ML20092K674
| ML20092K674 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/20/1992 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20092K666 | List: |
| References | |
| NUDOCS 9202250338 | |
| Download: ML20092K674 (26) | |
Text
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- i t .
).
f ENCLOSURE 1 i
PROPOSED TECllNICAL SPECIFICATION CllANGE ,
! SEQUOYAH NUCLEAR PIANT UNIT 1 t
f DOCKET NO. 50-327
('IVA-SQN-TS-92-02) i LIST OF AFFECTED PAGES i
j UniL1 1
4 2-6.
j 2-10 1
3/4 3-11 3/4 3-27 3/4 3-37 i
b 1-f 4
a d
4 I
4 d-4 1-i
' kRO2250338.920220 p ADOCK 05000327
, PDR
v, E TABLE 2.2-1(Continuedl
~
8< -
I REACTOR TRIP SYSTEM Ill5iRUMENTATION TRIP SETPOINTS l E .
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E l q 13. Stea:a Generator Water
.y Level--Low-tow !
t
h MS f
Coincident with :
Steam Generator Water 1 15.0% of narrow range Level -- Low-Low (Adverse) instrument span
> 14.4% of narrow rang = juss instrtment span
' and Containment Pressure - EAM i 0.5 psig 1 0.6 psig "" ,
or
{t Steam Generator Water 1 10.7% of narrow range > 10.1% of narrow range i
y Level -- Low-Low (EAM) instrument span instrument span lus3 -
with A time delay (TS) if one -<T 5 (Note 5)
Ster:: Generator is affected -< (1.01) T5(Note 5) w g or q E A time delay (T ) if two or R143 '
S moreSteamGeneYatorsare -<T* (Note 5) 1 (1.01) T" (Note 5) .
7+ affected o
~
t
- b. RCS Loop AT Equivalent to y
i j Power > 50% RTP m Coincident with
]
Steam Generator Water ' > 15.0% of narrow range > 14.a% of narrow range i
g_g Level -- Low-tow (Adverse) instrument span jgi33 m
instrument span
~
and 2 -
Containment Pressure (EAM) 5 0.5 psig 1 0.6 psig -
gg3
-g _w or 5@ Steam Generator Water > 10.7% of narrow range > 10.1% of narrow range "5' Level -- L @ _ .iast rt=. um ,
O I3 amenswr sg Tes Res Loor AT Yn.tMtc LP.;Pur Au oCABLE N uc Cr 6: Tor 5ctrorvr z.s%g7 7gtcoutco 1 To LTee screomT + 2A% F:re Djuw uurri tyuc 6 Orat$ rico Becaust c, Lescasen +q70 -
untrsTeactics.~rms N.cersom %.us ReputTies Exciao ArTse ho Or timri tytte 6 ortsme, _
m TABLE 2.2-1 (Continued) m .
g REACTOR TRIP SYSTEM INSTRLHENTATION TRIP SETPOINTS E
g NOTATION (Continued) x NOTE 2: (Continued)
E T
3
= Time constant utilized in the rate-lag controller for T,, ,T 3
= F secs. W K = 0.0011 for T > T* and K 6 6 = 0 for T 1 T" f
T = as defined in Note 1 T"
avg at RATED THERMAL POWER (Calibration temperature for
= Indicated T AT instrumentation, i 578.2*F) I S = as defined in Note I f (AI) = 0 for ell AI 2
E>
NOTE 3: na trip setpoint shall not exceed its computed trip point by more than 1.9 percent AT span.*
i l NOTE 4: ch=are!' -t W ri 1p point by more than R145 l
2g 1.7 percent AT span. $ p setpoint shall not exceed its computed t:
$ , h,
-$ s Tat Rec uistE M EWT FD e. THE O V ERhPER ATrut AT Bucer. rot _c fAwe Cr /.9 A xecr/r 47 $ppu /
7 (!get Tsc CcetPuTED IP le ScTPoinT- L keoutco To 1.t , P E R m rt- AT S een Duoc uom i !
85 tyttt is OeceariorJ betAt4se Or Inca.ewsco R T O t4 " ce t T n era ri es . T si3 Atteenacv%uc
- Renatrion EX PI RE3 bT THE EINd 0F LIMIT 1 C y tt.E ls C PER ATs c rJ. ,
4 Tus Recuite ment Fen Tat OvctroWER AT Anuc u n e,tc Vat ut or I-7 Potteur AT 3ra N )
E. houc Tee Co m raTrn Trip stTPo int I.s Rcouten To 1.t, PcRteur AT SPAe Duurs (Arnr 1 4
Cy tt_t lp CPERAT1ou BctAttse CF L cR.E%5ED RID LA NtCETA I NTI ES . TH is $'tc W A 8 tE V ALu E Rtnutrien Eamms Ar Tat End CF LiutT ycLE (c CPER%TioM. ,
- ~ w -
i - . A }
- w .
