ML20087D380
| ML20087D380 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/07/1995 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20087D378 | List: |
| References | |
| NUDOCS 9508100304 | |
| Download: ML20087D380 (40) | |
Text
...
'eiS - m mg g TABLE 2.2-1 (Continued) g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS CO
@y f FUNCTIONAL UNIT 7:a TRIP SETPOINT ALLOWABLE-VALUES om E 21. Turbine Impulse Chamber Pressure - < 10% Turbine Impulse R145 8
$$ Q (P-13) Input to Low Power Reactor Trips < 12.4% Turbine In mlse l Pressure Equivalent Pressure Equivalent
.y88 s Block P-7 em"O 22. Power Range Neutron Flux - (P-8) Low a 5 35% of RATED 5 37.4% of RATED Reactor Coolant Loop Flow, and Reactor THERMAL POWER R145l Trip '
THERMAL POWER
- 23. Power Range Neutron Flux - (P-10) - R145
> 10% of RATED > 7.6% of RATED Enable Block of Source, Intermediate, THERMAL POWER THERMAL POWER and Power Range (10w setpoint) Reactor Trips Reactor Trip P-4
] 24. Not Applicable Not Applicable
- 25. Power Range Neutron flux - (P-9) - < 50% of RATED E Blocks Reactor Trip for Turbine < 52.4% of RATED THERNAL POWER THERMAL POWER Trip Below 50% Rated Power NOTATION NOTE 1:
- 5 Overtemperature AT (1
- T4 S) 5 AT, {K y -K 1+ISS 2 (1 + r2 )[T-T'] + K3 (P-P') - (1(AI)}
1 '
Where: I * *4b = Lead-lag compensator on measured AT
- $ 1+T 5 >
[$
cc P 4'r5
= Time constants utilized in the lead-lag controller for AT 1 2)5( secs r 73sec.
4 4
$g* AT, = Indicated AT at RATED THERMAL. POWER
. ,:h, g K $ 1.15 1
5 g
K 2 / 0.011 3
- g. -
WA'
-_.___._.____.___.___._______-_.-_____.____.__.____.____s ________m__ _ _ _ _ _- _ _ _ _ _
i :;
y, .
- 18 TABLE 2.2-1 (Continued) 8 REACTOR TRIP SYSTEM INSTRIMENTATION TRIP SETPOINTS Y
x e NOTATION (Continued)
@ NOTE 1: (Continued) '
"s g 1 + r,5
= -
1+tS2 The function generated by the lead-lag controller for T,,g @namic compensation R145-r,&r =
y 2 Time n 1
2g4 secs. ts utilized in the lead-lag controller for T,yg, ty 733 secs.,
T = Average temperature *F R145-T' <
_ 578.2*F (Nominal.T,yg at WED THM NR)
'y K = 0.00055 a> 3 P = Pressurizer pressure, psig P' =
2235 psig (Nominal RCS operating pressure)
S = Laplace transform operator (sec-1) and fy(AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for qt ' 9 h between - 29 percent and + 5 perceiit fy (AI) = 0 (where qt and qb gy are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, 2 i. g and qt*9b is total THERMAL POWER in percent of RATED THERMAL POWER). '
-< a
~
Q U cs E co
c . _ ,
m TABLE 2.2-1 (Continued) .
E 8 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS slE I
NOTATION (Continued)
E NOTE 1: (Continued)
Z g (ii) for each percent that the magnitude of (qt-- 9 b) exceeds -29 percent, the AT trip set-point shall be automatically reduced by 1.50 percent of its value at RATED THERMAL POWER.
R23 (iii) for each percent that the magnitude of (qt - 9 )b exceeds +5 percent, the AT trip set-point shall be automatically reduced by 0.86 percent of its value at RATED THERMAL POWER.
5 NOTE 2: Overpower AT (1 + T45) 5 AT, {K4 -K 3 ) T - K6 (T-T") - f2fAI)I 1+T55 5 (1 + T3 5
'?
Where:
1+T54 = as defined in Note 1 1+155 T4,T5
= as defined in Note 1 AT"
= as defined in Note 1 R118-h K 4 5 1.087 K 7 0.02/*F for increasing average temperature and 0 for decreasing average 5
,_. g h temperature 8z o? r5 3 = R145 N I+TS3 The function generated by the rate-lag controller for T,yg dynamic
- compensation U
b
_.-_.__m_ _ _ __m' - - _ ..- __-_e-
m ,
37..
m
, _ . . -3
..~- t
.'4:. .
+
TABLE 2.2-1-(Continued)
)'8 REACTOR TRIP SYSTEM INSTRtMENTATION TRIP SETPOINTS y NOTATION (Continued) 1.
NOTE 2: (Continued)
E h RW, Z t 3
=
Time constant utilized in the rate-lag controller for T,yg 1 # 10 secs.
