APPLI.CABILITY: MODE 1 with THERMAL POWER > 50% RTP. 2.b0#J Vf ACTIONS ) CONDITION REQUIRED ACTION PLETION TIME Limit be low A.1Y QPTR not within limit.A.1 Reduse HERMALPOWER/o 2 hours 2: 3% RTP for each 1% of QPTR S
> 1.00.
N.i 9e/4r>r1 M 3'1'Y' I" "*[' o rs AND th;r;;ft:r E < woc-oz./ C.9 / R. i Ag3 Perform SR 3.2.1.1 and SR 3.2.2.1. #p ~.. , 6
- Once per 7 days
- thereafter f 12.tqyired Ae.eas *A.h- -- - - N()TE -
mvu4 be C.omplfled
~ ~' -~l A. Reevaluate safety Prior to
( whenever IRgwred 4tih analyses and confi m increasing A,6 6 m eeafed results remain val id THERMAL POWER
,,,_,,,,.' -- for durati , of above the limit operation 6.her th is of Required condition. ActionA. land b 2. 2.
(continued) (d)tW1/n 7.4 bow:l}Atf ac/Weenty 6qttilibtiwrionJHiM.5yy)j
' 171fAMet /@WEf jimifec/by ;
I htitdited hisns A*J MAL A.L ea WOG STS 3.2-18 Rev. O,09/28/92 { 9603200253 960305 ~ ' PDR ADOCK 05000424 P PDR
'. 1 QPTR 3.2.4 4 1
l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
.5 S A. (continued) A. M --------
--------NOTfired Perform Ret u 1 Action A.5 E only I after Required l
'[j Action A. is s
completed. Calibrate excore Prior to
/C detectors to show increasing asse-QPTRi= / 00 THERMAL POWER above the limit of Required Action A.104 4 g A .1.. L tev-osopI A. h --------No ------- C- --tM 76 ---
PerfoFm 'Re auired on age o//de g//fua Action A. only after Required g 7/g fy+ ,, j-Action A. is ggh #g completed._____________3________
, ,,g .
_ ,,,, __. _ _,, _ _ _ y l Perform SR 3.2.1.1 bithin24 hours and SR 3.2.2.2. after reaching g worn-02 RTP C.1 - 93 Within 48 hours after increasing THERMAL POWER above the limit of Required Action A.1 oncl A.L7 O B. Required Acti'on and B.1 Reduce THERMAL POWER 4 hours associated Completion to s 50% RTP. Time not met. s WOG STS 3.2-19 Rev. O, 09/28/92 . :_ . x-_________-_ _:_: - x - _-
et ne deredtn ef e emMen 10 QPTR k, ' oc.cichar.cc u.nik Ce tiers R of tMS B 3.2.4 BASES [ ACTIONS (continued
}he. incore
-psreg de*c/ ors
. guess radial p r distribution that requ es an OFSg investigation a d evaluation that is acc lished by examining the power distributio . Specifically, the core peaking f ctors rf ti: ;ria.. ".;P. must be evaluated
,/ because they re the' factors that best characterize the core
,[! power distri ation. This re-evaluation is required to j ensure that, efore increasing THERMAL POWER to above the limit of Required Action A. , the reactor core conditions $eokig )
Ahvs) Hire consistent with the as options in the safety analyses. , l* A WA NWM - LG 0W-I M0A,2,2 as/ w il k ai w l 4cmr) M afgr.jpgafzu h l h 00 MP , [ fry /h.-few/G:,n?:c.d if the QPTR has exc ded he 1.02 limit and a re-evaluation 8pgrfycherHv/ of the safety analysis is completed and shows that safety requirements are meu, the excore detectors are recalibrated M h/M i 0 laws to show a4epe QPTR prior to increasing THERMAL POWER to g I th0y // S p R /D above the limit of Required Action A.1C This is done to '
/cfum 4 4/#4 /ArJ i detect any ,sy pequent W changes in QPTR.#
; ###I # Require Action M i iYi ya o that states that reca/s}fd/c @ the QP is not nn;d ;_ until after the re-evaluation of r w yr c/r/tcr or, the safety analysis has determined that core conditions at TSC RTP are within the safety analysis assumptions (i.e.,
h Required Action A7 M ). This Note is intended to prevent any ambiguity aberut the required sequence of actions. 4 - A A,5 (~U Once' ,th: fl:_gt "-t QPTR is :(.cnd QWlh"l =t (i .e. , lequired Action A #eb is performed), it is acceptable to return to full power operation. However, as an added c 1eck that the core power distribution at RTP is consistent with the safety analysis assumptions, Required Action Arht requires verification that F,(Z) and F2,, are within their specified limits within 24 hours of reaching RTP. As an added precaution, if the core power does not reach RTP within 24 hours, but is increased slowly, then the putking factor surveillances must be performed within 48 ho'a s of the time when the ascent to power was begun. These onoletion Times are intended to allow adequate time to inc ease THERMAL POWER to above the limit of Required Action A.1 while not Md 42.7. (continued) WOG STS B 3.2-45 Rev. O,09/28/92 _ ._- . . = - . . _ _ _