' ,Y TABLE 4.3-1 E
REACTOR TRIP SYSTEM INSTRUMENTATICH SURVEILLANCE REQUIREMENTS OIAt03EL PDCES IN WICH i CIIAt#4EL CHANNEL FUNCTIONAL E FUNCTIOilAL UNIT SURVEILLANCE g OlECK CALIBRATION TEST REQUIRED w 1. Manual Reactor Trip N. A. N.A. S/U(1) and R(9) 1, 2, and *
- 2. Power Range, fleutron Flux 5 D(2), M(3) Q 1, 2 and Q(6)
- 3. Power Range, Neutron Flux, N. A.
fligh Positive Rate R(6) Q 1, 2
- 4. Power Range, Neutron flux. N.A. R(6) Q liigh Neg:stive Rate I* 2 g 5. Intermediate Range. S R(6) S/U(1) I= 2* and "
a Neutron Flux Y 6. Source Range, Hautron flux 5(7) R(6) v
" M and S/U(1) 2, 3, 4, 5. and *
- 7. Overtepperature Delta I S - R 56 s Q 1, 2 R145
- 8. Overpower Delta T S 2 ** ! Q 1, 2
- 9. Pressurizer Pressure--Low 5 R Q 1, 2
- 30. Pressurizer Pressure--High 5 R Q 1, 2
- 11. Pressurizer Water level--ifigh 5 R Q 1, 2
- 12. Loss of flow - Single Loop S R Q 1
- 13. toss of Flow - Two Loops S R N.A. I
' fC y 14. Main Steam Generator Water
, 3@ 'n Level--Loc low 3 v. 3 A. Steam Generator Water Level -- S R Q 1, 2 R145
- c ,", Low-Low (Adverse) a x B. Steam Ger,erator Water Level --
' 5 R 0 1, 2
$,.C .o,,,
low-low (EAM)
. C. RCS loop AT 5 g gg
- D. Containment Pressure (Ef41) S g
{ y Q 1, 2
- Fee uw i Cytu (, O FatATicab L LIEU CF A CuArm h C A t l6R ATv >> /7 "TECNrJ/ce/L '
E V A L L M T icrJ 1.l As BEEN ?atrcst mED fis DEtiNE ATTO lu istwnicgt SPctiricgr;cf; (gge;gg i D
Q uesT '4 2 - 0 2. .
~
p E *
?
, TASLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURE ACHIATION SYSTEM INSTRUMENTATION TRIP SETPOINTS .
FUNCTIONAL UNIT TRIP SETPOINT ALLOVABLE VALUES g 6. AUXILIARY FEEDWATcR
[ a. Manual Not Applicable Not Applicale l
- b. Automatic Actuation Logic Not Applicable Not Applicable
- c. Main Steam Generator Vater Level--Low-Low
to Power $ 0% input <50% RTP
- ~ W setpcint gi45 l
1
+2.5% RTP#h l
{ Coincident with Steam >15.0% of narrow range >14.4% of narrow lgi33 l Generator Water level-- instrument span range instrument span Y Low-Low (Adverse)
O and RI45 Containment Pressure-EAM $0.5 psig 10 6 psig or Steam Generator Water >10.7% of narrow range >10.1% of narrow Instrument span instrument span l R155 Level--Low-Low (EAM) 3' with I
@ A time delay (T ) if one i T3 (Note 5, Table 2.2-1) $ (1.01) Tg (Note 5, g 3 R145
. Steam Generator is affected Table 2.2-1) 5 or g A time delay (T,) if two 1 T,(Note 5. Table 2.2-1) $ (1.01) T,(Note 5,
.or more Steam Generators Table 2.2-1)
,3 -
are affected .
V C
CE 4 'Tgt Emut Re r4Ewr PC* Tar RCS Loco AT VAetaste Icapur AttoC As't VA'uc Cr A.
6 Tgie s tT Poi err + 2. 5 % R.TP Is ICD"tg T L 'Teie serPcirer + 2.1% q Due,a M uen 1 Cycte & Cett ATico kt^"Sc Cr _L ctmsen RTD uucccrawTics. tais g5 Auccene,tt VAtus Renu tTico EXPRES NT C ECTO CF UPJff 1Y C E'E b0
- j co A /
r ^~ f
] .
t b
- v. TABLE 4.3-2 (Continumi) l m ^
E EriGINEERED SATETY FEATURE ACTUATIOM SYSTEM IMSTRUMENTATION j SURVEILi#tCE REQUIREMENTS
- CHANNEL M) DES IN WICH j SURVEILLANCE j FUNCTIONAL E CHAMMEL CHANNEL TEST REQUIRED O FUNCTIONAL UtlIT CHECK CALIBRATIOM v
- c. Main Steam Generator Vater Level-tow-Low f
R Q 1, 2, 3 .
L Stean Genarator Water S Level -- Low-tow ;
(Adverse) i 5 R Q 1, 2, 3
- 2. Steam Generator Water gg43 ,
Level --- Low-Low s (EAM) w 3. RCS Loop AT S R* Q 1, 2, 3 i 5 R Q 1, 2, 3 b 4. Containment Pressure (EAM)
- d. 5.1. See 1 above (ali SI surveillance requirements)
Station Blackout N.A. R N.A. 1, 2, 3 e.
N.A. M.A. R 1, 2
- f. Trip of Main Feedwater '
Pumps Auxiliary Feedwater Suction N.A. R M 1,2,3
- - 3 g. !