3 K
6 0.0011 for T > T* an K 6 0 for 7 5 T" T = as defined in Note 1 .
T" = Indicated T at RATED THERMAL POWER (Calibration temperature for AT instrumantation, 1 578.2*F)
S = ~ as defined in Note 1 f2 (AI) = 0 for all AI E
~
NOTE 3: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 1.9 percent AT span.
NOTE 4: The channel's maximum trip setpoint shall not exceed its computed trip point by more than' R145 gg 1.7' percent AT span.
O g'
. .g G
W G
o 4
4 1 m
. _ _ _ _ _ _ _ ______1___________. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ ,_ . _ _ _ _ _ . _ _ ._. - _ _ . __ ,
- c. 3.x m TABLE 2.2-1 (Continued) -
18 g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS x
e NOTATION 5
NOTE 1: Overtemperature AT (1
- 4I 5) $ AT, {K y -K g (1
- I lb)[T - T'] + K3 (P-P') - f y(AI)} R132 1+155 1+152 1+TS4 .
where: = Lag com ensator on measured AT 1 * '53
- 2: S T4,r5 = tan s utilized in the lead-lag controller for AT,14# W secs, T
5{3 secs AT = Indicated AT at RATED THERMAL POWER i
ff e
R21 K 5 1.15 1
Kg
{0.011 1+tb1 4
=
7 73 The function generated by the lead-lag controlhr for T,yg dynamic compensation R132 E
r,&T = T' y
y 2 tp g st nts utilized in the lead-lag controller for T,yg, t y 7 33 secs.,
o
_ 4 secs.
g" T = Average temperature *F '
R132
$3 2 T'
5 578.2*F (Nominal T,yg at RATED THERMAL POWER) f] P K 3
= 0.00055 U3 - P = Pressurizer pressure, psig to g
'3 2 P' = 2235 psig (Nominal RCS operating pressure)
C N
. .m ,
as .-
5 2
I4 TABLE 2.2-1 (Continued) *
$ REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5x-e NOTATION (Continued)
E NOTE 1: (Continrad)
S = Laplace transform operator. sec -1 and fy(AI) is a function of the indicated difference between top and bottom detectors of the power range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
, (1) for qt 9b between - 29 percent and + 5 percentyf (AI) = 0 (where gt and q b
$ are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, R21 and qt
- 9b is total THERMAL POWER in percent of RATED THERMAL POWER).
(ii) for each percent that the magnitude of (qt 9b ) exceeds -29 percent, the AT trip set-
-point shall be automatically reduced by 1.50 percent of !ts value at RATED THERMAL POWER.
(iii) for each percent that the magnitude of (qt ~ 9 b) exceeds +5 percent, the AT trip set-point shall be automatically reduced by 0.86 percent of its value at RATED THERMAL POWER.
l N
5 g NOTE 2: Overpower AT (1
- 4I 5) < AT, {K4 -K 3 ) T -K6 [T - T"] - f2(AI)I g a 1+TSS 5 (1
- T3 5 1+tS4 I '
where: = as defined in Note 1 1+rS5 r1 l3 .
g
- )
~. y
~
w.
m- . TABLE 2.2-1 (Continued) .
E g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5
x NOTATION (Continued)
E Q
NOTE 2: (Continued) 7 7
= as defined in Note 1 R132
= as defined in Note 1 AT, K < 1.087 (104 4 -
K 5 / 0.02/*F for increasin average temperature and 0 for decreasing average h lemperature
- f
'? T3 3
0 1+r5= R132-3 Thecompensation function generated by the rate-lag controller for T,yg dynamic-T 3
=
Time constant utilized in the rate-lag controller for T T3 2 10 secs.
~
avg K x 0.0011 for T > T" and K,7 0 for T < T" 6
g _ , -_ 21 g T. = as' defined in Note 1 o.
C3 5 T" = Indicated Tavg at RATED THERMAL POWER (Calibration temperature for cm g '
-t , AT instrumentation, < 578.2*F)
- r.3 P
- g. ,. , S = as defined in Note 1 y f 2(AI) = 0 for all'AI 7 . .a ,2 U
~ .,
- .- 1:.--_
mz -
,t-ENCLOSURE 2 ;
PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE l
SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-95-11, REVISION 1)
DESCRIPTION AND JUSTIFICATION FOR REVISION OF OVERTEMPERATURE AND OVERPOWER DELTA TEMPERATURE EQUATION TIME CONSTANTS ,
3 I
Descriotion of Chance TVA proposes to modify the Sequoyah Nuclear Plant (SON) Units 1 and 2 technical specifications (TSs) to revise the overtemperature delta temperature (OTAT) and overpower delta temperature (OPAT) equation constant numerical value r4 in TS Table 2.2-1. The r4 value will be changed from 12 seconds to 5 seconds and the rs value will remain at a value of 3 seconds. An additional enhancement has been incorporated to change the equality signs for the r and K constant numerical values to either a less than or equal to or a greater than or equal to function as applicable.