F56 INSERT FOR REQUIRED ACTION A.4 BASES PAGE B 3.2-45 However, if prior to performing SR 3.2.1.1 and SR 3.2.2.1, QPTR is restored to within the limit, either due to prior completion of Required Actions or due to core performance characteristics that result in the QPTR out-of-limit condition correcting itself, Required Action A.3 and any other Required Actions would no longer apply because Condition A of LCO 3.2.4 would be exited in accordance with LCO 3.0.2 due to restoration of full compliance with LCO 3.2.4. g g ,z 4.cV-023-J Ifit is determined that a sustained change in the dial power distribution has occurred, and Required Action A.3 has been completed with s isfactory results, an increase in THERMAL POWER above the limit of Required Action A. may be appropriate. The necessary sequence of Required Actions, beginning with Required A.4, would be as follows prior to increasing i THERMAL POWER above the limit ofRequired Action A.1@d A,Z,E CV-0fd3-I j
- 1. Verify by the reevaluation of the safety analyses that after the sustained change in radial power distribution, the core conditions remain within the assumptions of the safety analyses and will remain so after return to RTP (Required Action A.4), ,
i and
- 2. Recalibrate the power range detectors to reset QPTR to 1.00 (Required Action )
A.5). If these actions are completed with satisfactory results, THERMAL POWER may be increased ) above the limit of Required Action A. After power is increased, the peaking facto. .re again verified to be within limits. Upon the atisfactory completion of Required Action A.6, Condition A of LCO 3.2.4 can be exited. 1 and A.L. 2. Lc)-0(oo3-I ,
~
i RCS Loops - MODE 5, Loops Filled l B 3.4.7 BASES LCO a. No operations are permitted that would dilute the RCS (continued) boron concentration, therefore maintaining the margin to criticality. Boron reduction is prohibited because-a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation; and
- b. Core outlet temperature is maintained at least 10*F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction. l Note 2 allows one RHR loop to be inoperable for a period of up to 2 hours, provided that the other RHR loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time whegugtestin is safe and possible. l 75C Note 3 requ' e t t,the secondary side water temperature of eachSGbe$ 0 F above each of the RCS cold leg tem)eratures before the start of a reactor coolant pump I during Mode 5 WiMa - (RC w ith = ecs = u n ; t m t = = ; = ; 7 . Tus j
.//g gCS /osgs8//ec/. restriction is to prevent a low temperature overpressure j event due to a thermal transient when an RCP is started.
Note 4 provides for an orderly transition from MODE 5 to f4ddihma/nguinmeAh M MODE 4 during a planned heatup by permitting removal of RHR i on SG % de D*//06/s 434 loops from operation when at least one RCS loop is in
/)ed 5/ok aret operation. This Note provides for the transition to MODE 4 o,4c5 / oops ed red'/#' where an RCS loop is permitted to be in operation and l gruwrarests/4//'N*'d i
knfing conspleftifnd replaces 1 oops. the h:f{n.~d function provided by the RHR*- Wk"% O'N KS circ.u/afton l b, 8Cdpreswe dd/4/d/# RHR pumps are OPERABLE if they are capable of being powered
>/00/ m3 5/&c /k and are able to provide flow if required. An OPERABLE SG mas /recan/8/hp ) can perform as a heat sink when it has an adequate water l
' Md rtr#ig, j level and is OPERABLE in accordance with the Steam Generator Tube Surveillance Program.