Pressure-tow
.$..< og N.A. 1, 2, 3
- h. Auxiliary Feedwater Suction N.A. R f !} Transfer Time Delays O>gl' ,+
- 7. LOSS Of POWER lR145 g
O - a. 6.9 kv Shutdown Board -
O y Loss of Voltage R M 1, 2, 3, 4 t
l g 1. Start Diesel Generators S N.A. 1, 2, 3, 4 l
S R
@ 2. Load Staeddi
- $ Fo n. LAra
- T .1 Cycte is C PER A T icro, Ira 1_i eta OF b (W ANN EL ( A ll6 ilATICrJ ,.4, O 7EtH1Vit4L EV A ttA ATIC f3 NAS EEEN MFCRmsD b5 hL:NE AYi'O M TEt.H fontA t- I SPEtt F IC ATIOPJ CHAtor,E kEGlAEST 97-02. A 2 ' % j f
~ - - - - - - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
TJ4 CLOSURE 2 PROPOSED TECilNICAL SPECIFICATION CilANGE SEQUOYAll NUCLEAR PLANT UNIT 1 DOCEET NO. $0-327 (IVA-SQN-TS-92-02 )
DESCRIPTION AND JUSTIFICATION FOR REVISION OF OVERTEMPERATURE DIFFERENTIAL TEMPERATURE, OVERPOWER DIFFERENTIAL TEMPERATURE, AND REACTOR COO! ANT SYSTIM (RCS)-140P DIFFERENTIAL TFNPERA1URE ALLOWABLE VALUES AND CALIBRATION REQUIREMENT FOR RCS RESISTANCE TEMPERATURE DETECTORS h
_J
Description of_ Change TVA proposes to modify the Sequoyah Nuclear Plant (SQN) Unit 1 Technical Specifications (TSs) to revise the reactor trip allowable values for overtemperature (OT) differential temperature (AT) and overpower (OP) AT from 1.9 and 1.7 percent AT span respectively to 1.6 percent AT span. These changes affect items 7 and 8 of Table 2.2-1 through the associated notes 3 and 4 of this table. In addition, the reactor coolant system (RCS) loop AT allowable value of +2.5 percent reactor thermal power (RTP) for item 13.a in Table 2.2-1 and item 6.c.i in Table 3.3-4 will be revised to 42.4 percent RTP. The channel calibration requirements for the RCS resistance temperature detectors (RTDs) associated with OTAT, OPAT, and RCS loop AT functions, items 7, 8, and 14.0 respectively of Table 4.3-1 and item 6.c.3 of Table 4.3-2, will be revised to utilize the technical evaluation presented in this TS change request in lieu of an RTD cross-calibration.
These changes are requested for Unit 1 Cycle 6 operation only and will expire at the end of this fuel cycle.
Reaso:Lior_ Change During the start-up of Unit 1 f or Cycle 6_ operation, TVA performed RTD cross-calibrations to verify accuracy. At the upper-temperature plateaus, the data for this calibration effort appeared to be skewed. TVA investigated and identified errors in the data as a result of test-instrwnentation appilcation. The use of a data logger on the Eagle 21 analog test points had introduced a random bias on the temperature reading.
This bias skewed the data such that the intent of the cross-calibration to verify a tolerance of 10.5 degree Fahrenheit (F) could not be achieved.
Westinghouse Electric Corporation was requested to provide an evaluation of the RTD accuracy based on Cycle 5 factory RTD data and maximum expected RTD uncertainties. This evaluation concluded that the RTDs would meet an accuracy of 11.2 degrees F for the beginning of'the Cycle 6 fuel cycle.
With the RTD accuracy of 31.2 degrees F, Westinghouse determined that the effect would require a slight reduction in the OTAT, OPAT, and RCS loop AT allowable values to support the safety analysis. TVA discussed these changes with NRC and-agreed to administrative 1y control these values while processing this TS change for the reasons described above. The Westinghouse evaluation also serves as the basis that SQN Unit 1 has met the intent of the 18-month channel calibration requirement for OTAT, OPAT, and RCS loop AT functions that utilize RCS RTD inputs. This evaluation will be used for Unit 1 Cycle 6 operation instead of the RTD cross-calibration data that was skewed as previously described.
The proposed changes are needed to properly reflect more conservative allowable values that are consistent with the safety analysis and to prevent an unnecessary midcycle unit shutdown to perform an additional RTD cross-enlibration.
Juntification_for_ Change The trip setpoint and allowable value for the OTAT reactor trip function have been designed to protect the reactor from reaching an unacceptably 109 departure from nucleate boiling ratio. The RCS hot- and cold-leg RTD
4
.- 2-temperatures are inputs to the process circuitry that calculates this trip setpoint on a continuous basis. The 'tTD inputs are provided to the circuitry with one set of temperature measurements for each RCS loop. The OPAT reactor trip iunction protects against excessive power that could exceed the power per foot rating of the fuel rods. This setpoint is also calculated on a continuous basis and utl11res RTD temperature inputs from each RCS loop.
The reactor trip function of the low-low steam generator (S/0) levet protects against the loss of teactor heat sink in the event of loss of feedwater to the S/Cs. With the reactor protection system upgrade that included the installation of the Westinghouse Eagle 21 system, SQN implemented a trip time delay (TTD) leature associated with the low-low S/0 level reactor trip. This TTD f eature provides a variable tirne delay f or actuation of the reactor trip based on reactor power levels below 50 pareent RTP. The RCS hot- and cold-leg RTD temperatures are utilized to determine this power level by calculations performed with process circuitry. This logic also applies to the engineered safety feature that initiates auxiliary feedwater on low-low S/G 1evel. This function is designed to protect against the loss of reactor heat sink as well.
The functions described above utilize the RCS hot- and cold-leg RTD temperature measurements to perform their design functions. The channel calibration requirernents for these functions ensure that the RTDs have not experienced instrumentation drift or failures that would invalidate the assumptions used in the safety analysis. In addition, control and protection functions for low T ay signal for feedwater isolation, PermissiveP-12forsteamdump,hressurizerlevelcontrol,RCST measurementsforrodcontrolsystemautomaticoperation,andcalEu$ated value of RCS flow measurement uncertainty also utilize these RTD temperature measurementa. Enclosure 4 provides Westinghouse's safety evaluation that addresses the effect on each of these functions in relationship to the safety analysis for SQN. This evaluation provides the justification for the proposed TS changes needed (4s a result of the increased RCS RTD calibration uncertainty.