Reason for Chance SON has experienced OPAT turbine runback alarms on individual channels resulting in partial runback signals. During functional testing at power, as required by TSs, these occurrences could result in turbine runbacks or reactor trips because the tested channelis placed in the trip condition completing the required logic for actuation. The r constant numerical values have been reanalyzed to provide additional margin to these setpoints and minimize the potential for turbine runback and reactor trip signals. Incorporation of the new r numerical value will provide a reasonable margin to the turbine runback and reactor trip setpoints and minimize the challenges to safeguard function actuation. The revision of the equality signs for the equation constants is consistent with standard TS (WUREG-1431). The use of greater than or equal to or less than or equal to functions for these constant numerical values is more appropriate and will provide the flexibility to set these values within the limits utilized in the safety analysis.
Justification for Chanaes The OTAT trip provides core protection to prevent departure from nucleate boiling for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to transit, thermowell, and resistance temperature device (RTD) response time delays i from the core to the temperature detectors, and pressure is within the range between the high and low pressure reactor trips. This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature and dynamic compensation for transport, thermowell, and RTD response time delays from the core to the RTD output indication. With normal axial power distribution, this reactor trip limit is always below the core safety limit. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip setpoint is automatically reduced.
The OPAT reactor trip provides assurance of fuelintegrity, limits the required range for OTAT protection, and provides a backup to the high neutron flux trip.
The setpoint includes corrections for changes in density and heat capacity of water with temperature, dynamic compensation for transport, thermowell, and RTD response time delays from the core to the RTD output indication. The OPAT provides protection to mitigate the consequences of various size steam breaks.
Westinghouse Electric Corporation has evaluated SON's setpoints for OTAT and OPAT and has verified that sufficient margin exists in the SON accident analyses to support relaxation of the lead / lag dynamic compensation. This evaluation addresses the Final Safety Analysis Report (FSAR) Chapter 6 and 15 accidents that are affected by a revised lead / lag dynamic compensation.
These evaluated accidents only include the non-loss-of-coolant accidents (LOCA) because the LOCA and containment integrity accidents are not affected by the revised compensations. The results of the evaluation, which are included in Er closure 4, indicate that the proposed revision of the r4 constant numerical value is acceptable in that all acceptance criteria of the current safety analyses described in the FSAR continue to be met.
The revision of the equality signs for the constant numerical values of the OTAT and OPAT equations does not change these functions. This change will clarify that the adjustment of the related setpoints for these values can be set more conservatively than assumed in the analysis. This change is consistent with the provisions in NUREG-1431.
Environmental Imoact Evaluation The proposed change does not involve an unreviewed environmental question because operation of SON Units 1 and 2 in accordance with this change would not:
l
- 1. Result in a significant increase in any adverse environmentalimpact previously evaluated in the Final Environmental Statement (FES) as modified by NRC's testimony.to the Atomic Safety and Licensing Board, supplements to the FES, environmental 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 SON that may have a significant environmental impact.
- wm._.--
in.
s 1
ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-95-11, REVISION 1)
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION
--,_ y
6 l
Significant Hazards Evaluation ,
l 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 (SON) in accordance with the proposed amendment will not:
- 1. Involve a significant increase in the probability or consequelces of an accident previously evaluated.
The revision of the r4 constant numerical value in the overtemperature delta temperature (OTAT) and overpower delta temperature (OPAT) equations have been analyzed by the enclosed Westinghouse Electric Corporation evaluation and have been found to have sufficient margin for the proposed change. This evaluation shows that the proposed changes are bounded by the existing analysis for Chapter 6 and 15 accidents. The setpoint change will continue to meet the applicable safety analysis acceptance criteria for the transients evaluated. The offsite dose rates for postulated accidents have not exceeded the values stated in the Updf ted Final Safety Analysis Report as a result of this change. The clarification of the equality signs for the constant numerical values does not change plant or accident mitigation functions. Therefore, the proposed changes will not increase the consequences of an accident.
This change affects the OTAT and OPAT functions that are designed to mitigate the consequences of an accident and are not considered to be an accident initiating source. Therefore, the probability of an accident is not increased by the proposed change.
- 2. Create the possibility of a new or different kind of accident from any previously analyzed.
The revision of lead / lag dynamic compensation for the OTAT and OPAT functions do not impact accident initiators because these functions are used for accident mitigation and are not postulated as a source.
Therefore, the possibilit/ of a new or different kind of accident is not created by the proposr,d revision.
- 3. Involve a significant reduction in a margin of safety.