[ Ps E l APPLICABILITY In MODE 5 with RCS loops filled, this_LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, j , (continued) Het 2 B 3.4-34 "e.. O, M/Zoisi Yog-{lt h SUN
- . . . _ - _ . ~ . . -- - _. - .
INSERT FOR PAGE B 3.3-53 Axial offset is the difference between the power in the top half of the core and the bottom half of the core expressed as a fraction (percent) of the total power being produced by the core. Mathematically,it is expressed as: 4 (Flux7 - Flux,) Ao = 100 x .p LcV-or,03-1 (Power)(Flut -Flux,) where Fluxr= neutron flux at the top of the core, and Fluxa= neutron flux at the bottom of the core The relationship between AFD and axial offset is: AFD= AO x(Power (%)/100) AFD as displayed on the main control board and as determined by the plant computer use inputs from the power range NIS detectors which are located outside the reactor vessel. Axial offset is measured using incore detectors. The surveillance assures that the AFD as displayed on the main control board and as determined by the plant computer is within 3 % of the AFD as calculated from the axial offset equation. Agreement is required so that the reactor is operated within the bounds of the safety analysis regarding axial power distribution.
. . . . . - .~ _ -. . _ - .-- . --
~. Definitions 1.1 1.1 Definitions (continued) CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. CHANNEL OPERATIONAL A COT shall be the injection of a simulated or TEST (C0T) actual signal into the channel as clos 3 to the sensor as practicable to verify the FERABILITY of and trip required functions. Thealarm, COT shallinterlock, include a d:;;.g)djustme necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. CORE ALTERATION CORE ALTER #shall be the movement 3 any fuel, SWR 05 sources,#eactivity control components,Cor othTb c3 c: ;:r" 7":" t; . ...t '. . ' t., within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of COE ALTERATIONS shall not preclude completion of movement of a component to a safe position. _ CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each c .f_ o g _ z j ___, g reload cycle in accordance with Specification
@ . . ^ . . 0.
addressed i operation within these limits is individual Specifications. UnH l DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and 1-135 actually present. The thyroid-dose conversion factors used for this calculation-shall be those listed in F d k ::: J T!; ; 00,
"::, .:::, ":::::::M :. :? ::.1. ..... .% . . . . ^. . .
' ~
":u: :-M T;;; ^ .. wr E'.;;;," ;r M. .; i k u ,..
Table E-7 of Regulatory Guide 1.109, Rev.--1, NRC, 1977 . (continued)
"O^ 373 1.1-2 90'? ? , "" '9 /S (9 He Mih .tdc
Definitions l l'1 Ok/}he. nminn / POAVSeQinf.S 1.1 Definitions Sf & cow gnhcgjgg WffMWC fj,f4,7 , i '
)
PHYSICS TESTS c. Ot lerwise approved by the Nucleer Regulatory (continued) Consnission. PRESSURE AND The PTLR is the unit specific document that g ,o,gq TEMPERATURE LIMITS provic es the reactor vessel pressure and REPORT (PTLR) tempe}'ature limits, including heatup and cooldown j ) rates for the current reactor vessel fluence 5 pertod. These pressure and temperature 1 #R shall be determined for each fluence iod in accordance with Specification 5.0.1.,. Mmt l/4// operation within these operating limits is ! /4dividro/5pec/Aa//r.t. addrgsed ing3.(.g - :: [. m..y.mg ,,"""
, 6 C E :i _: L'r ..
. m m un'nd"" '?"
"I W r.ori
~,m....
QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper RATIO (QPTR) excore detector calibrated output to the average of the upper excore detector calibrated output 4, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) MWt. ratetothereactorcoolantof;00'b .35(o REACTOR TRIP The RTS RESPONSE TIME shall be that time interval SYSTEM (RTS) RESPONSE from when the monitored parameter exceeds its RTS TIME trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. 1 SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of ' reactivity by which the reactor is subtritical or would be subtritical from its present condition assuming:
- a. All rod cluster control assemblies (RCCAs) are l fully inserted except for the single RCCA of BWR-l8 highest reactivity th, which is assumed to C.L be fully withdraw me (A>/M OAy 4CC4 M/
Capoebk. d bei I/y in:,erfeof -Mc(VJcHvil ? (mf# cf.Me AC MWs/ bc. Mca,MNdAf Jh 'j,Ac f 1 dehrminok asswt id / (continued) i
'a , 1.1-5 L,. O, G^/^0/02-
VEGP SPECIFIC INSERT High Flux At Shutdown Alarm 3.3.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l SR 3.3.8.1 Perform COT. 92 days i SR 3.3.8.2 Perform CHANNEL CALIBRATION. 18 months
- m M0T E -
L _ got required be g en ft,rin
. perarmed Prior MODC3 frm Mo06 Z Mh.l # M '
I cd- oas .t . g en} snh M006 3. l
- _ _ ~ } ,
L__ , I 1 Vogtle Units 1 and 2 Page 3.3.8-2 Rev. O,10/18/94
, HFASA l B 3.3.8 L BASES l 2 SURVEILLANCE The HFASA channels are subject to a COT and a CHANNEL REQUIREMENTS CALIBRATION. SR 3.3.8.1 SR 3.3.8.1 requires the performance of a COT every 92 days to ensure that each channel of the HFASA and its setpoint are OPERABLE. This test shall include verification that the HFASA setpoint is less than or equal to 2.3 times background. The frequency of 92 days is consistent with the requirements for the source range channels. J SR 3.3.8.2 SR 3.3.9.2 requires the performance of a CHANNEL CALIBRATION every 18 months. This test verifies that each channel responds to a measured parameter within the necessary range and accuracy. It encompasses the entire instrument loop, l including the sensor. The frequency is based on operating experience and consistency with the typical industry refueling cycle. REFERENCES 1. FSAR, Chapter 15.4.6. l Thi; $dotill04Cl Atf Mif'**0l 5I!"Ob$'CIby a bl*N lbdlff0 VICES A 9-b94r de/g in-lh regwru,uf h ahw f ni'swvei//osce 4<- JAe //hene inskerfong uptenkring M@f3 R>m mod 6 2. TNs /\We o/hw.s aiswmo/ she%; ibprwed wHAoutddoy &/k pvfor>ronce. of #< surveilloxe. h n,eef-
& ayfa:a'b'//h itgwreme.<rk H M00s 3.
Ld- OG05- f 1 Vogtle Units 1 and 2 B 3.3 - xx
L tv- oc, 0 3 - I L ! CHAPTER 5.0 ADMINISTRATIVE CONTROLS i INSERTS l l INSERT 21 TO PAGE 5.0-21 5.6.6 PTLR 1 l LC " CS Press perature Limit! LCO 3. Protection System! l 1 The RCS pressure and temperature limits for Unit I shall be those previously i reviewed and approved by the NRC in Amendment No. 87 to Facility Operating License NPF-68. The RCS pressure and temperature limits for Unit 2 shall be those previously reviewed and approved by the NRC in Amendment No. 65 to Facility Operating License NPF-81. The acceptability of the P/T and COPS limits , are documented in NRC Letter,"Vogtle Electric Generating Plant, Units 1 and 2 - Acceptance for Referencing of Pressure Temperature Limits Report," February 12,1996. Specifically, the limits and methodology are described in the following : i documents:
- 1. Amendment No. 87 to Facility Operating License No. NPF-68, Vogtle t Electric Generating Plant, Unit 1, June 8,1995.
l
- 2. Amendment No. 65 to Facility Operating License No. NPF-81, Vogtle Electric Generating Plant, Unit 2, June 8,1995. !
- 3. Letter from C. I. Grimes, NRC, to R. A. Newton, Westinghouse Electric Corporation," Acceptance for Referencing of Topical Report WCAP-14040, i Revision 1, ' Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,'" October 16,1995.
- 4. Letter from C. K. McCoy, Georgia Power, to U. S. Nuclear Regulatory Commission, Attention: Document Control Desk,"Vogtle Electric Generming Plant, Pressure and Temperature Limits Report," Enclosures 1 and 2, January 26,1996.