The conclusion of the Westinghouse safety evaluation is that with the proposed TS changes and the normalization of appropriate AT and T ay values, theRTDperformancewillbeacceptablewiththeoriginal'RTDfactorf calibration constants for the remainder of the current Unit 1 Cycle 6 fuel cycle. Additionally, the Westinghouse evaluation provides the basis for the conclusion that the RCS RTDs are accurate to 11.2 degrees F at the beginning of the Unit 1 Cycle 6 operation.' These combine to provide the technical basis for meeting the intent of the 18-month channel calibration for OTAT, OPAT, and RCS loop AT functions. This evaluation will be used in-lieu of an RCS RTD cross-calibration. TVA operates the SQN units within the constraints specified by Westinghouse for AT and T 3yg normalization and therefore ensures the validity of this evaluation
Environmental _ImpacLEvaluation The proposed change request revises the OTAT, OPAT, and P.CS loop AT allowable values and associated channel calibration requirements to ensure the operability of the associated reactor trip and engineered safety feature functions. This change does not involve an unreviewed environmental question because operation of SQN Unit 1 in accordance with this change would nott
- 1. Result in a significant increase in any adverse environmental impact '
previously evaluated in the Finni Environmental Statement (FES) as modified by the Staff's testimony to the Atomic Safety and Licensing floard, supplements to the FES, environinental impact appraisals, or decisions of the Atomic Safety and Licensing Board.
- 2. Result in a significant change in effluents or power levels.
- 3. Result in matters not previously reviewed in the licensing basis for SQN that may_have a significant environmental impact.
Enclosure 3 l'R01'0 SED TECilNICAL SPECIFICATION CilANGE SEQUOYAll NUCLEAR PIANT UNIT 1 DOCKET NO. 50-327 (TVA-SQN-TS-92-02)
DETERMINATION OF NO SIGNIFICANT !!AZARDS CONSIDERATION l
_ .- =-
e Sigulficant llaeards Evaluation TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear plant (SQN) in accordance with the proposed amendment will not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated.
As documented in the attached Westinghouse Electric Corporation evaluation (Enclosure 4), this TS change will provide different allowable values for overtemperature (OT) differential temperature (AT),
overpower (Op) AT, and reactor coolant system (RCS) loop AT and documents the technical evaluation used in lieu of the associated channel-calibration requirements.- These changes ensure that the accident analysis for SQN remains valid and that the associated surveillances remain in frequency. The impact on control and protection functions considering these changes is shown not to increase the probability of any accident because no accident initiator is affected.
With these changes, the consequences of an accident have been evaluated to ensure no increase in the radiological consequences would result.
Control and protection functions will continue to operate acceptably to maintain all the assumptions in the SQN safety analysis for the remainder of the Unit 1 Cycle 6 operation. The proposed TS changes will compensate for the increased calibration uncertainty of the RCS resistance temperature detectors (RTD).
- 2. Create the possibility of a new or different kind of accident from any previously analyzed.
These TS changes only affect protection functions that required additional conservatisms to support the SQN safety analysis because of the increase in RTD calibration uncertainty. All other effects resulting from the calibration uncertainty have been evaluated by Westinghouse and= verified not to impact the intended functions or operability of control or protection features. Accordingly, no new accident scenarios have been created by these changes to the TSs or tae calibration uncertainty.
- 3. Involve a significant reductico in a margin of safety.
The increase in the RTD calibration uncertainty did not adversely impact the safety-analysis limits or nominal trip setpoints of any protection function. To accommodate the increase in RTD uncertainty, the TS allowances for OTAT, OPAT. and RCS loop AT setpoints are reallocated.
The SQN safety analysis remains valid with these changes and does not involve a reduction in the margins of safety.
i Enclosure 4 l'R01'0 SED TECilNICAL Sl'ECIFICATION CllANGE SEQUOYAll NUCLEAR l'IANT UNIT 1 DOCEET NO. 50-327 (1VA-SQN-TS-92-02 )
WESTINGil0USE ELECTRIC CORI' ORATION SAFETY EVALUATION CllECK LIST (SECL)91-459. REVISION 2
FEB 19 ' M 14t43 FRCri LICDG1tG PfGE 002 SECL 91-459 Rev 2 Customer Reference No(s).
Westinghouse Reference No(s).-
WESTINGHOUSE SAFETY EVALUATION CHECX LIST
- 1) NUCLEARPLANT(S)__SE000VAHUNIT1
- 2) CHECK LIST APPLICABLE (SubjectofChange) TO:_ INCREASED RTD CALIBRATION UNCERTAINTY
_, 4 /-0. 5 ' F t o 4 /- 1. ? ' F
- 3) The written safety evaluation of the revised procedure design change or modification required and is required attached. by 10CFR50.59 has been prepare,d to the extent is incomplete for any reason, explain on Page 2.If a safety evaluation is not require Parts A and B of this safety Evaluation Check list are to be completed only on the basis of the safety evaluation performed.
CHECX LIST - PART A - 10CFR50.59(a)(1)
(3.1) Yes l No _ A change to the plant as described in the FSAR?
(3.2) Yes_
No 1 A change to procedures as described in the FSAR7 (3.3) Yes_
No1 A test or experiment not described in the FSAR7 (3.4) Yesl No_,__ A change to the plant technical specifications
- 4) CHECK LIST - PART B - 10CFR50.59(a)(2) (Justification for Part B answers must be included on Page 2.)
(4.1)
Yes__ No1 Will the probability of an accident previously (4.2) Yes evaluated in the FSAR be increased?
No 1 Will the consequences of an accident previously.
(4.3) evaluated in the FSAR be increased?
Yes,__ No 1 May the possibility of an accident which is diff"ent than any already evaluated in the FSAR be , wated?