The proposed revision to the lead / lag compensation for the OTAT and OPAT functions does not invalidate the conclusions in the safety analysis.
Margins provided for in the safety analysis are maintained with the proposed changes such that no reduction in the margin of safety is involved.
e' ENCLOSURE 4 PROPOSED TECHNICAL SPEdlFICATION CHANGE' SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 s
DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-95-11, REVISION 1)
WESTINGHOUSE ELECTRIC CORPORATION EVALUATION OF REVISED TIME CONSTANTS-i
, . . .. .. .- ~ . . .n,. . -.. -
4' ' q ,
SECL-95-063, Revision 0 Customer Reference N' o(s). i N/A Westinghouse Reference No(s).
N/A WESTINGHOUSE i SAFETY EVALUATION CHECF. LIST -
- 1) - NUCLEAR PLANT (S)- Sequoyah Units 1 & 2
- 2) CHECK LIST APPLICABLE TO: OPDT / OTDT Ooeratina Martin Imorovement Via Revised Lead / Lar Comnensation Grom 12/3 To 5/3)
- 3) The written safety evaluation of the revised procedure, design change or modificatten required' by i 10CFR50.59 has been prepared to the extent required and is attached. If a safety evah.:!vc is not -
required or is incomplete for any reason, explain on Page 2. Parts A and B of this Safety . .
1 Evaluation Check List are to be completed only on the basis of the safety evaluation performed.-
CHECK LIST - PART A E
i 3.1) Yes_X_ No_ A change to the plant as described in the FSAR?
3.2) Yes_ No_X. A change to procedures as described in the FSAR?
3.3) Yes_ No_X_ A test or experiment not described in the FSAR?
3.4) Yes_X No_. A change to the plant Technical Specifications (Appendix A to the Operating License)?
l
- 4) - CHECK LIST - PART B (Justification for Part B answers must be included on page 2.) -
, 4.1) Yes._ No_X. -Will the probability of an accident previously evaluated in the FSAR be .
. increased? '
4.2)- Yes_ - No.X_ Will the consequence,: of an accident previously evaluated in the FSAR be increased?
4.3) Yes_ No_X. -May the possibility of an accident which is different than any already evaluated in the FFAR be created?
4.4) Yes_ No_X_ .Will the probabiliy of a malfunction of equipment important to safety ' ,
previously evaluated in the FSAR be increased? '
4.5) Yes__. No_X_ Will the consequeaces of a mai' unction of equipment important to safety--
previously evaluated in the FSAR be increased?
4.6) Yes_ No_X May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?
4.7) Yes_ No_X_ Will the margin of safety as defined in the bases to any Technical Specification be reduced?
4
- j. i.vr uma Page 1 of 10 n ---v v - ~ - - - . , ~. --:,.-,,m ,e,, nv . m rw, , w r-se v - q
I l
l SECL-94-063 -1 Revision 0 If the answers to any of the above questions are unknown, indicate under 5) REMARKS and explain below.
If the answer to any of the above questions in Part (3.4) or Part B cannot be answered in the negative, the change review requires an application for license amendment in accordance with 10CFR50.59(c) and submitted to the NRC pursuant to 10CFR50.90.
- 5) REMARKS:
The answers given in Section 3, Part A, and Section 4, Part B, of the Safety Evaluation Checklist, are based on the attached Safety Evalua. tion.
1 Reference document (s):
l N/A I 1
FOR FSAR UPDATE See separately transmitted UFSAR mark-ups [For 5/3 Lead / Lag: Section 15.2.2 pages 15.2-5 through 15.2-9 Table 15.21 (Sheet 1), and Figures 15.2.2-1 through 15.2.2-7)
SLB w/RWAP mark-ups (Ref. WCAP-12504)
See attached Tech Spec mark-ups [ Table 2.2-1 (Page 2-7 and 2-8, Unit 1; Page 2-9, Unit 2)]
See attached PL&S mark-ups.[Pages 14 and 15 of 102]
.I SIGNATURES Prepared by: L.V. Tom EMM Date: )
- I b * -
b Reviewed by: T. L Kitchen -
Date: 7k9 /iI v <
f s mm2 a Page 2 of 10
.r l
l
- SECL-94-063 Revision 0
SUMMARY
The overall purpose of this safety evaluation is to increase the Overtemperature Delta-T and the Overpower Delta-T (OPDT / OTDT) operating margin to trip in order to increase the turbine runback margin (associated with OPDT / OTDT) by relaxing the lead / lag dynamic compensation.
His' safety evaluation addresses revising the OPDT and OTDT lead / lag dynamic compensation from 12/3 to 5/3 and concludes that this revision does not represent an unreviewed safety question pursuant to 10CFR50.59,(a), (2), criteria, provides the basis for the attached marked up Technical Specifications, the attached marked-up PL&S document and the separately transmitted marked-up FSAR revisions, and supports a no significant hazards consideration pursuant to 10CFR50.92, (c), criteria.