I I d. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
- Reporting Requirements
. 5.6 5.6 Reporting Requirements (continued) 5.6.4 Monthly Ooeratina Reoorts Routine reports of operating statistics and shutdown xperienc ,
including documentation of all challenges to the pre urizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th f each month following the calendar month covered by the report. 5.6.5 CORE OPERATING LIMITS REPORT (COLR)
- a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
IhlSERT) The indivtauas now;Tk:t'a-e +b+ t:;; sm e operating L ji limite -"-t L ..iaranced here,
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
M haical Report (s) by number, tit
~
fgg/) s NRC staff approval e staff Safety D ' Evaluation Re an speci o by NRC
, ate.
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d. The COLR, including any midcycle revisions or supplements,
, shall be provided upon issuance for each reload cycle to the i NRC. l 1 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LM REPORT (PTLR)
- a. RCS pressure and, temperature limits for heat up, cooldown, I 1m ' : ;rh peration, criticality, and hydrostatic l Ld-0603-I (continued)
WOG STS 5.0-20 Rev. 11/16/94
m ycyeohcl/t//c[Mt //8JE8/r4s v//ec/A Reporting Requirements sa ge 56 (c$ps)t M Co/dOkw sho/Ibe es reuun liskd and'docu hrfecMdit clitylhe (htR EvHeA/Ated: I. . L eReporting
.6 e 3 ,4,IRequirements z ** M Q W f ttJ M t hoh clion 91&s
5.6.6 Reactor Coolant Svitem (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR1 (certinued) testing as well as heatup and cooldown rates shall be established a-d documented in the PTLR for.the following: E O b U dF .". N Y N? [ '" ~ ~ ~"' " ""
~
b. W5N$ 55.S Pteii5t[Mi&ttMhre llVT)Limils" ine nai u im9tno use to ete - ne t e nu pre sur [ an te er ure imi s sh 1b the pre ious y re 1e d i a a ro db th NRC, spec fica yt se d scri ed n oil in doc en : [ dent fy t NR staf ap ov idoc n by 'te. t , r
- c. 'Th PT sh 'b pr ide to t e NR upon iss nc f e lMS627 r et ye - lu ce p iod nd f any rev io o ll papp me t et . j s E , -
Neviewers' Notes: The methodology for the calculation of the P-ts for NRC ape-nval should include the following provisio .
- 1. methodn'- ' shall describe how the neutr)n flue is calc ated ' 'erence new Regulatory Guide when i ued).
- 2. The Reac r " sel Material Surveillance Pr am shall comply wit ' m ndix H to 10 CFR 50. Th eactor vessel material ir '$ tion surveillance spec' en removal schedule shall be pr- d
.. along with how t specimen examinations shall be us 'o u ate the PTLR rves.
- 3. Low Tempera' n Overpre ur rotection (LTOP) System lift setting lir for the Po Operated Relief Valves (PORVs),
developed u- NRC-ap ove ethodologies may be included in the PTL".
- 4. 'The adjustn' .crence temperature T) for each reactor beltline ir "1 shall be calculated, counting for radiatio -
'tlement, in accordance wi Regulatory Guide 1.99, vis' '. ,
- 5. limitir '- shall be incorporated into the culation of the pree -
and temperature limit curves in acc ance with N'JREG ~ 9 Standard Review Plan 5.3.2, Pressure-Temperatur- "its. _ L.CN' 0(p0 (continued) WOG STS 5.0-21 Rev 1, 04/07/95
e, . e
, ECCS Recirculation Fluid pH Control System j ID 3.5.6 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.6 Recirculation Fluid pH Control System LCO 3.5.6 The Recirculation Fluid pH Control System shall be OPERABLE.
APPLICABILITY: MODES 1,2,3, and 4. l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation Fluid pH A.1 Restore system to 72 hours Control System OPERABLE status, inoperable. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 84 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.6.1 Perform a visualinspection of the Recirculation 18 months Fluid pH Control System and verify the following: a) Three storage caskets are in place, and b) have maintained theirintegrity and g -caff8Mo M /Of LcV- Of,0.5- I c) Mbasked!! .d ./,;h a 11,484 (220 cubic feet) and s 14,612 pounds (260 cubic feet) of trisodium phosphate crystals. Vogtle Units 1 and 2 3.5-1 DRAFT - .-- ._.}}