(4.4)
Yes._, No1 Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
(4.5) Yes No1 Will the consequences of a malfunction of equip-ment important to safety previously evaluated in the FSAR be increased?
(4.6)
Yes___, Nel May the possibility of a malfunction 'of equipment important to safety different than any already (4.7) evaluated in the FSAR be created?
Yes.___ No1 Will the margin of safety as defined in the bases to any technical specification be reduced?
Page-I
FEB 19 '92 14843 FRCF1 LicD6Ito PAGE.003 i
SECL 91-459 Rev 2 i !
i If the answers to any of the above questions are unknown, indicate under (5) REMARKS and explain below.
If the answer to question (3.4) of Part A or any of the questions in Part C !
cannot be answered in the negative, based on the written safety evaluation, the Change review requires an application for license amendment as stated in 10CFR50.59(c) and must be suamitted to the NRC pursuant to 10CFR50.90. .
- 5) REMARKS:
The tnswers given in Sections 3 and 4, Parts A and b of the Safety Evaluation Checklist are based on the attached safety evaluation.
SEE ATTACHED SAFETY EVALVATION O
FOR FSAR UPDATE Section: Pages: Tables: Figures:
Reason for / Description of Change:
Changes to Technical Specifications: See attachment Table 2.2-1 Allowable values for RCS Loop Delta-T Equivalent to Power, Overtemperature Delta-T, and Overpower Delta-T Table 3.3-4 Allowable Value for RCS Loop Delta-T Equivalent to Power Approvals
- Prepared by: !- AM < Oate: 1 (6/
M rown /V/ 3' GroupManageb'a.'Stedis
/ -Date: /f '
f/
Page 2
-.---,--,-,-.,-,.-.,,,,.,,------,,,,,-----n ,,.n, .r n ,--~~.-w,-. , ,...-,,-,,,,c,--,,,,~n ,ne,,,m,,,, . , . - . n r-
i FEB 19 '92 14144 FROM LICEtGit0 N E.004 4
SECL 91-459 Rev 2 Page 1 of 7 Sequoyah Unit 1 Increase in RTD Calibration Uncertainty From +/-0.5'F to +/-l.2'F SAFETY EVALUATION 1.01hTR000CT10N Reference 1 provides documentation of the Westinghouse calculati>ns performed to determine the setpoints for the Sequoyah units witt the Eagle-21 protection system process racks installed. One of the assumptions made in the Westinghouse calculations is that the Hot Leg and Cold Leg Narrow Range RTDs are calibrated to within +/-0.5'F. This calibration is normally performed during plant heat-up after refueling via the cross-calibration process.
This safety evaluation addresses the impact on the safety analysis performed for Sequoyah Unit 1 Cycle 6 d increasing the RTO calibration uncertainty from +/-0.5'F to +/-1,2'F, The total RTD uncertainty included in the protection system set plus +/-0.7'F (drift over the cycle) pt for<nts is +/-l.2*F a total of + (calibration)
The Reference 1 analysis was performed assuming an RTD ca/-l.9'F.
libration uncertainty of +/-0.S'F and drift of +/-0.7'F for a total RTD uncertainty of
+/-l.2'F.
This safety evaluation provides the basis for the changes-recommended to the Technical Specifications provided in the attachment and supports a "No Significant Hazards
- determination pursuant to 10 CFR 50.92.
2.0 W " W NG BASIS The evaluation performed by Westinghouse is based on comparison of the results of calculations performed with the revised RTD calibration uncertainties to results and limits noted in Reference 1 and the Sequoyah '
Unit 1 Technical Specifications, The Total Allowance noted in Reference I was compared with the Channel Statistical Allowance values calculated for the revised RTD calibration uncertainties. Changes in the Allowable Values noted in the Technical Specifications were made where necessary.
The analyses given in Chapters 6 and 15 of the Sequoyah Unit 1 FSAR make explicit allowances for instrumentation errors for some of the reactor protection system setpoints. In addition, an allowance is made on the initial average reactor coolant system (RCS) temperature, pressure, and power as described in FSAR Section 15.1.2.2. The-following protection and-control functions are affected by increases in the Narrow Range RTD calibration accuracy: Overtemperature Delta-T (OTDT) reactor trip, Overpower Delta-T (0PDT) reactor trip, Vessel Delta-T Equivalent to Power (input to Steam Generator Level -Trip Time Delay), Low Tavg signal for
REB 19 '92 14115 FRCM LICDG!to F M E.fC*5 j SECL 91-459 Rev 2 Page 2 of 7 ,
feedwater isolation, Permissive P-12 for steam dump, Pressurizer Level for !
pressurizer level control, the rod control system, and the calculated ;
value of RCS flow measurement uncertainty.
3.0 [VALUATIOh 3.1 Instrumentation and Control Systems Evaluation While -/-0.7'F drift is utilized in this analysis, industry experience with RTDs show that little drift occurs. From the readings of the MMI of T-hot and T-cold, no anomalous indications are noted and a reasonable assurance that the Sequoyah Unit 1 RTDs have not experienced excessive drift during the time period can be made. Therefore, the +/-0.7'F assumed for drift-uncertainty can be considered conservative and would still be valid. The original factory calibration is +/-0.2*F and assures that +/-0.5'F could be verified through cross calibration.
Due to the vintage of the RTDs at TVA, and industry experience with other RTDs of the same vintage recent installation with RTDs calibrated in placeusingthecross-calibrationprocesshaveshownthatthefactory calibration uncertainty in addition to any installation errors have typically yielded RTDs within the +/-0.S*F range. As discussed above, it is reasonably expected that Sequoyah Unit 1 RTD performance is consistent with industry experience. Therefore, given that the factory calibration constants are Utilized, +/-0.5'F uncertainty along with the additional +/-0.7'F is judged to be sufficient to cover any potential deviation from average temperature for any RTD as an initial condition for beginning of Cycle 6 of Unit 1.