]
Decreasing the lead / lag compensation from 12/3 to 5/3 is predicted to increase the OPDT /OTDT margin to trip from 1.0% to 3.6%. The turbine runback margin would increase from 1.0% to 2.6%.
This safety evaluation addresses the FSAR Chapters 6 & 15 accidents that are affected by a revised i lead / lag dynamic compensation, which are the non-LOCA accidents described herein. The LOCA related '
accidents ( large and small break LOCA, reactor vessel and loop LOCA blowdown forces, post-LOCA long term core cooling subcriticality, and hot leg switchover to prevent boron precipitation), and the Containment integrity accidents are not affected by revised lead / lag compensations. The steam generator tube rupture (SGTR) accident is not adversely affected by the revised lead / lag compensation. The i
UFSAR SGTR analysis was performed to evaluate the radiological consequences resulting from the double ended rupture of a single tube in one steam generator. The major factors that affect the extent of
' radioactive release and the resultant offsite radiation doses for an SGTR are the amount of fuel defects (level of reactor coolant contamination), the primary-to-secondary mass transfer through the ruptured ,
tube, and the steam released from the faulted steam generator to the atmosphere. He time of reactor trip l is important for offsite doses since the amount of radioactivity discharged to the atmosphere is directly impacted by the duration of steam release via the steam generator safety and/or power operated relief 1 valves (coincident loss of offsite power or failure of the condenser dump system assumption). Tube rupture recovery actions are assumed to be completed within 30 minutes of accident initiation, {
4 consequently, an earlier trip is conservative per current SGTR methods. For an SGTR event, reactor trip cr.n occur either on OPDT/OTDT or low pressurizer pressure. The change in lead / lag compensation j would effect the OPDT and OTDT setpoint algorithm and result in an increase in the time of reactor trip.
4 However, this will not adversely effect the radiological consequences of a SGTR accident as the SGTR accident analysis trips upon low pressurizer pressure. He change in OPDT/OTDT lead lag compensation will not affect the pressurizer low pressure trip setpoint. The doses from the SGTR will remain well !
within the limits as defined in 10CFR100 (25 rem whole body and 300 rem thyroid). Additionally, in order to predict the margin to trip improvement for the lead / lag relaxation, an OPDT/OTDT margin parametric study was performed.
He curves showing the results of the parametric study for the margin to trip are included in Attachment 1, the PL&S mark-ups are included in Attachment 2, and the Technical Specification mark-ups are included in Attachment 3.
l ,
LYT 95M2.dA Page 3 of 10
\ i 1
SECL-94-063 Revision 0 NON-LOCA BACKGROUND Overpower AT turbine runback alarms are intermittently occurring on Sequoy margin to the overpower AT turbine runback alarm. Unit I and Unit 2 The change is a decrease in the ratio of the time constants in the lead-lag compensa!
coolant from 12/3 tosystem 5/3. (RCS) AT. The time constants in the lead lag compensator on RC LICENSING BASIS ne lead lag compensated RCS AT signals and lag compensated hot and cold leg t modeled in SQN licensing-basis safety analyses which rely on the reactor protection sy OTAT reactor trip functions for primary protection. Table I lists the SQN safety the OTAT or OPAT reactor trip functions for primary protection.
Table 1 SQN Safety Analyses which Trip on OTAT or OPAT Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (UFSAR 15.2.2 Loss of External Electrical Load and/or Turbine Trip (UFSAR 15.2.7)
Uncontrolled Boron Dilution (UFSAR 15,2,4)
Accidental Depressurization of the Reactor Coolant System (UFSAR 15.2.12)
Steamline Break with Coincident Rod Withdrawal at Power (WCAP-12504)
Steamline Bred Mass / Energy Outside Containment (WCAP-10961)
ANALYSES / EVALUATIONS IMPACT OF THE TIME CONSTANT CHANGES TO THE RCS AT LEAD-LAG Table 2 presents the analysis values of the RCS AT lead-lag compensator time consta licensing basis safety analyses, wr mm a Page 4 of 10
SECL-94-063 Revision 0 Table 2 12ad-1ag Time Constants in the RCS AT L ad-Lag Compensator for the SQN Safety Analyses which Trip on OTAT or OPAT Lead _ Lag Eycat (seconds) (seconds)
Uncontrolled Rod Cluster Control Assembly 12 3 Bank Withdrawal at Power Loss of External Electrical Load and/or Turbine Trip 0 0 Uncontrolled Boron Dilution 12 3 Accidental Depressurization of the Reactor 0 0 Coolant System Steamline Break with Coincident Rod Withdrawal 12 3 at Power Steamline Break Mass / Energy Outside 0 0 Containment As shown in Table 2, three of the SQN safety analyses, Loss of External Electrical Load and/or Turbine Trip, Accidental Depressurization of the RCS, and Steamline Break Mass / Energy Outside Containment model a 0-0 lead-lag. Use of the 0-0 lead-lag compensation is more limiting than the 5-3 lead-lag compensation and provides a bounding analysis model. This is because the 0-0 compensation results in a slower reactor trip on OTAT and OPAT. Since these three events are analyzed with a conservative lead- i lag model, there is no adverse impact on the analysis results with the implementation of the 5-3 lead-lag.
Consequently, the UFSAR conclusions for the Loss of External Electrical Load and/or Turbine Trip, Accidental Depressurization of the Reactor Coolant System, and Steamline Break Mass / Energy Outside Containment events continue to remain valid.
The remaining Table 2 analyses, Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Uncontrolled Baron Dilution, and Steamline Break with Coincident Rod Withdrawal at Power, assume a 12-3 lead-lag. Use of the 12-3 lead-lag compensation is less limiting than the 5-3 lead-lag compensation and does not provide a bounding analysis model. This is because the 5-3 lead lag compensation results in a slower reactor trip on OTAT and OPAT. 'Ihe three events which are adversely impacted by the change in lead-lag time constants are evaluated as follows.
Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power The Rod Withdrawal at Power event is analyzed with a variety of reactivity insertion rates at 10%,60%,
and 100% of rated thermal power. Depending on the case analyzed, either the High Neutron Flux or OTAT reactor trip occurs. This event was conservatively reanalyzed with 0-0 lead-lag time constant values. Based on the results of the analysis, all applicable safety analysis criteria continue to be satisfied and the UFSAR conclusions remain valid.
i.vr mo m Page 5 of 10
SECL-94-063 Revision 0 Uncontrolled Boron Dilution The Baron Dilution event is analyzed to identify the amount of time available for operator or automatic mitigation of an inadvertent boron dilution prior to complete loss of shutdown margin. This transient is required to be considered for Sequoyah for refueling (Mode 6), startup (Mode 2) and power operation (Mode 1). The subcritical modes of operation: hot standby (Mode 3), hot shutdown (Mode 4), and cold shutdown (Mode 5) are not affected by the lead-lag time constant because Sequoyah adheres to the Westinghouse interim operating procedures (Reference 5) which do not rely on the OTAT or OPAT reactor trips for event mitigation. Following these administrative procedures ensures that 15 minutes of operator action tirr.e is available for mitigating an uncontrolled boron dilution event while in Mode 4 or 5 and operating on the residual heat removal (RHR) system. Operating in Mode 3 with at least one reactor coolant pump in operation is less limiting than operating in Mode 4 and 5 on RHR due to increased mixing ud decreased localized dilution. The Mode 6 analysis is unaffected by the lead-lag time constant charge decause it is administratively precluded by the limitations of the Technical Specification 3/4.9.1 The Mode 2 analysis is unaffected as it relies on a source range reactor trip to initiate operator action to mitigate a boron dilution transient. The source range reactor trip setpoint is not affected by the OPDT/OTDT lead / lag constant revision.
The Mode 1 event is analyzed in two separate cases which assume either that the control rods are in manual mode of operation or automatic. If the control rods are in automatic, the operator would be alerted to the occurrence of a boron dilution by the rod insertion limit alarms. If the rods are in manual, the first indication may be the OTAT reactor trip. The Mode I rods in manual analysis is impacted by the lead-lag time constant change because the time of reactor trip on OTAT is taken from the Uncontrolled RCCA Bank Withdrawal at Power (RWAP). The time of reactor trip is taken from the RWAP case which has a reactivity insertion rate equal to or less than that calculated for the boron i dilution event and then subtracted from the amount of time available between start of the event and loss of i shutdown margin. Acceptable results were obtained when the RWAP results from the previously discussed reanalysis were used to calculate the amount of time for operator action. Consistent with the SQN UFS AR, there are over 40 minutes available between the start of the event and the complete loss of j shutdown margin. Based on the results of the analysis, all applicable safety analysis criteria continue to l be satisfied and the UFSAR results and conclusions remain valid. l l
Steamline Break with Coincident Rod Withdrawal at Power !
In September of 1979, IE Information Notice 79-22 was issued by the NRC addressing a potential unreviewed safety question resulting from Control and Protection Systems interaction. One of the postulated scenarios identified was the operation of the rod control system following an inside containment steamline rupture (refer to Reference 4).