3.2 Setpoint Study Evaluation Westinghouse performed uncertainty calculations identical to those '
performed for Reference I with one change, the SCA (Sensor Calibration Accuracy) was modified to reflect the change from +/-0.S'F to
+/-l.2'F. After evaluation of the results of the calculations, the following conclusions were reached:
- 1) The rod control uncertainty increases from +/-4.5'F to i +/-4.6'F.
- 2) The Tavg --Low, low uncertainty increased by +/-0.2% of span, but because of the process rack change-out to Eagle instrumentation and the use of the New Steam Line Break Protection System, there is no protection system impact.
- 3) The uncertainty of the baseline RCS Flow Calorimetric remains unchanged since TVA is basing the Sequoyah RCS flow uncertainty on the use of an earlier cycle RCS flow calorimetric. .
- 4) The RCS flow indicated value on the control board or process computer is affected by a small increase in the Tcold RTD uncertainty and its subsequent atfect on cold leg density. The hffect of this increase is noted and accounted for within the round-off for the indicated flow I uncertainty.
- 5) The uncertainty for the loss of Flow Reactor Trip function is increased slightly due to the affect of the increased RTD error on Teold. Subsequent effects on cold leg density _are accounted for within the round-off of this function's- total uncertainty yalue.
.r A rts is *?2 ta w FRcn Ltcasuo Pr4E.006 SECL 91-459 Rev 2 Page 3 of 7
- 6) The Pressurizer Level Control uncertainty is determined to be +/-5.01%
of level span which is essentially the same as that assumod in the safety analysis as an initial condition (+/-5.00%).
- 7) Overtemperature Delta-T was evaluated and it was determined that the Allowable Value decreases from 1.9% of Delta-T span to 1.6% of Delta-T span, as noted on the attached Technical Specification page markups.
- 8) Overpower Delta-T was evaluated and it was determined that the Allowable Value decreases from 1.7% of Delta-T span to 1.6% of Delta-T span, as noted on the attached Technical Specification page markups.
- 9) Vessel Delta-T Equivalent to Power was evaluated and it was determined that it is appropriate to reduce the Allowable Value from 2.5% RTP
().7% span) to 2.4% RTP (1.6% span). This is noted on the attached Technical Specification page markups.
The probability and confidence level of the results of the protection function uncertainty calculations reported in Reference 1 have been previously identified to be accurate with a 95% probability at a 95%
confidence level. One assumption which supported this assertion was that a valid RTD cross calibration is performed which confirms an RTD uncertainty. While cross calibration data in this case can not be used to confirm an uncertainty, Westinghouse has no reason to believe that the prestr.t installation is not accurate with a 95% probability at a 95%
confidence level, given the discussion under Section 3.1, 3.3 Transient Analysis Evaluation The Chapter 6 and Chapter 15 safety analyses were performed assuming that, at steady state full power, the average RCS temperature was equal to the nominal value plus 5.5'F. This allowance it currently described in the FSAR as +/-4.0*F for rod controller accuracy and +/-1.5'F for accident evaluation. Based on the increase in the rod control uncertainty
(+/-4.5'T to +/-4.6'F) due to the increased RTD uncertainty, a portion of the accident evaluation margin can be allocated to maintain the same RCS temperature assumption in the safety analyses. Since the average temperature remains unchanged, the conclusions in the FSAR would remain valid when the effects of the estimated RTD errors are included for the Chapter 6 and Chapter 15 events which do not rely on the protection functions identified above.
To accommodate the increase in the RTD uncertainty, the Technical Specifications Total Allowances for the OTDT and OPDT reactor trip setpoints are reallocated. The reallocation ensures consistency with the FSAR Chapter 6 and 15 safety analysis assumptions for the OTDT and OPDT reactor protection functions. The modifications to the existing Technical Specifications which include the increased RTD calibration errors are provided in Attachment 1.
The Chapter 6 and Chapter 15 safety analyses do not take credit for the low Tavg feedwater isolation or Permissive P-12 steam dump functions.
Therefore, the impact of increased RTD calibration uncertainty on these functions does not affect the conclusions in the FSAR.
FEB 19 '92 14:47 FROM LICDGD4 PME.007 SECL 91-459 Rev 2 Page 4 of 7 Pressurizer Level uncertainty for pressurizer level control is increased to +/-5.01% due to the increased RTD calibration uncertainty. An increased pressurizer level can affect those analyses which use overfill of the pressurizer as an acceptance criterion. Overfill of the pressurizer is used as an acceptance criterion for the Loss of Normal feedwater (TSAR Section 15.2.8), Loss of Off-Site Power to the Station Auxiliaries (FSAR Section 15.2.9), and Rod Withdrawal at Power (FSAR Section 15.2.2) events. These analyses have been_ reviewed and it has been determined that the increased pressurizer level would not cause the pressurizer to overfill. Thus, the increased pressurizer level uncertainty is acceptable with respect to the non-LOCA accident analyses.
3.4 Operational Evaluation 3.4.' Delta-T Norm 112ation at Power One of the requirements for operability for the OTDT and OPDT protectici functions is the normalization to loop specific, indicated vessel Delta
- values. It has been recently noted tint with low leakage cores, the indicated vessel Delta-T values change in the non-conservative direction with increasing burnup (i.e., the indicated Delta-T decreases as the cycle progresses). This is attributed to the change in radial power 3 distribution with burnup. Renormalization of the protection channels should be performed if any indicated loop Delta-T is more than 1%
(0.6*F) smaller than the calibration value. Likewise, if the indicated loop Delta-T is more than 2% (1.2*F) larger than the calibration value, the channels should be renormalized. - The process of renormalization is compitcated by feedwater venturi fouling which has been observed in the past at Sequoyah. One of the mer' significant inputs to compensate for the magnitude of the fouling has been indicated vessel Delta-T. Since feedwater venturi fouling results in a conservative full power Delta-T, Westinghouse recommends the use of either, a value uT feedwater flow which is not compensated for the affects of venturi fouling, or feedwater flow which has been compensated by other means in order to renormalize the Delta-T used for the protection channels.
3.4.2 Tavg Input to OTDT and OPDT A secondary effect (which is not necessarily small in magnitude) is on the Tav9 input to-these protection functions. Generally, no effort is made to revise the T' and T" values to reflect the indicated, loop specific Tavg values at 100% RTP. However, Westinghouse makes the following recommendations in order to assure that the protection functions will respond appropriately, in light of the fact that the indicated Tavg has a possibility of being in error in the non-conservative direction and that the indicated Thot value will decrease with increasing burnup resulting in a change in indicated Tavg in the non-conservative direction.
it is recommended that T' and T" be modified to reflect the indicated, loop specific value for Tavg at 100% RTP at the beginning of the cycle. In addition, the T' and T" values should be modified to reflect loop specific, indicated values when the indicateo Delta-T values are
- FEB 19 * ?2 14
- 47 FROM LICEtG!tG P%E.009 SECL 91-459 Rev 2 Page 5 of 7 redetermined and the protection functions rescaled. This is a conservative position with respect to T' and is a restatement of the existing requirement for T' (see page 2-10 of the Sequoyah Unit 1 Technical Specifications). ,
When the plant is in automatic rod control with the reference Tavg set to the design nominal full power Tavg, the T' and T* values would not need modification. This is due to the control system maintaining the Tavg value to the same as tha' noted for T' and T' in the plant Technical Specifications.
When in manuti rod control, the operation of the plant should emulate the auto controller, i.e., maintain the plant about the same reference Tavg. Indicated temperature may vary above and below the reference Tavg as in auto control. A variance of +/-l.5'F about the reference Tavg would be consistent with the auto control system deadband.
3.5 Other Other safety related areas within the Westinghouse scope of supply have been reviewed and it was determined that none of these are affected by the increased RTD calibration uncertainty. These areas include:
mechanical components and systems integrity containment response radiological consequences LOCA and LOCA related transients including, large and small break LOCA LOCA hydraulic forces post-LOCA long term core cooling rod ejection mass releases hot leg switchover time to prevent boron precipitation steam generator tube rupture probabilistic rish assessment emergency operating procedures 4.0 ASSESSMENT OF UNREVIEWED SAFETY OVESTION The use of an increased RTO calibration uncertainty is evaluated below in accordance with the criteria of 10 CFR 50.59 as required to demonstrate that no unreviewed safety question is involved.
4.) The probability of an accident previously evaluated in the FSAR will not be increased.
The increase in the RTD calibration uncertainty does not adversely impact any control and protection functions. The setpoint study calculations setpoints wereperformed preserved. confirmed that in all cases the nominal trip Only a reallocation of Total Allowances was necessary to accommodate the increased calibration uncertainty. No initiators of any accidents are affected.
__ _ J
FEB 19 '92 14 4 FR31 LICDGitG FME 009 SECL 91-459 Rev 2 Page 6 of 7 4.2 The consequences of an accident previously evaluated in the FSAR will not be increased.
The setpoint study calculations showed that the nominal trip setpoints for the protection functions were unaffected by the increase in RTO calibration uncertainty, Also, the increased uncertainty on pressurizer level was shown to be acceptable with respect to the consequences of events which use overfill of the pressurizer as an acceptance criterion. Thus, system performance with respect to the control of radiological consequences is not adversely impacted.
4.3 The possibility of an accident which is different than any already evaluated in the FSAR has not been created.
No new limiting single failures are introduced due to the increase in the RTD calibration uncertainty. No previously incredible event is now made credible as a result of this change. All control and protection functions their intended functions, continue to be operable and capablo of performing 4.4 The probability of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
Neither system performance nor safety system functions are adversely impacted by the increase in the RTD calibration uncertainty. This activity has no affect on non-safety related equipment or functions which could in turn affect safety related equipment performance. This change does not affect the initiators of any event.
4.5 The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not be-increased.
The increase in the RTD calibration uncertainty does not adversely impact any contol or protection functions.
demonstrated that nominal trip setpoints are preserved.The setpoint calculations The affeet on pressurizer control can be accommodated in the analysis of those events which use overfill of the pressurizer as an acceptance criterion without any increase in radiological consequences. System performance is not compromised by this change.
4.6 The possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR will not be created.
No new failure modes for any equipment are created by the increase in the RTO calibration uncertainty. The affected control and protection functions continue to remain operable and capable of performing their intended functions.
4.7 The margin of safety as defined in the bases to any Technical Specifications will not be reduced.
The increase in the calibration uncertainty of the RTDs did not adversely impact the Safety Analysis Limit or Nominal Trip Setpoint of
r I
FEE 19 '?1 14:49 FRCr1 LICEtG!!G prGE . 010 SECL 91-459 Rev 2 Page 7 of 7 any protection function.
To accomodate the increase in the RTO uncertainty, the Technical Specifications Total Allowances for the OTDT and OPDT Power setpoints reactor trip setpoints and Vessel Delta-T Equivalent to are reallocated.