This analysis is simulated by modeling a steamline rupture and coincident withdrawal of control bank D at full power conditions. A spectrum of steamline break sizes was analyzed to determine the limiting condition. The following reactor trip functions may actuate during this postulated steamline rupture with a consequential rod withdrawal transient depending on the break size:
- a. OPAT
- b. A reactor trip is generated subsequent to safety injection system and steamline isolation actuation caused by low steamline pressure.
umm a Page 6 of 10
SECL-94-063 Revision 0 This event was reanalyzed with the change to the 5-3 lead-lag time constant values. Based on the results of the analysis, all applicable safety analysis criteria continue to be satisfied and the UFSAR conclusions remain valid.
ASSESSMENT OF UNREVIEWED SAFETY QLTSTION Operation of Sequoyah Nuclear Plant Units 1 and 2 with the revised RCS AT lead-lag compensator time constants do not constitute an unreviewed safety question. This conclusion is based on the responses to the seven questions below.
4.1 Will the probability of an accident previously evaluated in the FSAR be increased?
No. As addressed in this safety evaluation, all transients affected by the proposed modifications
, were reanalyzed or evaluated and found to adhere to the safety analysis acceptance criteria. He l proposed modification does not adversely affect the integrity of the RCS or Main Steam System I
pressure boundary. He operation of the plant with the revised RCS AT lead-lag compensator time constants do not impose any new performance requirements on the RCS. The probability of such accidents occurring remains unaffected.
4.2 Will the consequences of an accident previously evaluated in the FSAR be increased?
No. Per the discussion presented in the Analyses / Evaluations section, all the applicable acceptance criteria are still met for the transients evaluated and for the events reanalyzed. Additionally, no new limiting single failure is introduced by the proposed change. Therefore, there is no potential for an increase in the consequences of an accident previously evaluat a Nt he FSAR. The revised RCS AT lead-lag compensator time constants do not result in a chal'ent tc 'he fission product boundaries, i.e., fuel cladding, pressure vessel and containment. Offsa does for all events do not exceed the values reported in the FSAR.
4.3 May the possibility of an accident which is different than any already evaluated in the FSAR be created?
No. Revising the RCS AT lead-lag compensator time constants do not introduce a new accident i initiator mechanism. Hus, no new accident will be created.
4.4 Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
No. Revised RCS AT lead-lag compensator time constants will not adversely affect operation of the Reactor Protection System, any of the protection setpoints, or any other device required for accident mitigation. Operation with the revised RCS AT lead-lag compensator time constants will not affect the probability of safety related equipment malfunctions currently evaluated in the FSAR.
4.5 Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
l No. As discussed in the responses to questions 2 and 4, there is no possibility of increasing consequences of a malfunction of equipment for the revised RCS AT lead-lag compensator time Constants.
mmm Page 7 of 10
SECL-94-063 Revision 0 4.6 May the possibility of a malfunction of equipment important to safety different than that evaluated in the FSAR be created?
No. As discussed in question 4, revised RCS AT lead-lag compensator time constants will not impact any other equipment important to safety. Operation of the plant with the revised RCS AT lead-lag compensator time constants do not introduce any new failure modes for any equipment which are credible and have not been previously considered in the FSAR. No new performance requirements are imposed on any system or component such that any design criterion is exceeded.
4.7 Will the margin of safety as defined in the bases to any technical specification be reduced?
No. As discussed in the safety evaluation, the proposed RCS AT lead-lag compensator time constants will not invalidate any of the conclusions presented in the UFSAR accident analyses.
Therefore, the margin of safety, as defined in the bases to the Technical Specifications will not be decreased.
CONCLUSION Based on the preceding evaluation, it is concluded that the acceptance criteria for the SQN Unit 1 and Unit 2 licensing-basis safety analyses continue to be satisfied and the UFSAR conclusions remain valid for the revised RCS AT lead-lag compensator time constants.
REFERENCES
- 1. Tennessee Valley Authority Sequoyah Nuclear Plant Updated Final Safety Analysis Report
- 2. WCAP 12504, Summary Report Process Protection System EAGLE 21 Upgrade, RTDBE, NSLB, MSS, EAM and TTD Implementation, Sequoyah Units 1 & 2, March 1990
- 3. WCAP-10961 Rev.1, Steamline Break Mass / Energy Releases for Equipment Qualification Outside Containment - Report to the Westinghouse Owners Group High Energy Line Break / Superheated Blowdowns Outside Containment Subgroup, October 1985 4 SECL-86-451, Sequoyah Steamline Rupture with Consequential Rod Withdrawal-Core Response Analysis for Control and Protection System Interaction, Sequoyah Units 1 and 2, November 1986.
- 5. TVA-92-226, " Improved Interim Operating Procedure for Boron Dilution in Modes 4 and 5,"
December 1,1992, TVA-89-582, " Uncontrolled Boron Dilution Event Reanalysis and Interim Operating Procedure, FSAR Revision Section 15.2.4, " February 22,1989, TVA-84-162, " Boron Dilution Concerns at Hot and Cold Shutdown, " August 22,19984, and NS-TMA-2273, " Boron Dilution Concerns at Cold and Hot Shutdown,~ " July 8,1980.