Chapters 6 and 15 continue to remain valid.The conclusions of the FSAR found in
5.0 CONCLUSION
An increased RTD calibration uncertainty of +/-l.2'F plus +/-0 7'F Unit 1 Chapter 6 and Chapter 15 accident analyses.for cycle a with the attached changes to the Technical Specifications, the increasedIt is RTD calibration question. uncertainty does not constitute-an unreviewed safety Westinghouse finds that continued operation with the original factory calibration constants is acceptable for the remainder of the current Unit 1 Cycle 6 fuel cycle.
6.0 REFERENCES
1.
WCAP-ll239 Rev 5, " Westinghouse Setpoint Methodology for Protectio Systems - Sequoyah Units 1 & 2 - Eagle-21 Version,' Maren 1991 7.0 ACKNOWLEDGEMENTS.
The fnllowing personnel contributed to this safety evaluation:
P..
B. Miller, J. F. Hermigos, W.H. Moomau, C. R. Tuley and J T
. . Doman.
FEB 7 '92 15 14 FROM L1 CENSING PAGE.011 4
- ATTACHMENT 1 TECH SPEC CHANGES SEQUOYAH 4
4 4
4 4
l m
g TABLE 2.2-1 (Continued) hw 8
4 REACTOR TRIP SYSTEM INSTRtMENTATION TRIP SETPOINTS -,
m FUNCTIONAL UNIT.
tw TRIP SETPOINT q 13. Steam Generator Water
_ ALLOWABLE VAttlES U g level-tow-Low
- a. ' RCS toop AT Er;uivalent to 7 ,
RCS Loop AT variable i Power 150% RTP input 1 50% RTP RCS Loop AT variable '"
I at45 ;
Coincident with input i trip setpoint
- RTP' Steam Generator Water 3
L I level - Low-Low (Adverse) > 15.0% instrument of narrow span range > 14.4% of narrow range r-and instrument span lni33 F.
Containment Pressure - EM i 0.5 psig m 2
or 5 0.6 psig #3'5 E
c>
Steam Generator Water > 10.7% of narrcw range rf Level -- Low-low (EAM) > 10.1% of narrow range
- instrument span jx135 with instrument span A time delay (T ) if one Steam Generator,is affected i T, (Note 5) 5 $ (1.01) T3 (Note 5) or 2 A tirse delay (T if two or more Steam GeneYa) i T,(Note tors are 5)
E+
affected 1 (1.01) T" (Note 5)
~
E
- Coincident with Steam Generator Water ~
gg Level - Low-low (Adverse) > 15.0% of narrow m
and instrument span range > 14.4% of narrow range instrument span {n155
{ Containment Pressure (EAH) 5 0.5 psig 3 c'
-30 C
or 1 0.6 psig gg43
"$ -^
Steam Generator Water > 10.7% of narrow range n Level -- Low-Low (EAM) ~ fnstrument rf ' > 10.1% of narrow range
~
instrument I"I53
'Q n
. m J
v> TABLE 2.2-1 (Continued) "
y ,
8 REACTOR TRIP SYSTEM INSTR M IfTATION TRIP SETPOINTS
-a 3.
x NOTATION (Continued) "
NOTE 2: (Continued)
E p 1 = .
3 Time constant utilized in the rate-lag controller for T
~
K =
avg, t3 = 10 secs. R145 o
A 6
- 0.M11 for T > T" and K6 = 0 for T 1 P T = c-as defined in Note 1 "
m T" = z Indicated T,,, at RATED THERMAL POWER (Calibration temperat'ure for AT Instrumentatten < $78.2*F) h 1
- 5 =
as defined in Note 1 f2(AI) =
0 for all Al _
8 NOTE 3: The chan '
perc atl's ATmaximum span. trip setpoint shell not exceed its computed trip point by more than NOTE 4 The ch K per ont AT I's maximum span. trip setpoint shall not exceed its computed trip point by more than at45
~8 /. 6 , ,
W
-3 p -
a T
D
-[ O
'N O e
Y--
~
a
. . A
, TARLE 3.3-4 (Continued) /
Q -s g ENGINEERED SAFETY FEATURE ACTPATION SYSTEM INSTRIMENTATION TRIP SETPOINIS ,
FUNCTIDAAL UNIT TRIP SETPOINT .
ALLOWAELE VALUES G
g
- 6. AUXILIARY FEEDWATER ;
o,
[ a. Manual Not Applicable Not Applicable
- b. Automatic Actuation Logic Not Applicable Not ArpIfcable A S
- c. Main Steam Generator ,_
Water Level--Low-Low -
i 1. RCS toop AT Equivalent RCS Loop AT variable RCS Loop AT variable 5 5 RTP to Power 10% input $ 0%
5 RTP input < trb' setpoint 4145 g
+J M% RIP a
+ 2.Y %
R**
Coincident with Steam >15.0% of narrow range >14.4% of--darrow lnI55 Generator Water Laval-- Instrumer.t span range instrument. span
'? ' Low-Low (Adverse)
O and R145 l
Containment Pressure-EAM 10 5 psig 10 6 psig ,
l or Steam Generator Water >10.7% of narrow range . >10.1% of parrow Tastrument span Tnstrument span lat55 Level-tow-tow (EAM)
{
a with A time delay (TS) if one 1T3 (Note 5, Table 2.2-1) 1 (1.01) T3 (Note 5, k Steam Generator is affected Table 2.2-1)
S or g A time delay (T,) If two 1 T,(Note 5 Table 2.2-1) $ (1.01) T ,'(Note 5, or more Steam Generators Table 2.2-1) are affected Of .
C_. 3 c-- "?g a r- ."
w5
~ '
. -r -
Yo to
=
a .