I 1
i muona Page 8 of 10
SECL-94-063 Revision 0 CONTROL SYSTEM ,
. PARAMETRIC STUDY ON OPAT / OTAT DYNAMIC COMPENSATIONS f
Background
Sequoyah Unit 2 is experiencing sporadic hot leg temperature fluctuations. These fluctuations are postulated to originate in the reactor vessel upper plenum. He hot leg temperatures are measured using fast-response RTDs installed in thermowells located 120 degrees apart in the same plane. Three RTDs are used to account for temperature streaming to provide an average temperature in the hot leg. Since the average of the three RTDs is used to represent the hot leg temperature, temperature fluctuations at any RTD can adversely affect the calculation of the average T-hot temperature. This, in turn, impacts the average temperature, Delta-T (AT) and the margin to the overtemperature and overpower Delta-T (OTAT and OPAT) trips / turbine runbacks. Since there have been OPAT turbine runback alarms at 100%,
Sequoyah Unit 2 has been operating at a reduced power (98-99.5%).
In order to recover some of OPAT/OTAT margin, the dynamic comperuation used in the OPAT/OTAT setpoint margin calculations are being revised. TVA requested Westinghouse to predict the margin recovery if: the lead / lag compensations is reduced from 12/3. This report describes the results of a parametric study on the OPAT/OTAT margins.
Data Analysis Coincident plant data from all three hot-leg RTDs and cold-leg RTDs are required to reproduce the i OPAT/OTAT alarms. Since this data collection is very laborious, Westinghouse suggested using a , l
! hypothetical hot leg / cold leg RTD signal that would predict one percent OPAT margin at 100% power.
A simulation of the OPAT/OTAT trip system with current dynamic compensation would then be compared for several lead / lag and RTD filter time constants.
The hypothetical RTD signals are shown in Figure 1. It should be noted that there are several combinations of the temperature ramp rate and maximum amplitude that would result in a one percent OPAT margin. The T-hot RTD signals shown in Figure I have a rise time of four seconds, a typical RTD time constant. If the rate of increase is higher than that simulated, a smaller temperature increase -
would result in OPAT alarms / trips. Also, if the actual rate of temperature increases are lower than that simulated, the margin recovered will be more than that determined with a four second rise time.
l The hypothetical T-hot' signature is shown in Figure 1. The T-hot average (Figure 2) and i T-avg (Figure 3) are then determined using a corresponding T-cold. He delta-T calculated from the above T-hot /T-cold variations are shown in Figure 4 along with the lead / lag compensated delta-T for a
, lead / lag ratio of 12/3. His lead / lag delta-T is compared with the OPAT/OTAT setpoints (Figures 5 and '
.6) to calculate the margin to OPAT/OTAT (Figures 7 and 8). As shown in Figure 7, the current OPAT margin is about 1%. The same temperaturs fluctuations are used to calculate OPAT margin for a lead / lag ,
ratio of 8/3 and 5/3 (Figure 9). '!
Results
- The results are shown in Table 1. If the lead / lag stic is reduced from 12/3 to 5/3, the OPAT margin
, increased from 1.0% to 3.6%.
wr %>i.4a Page 9 of 10
r .,
SECL-94-063 ;
Revision 0 I 1
1 I
i I
i Table 1 Comparison Of GPDT Margin Lead / Lag Ratio i
12 0 BS 5H l l
1.0 2.5 3.6 l
tvimua Page 10 of 10
Attaciunent i Parametric Study Curves For Margin To Trip 4
.u. A rL E % ,,A- % V D..W .-
s e
e HYPOTHETICAL SEQUOYAH T-HOT SIGNATURE 612 -
611 I
610
u.
m 609-m E
O w 608 O
l-~
@607 4
m 606 -
x D
in g 605 -
2 604 -
s.
. i 603 -
602 ' ' ' ' '
0 5 10 15 20 25 30 35 Time, Seconds FIGURE 1
8r-"' ' $ a
-b TVA LOOP 3 T-HOT AVERAGE 610 -
609 -
608 -
u.
$ 607 -
m uj 606 -
O td y605 g 604 -
O I 603 -
A l l
602 -
601 -
600 . . . . . . . . . . . . . r >
i Time, Seconds l
FIGURE 2
577.4 - ,
577.2 -
i i
! ' I
. v>
577 -
m
. m tr o
W o 576.8 -
d '!
ki 576.6 -
g i
j 576.4 -
576.2 i
0 5 10 15 20 25 30 35
- Time, Seconds f
d h'
1
- FIGURE 3 ,
4
4 CALCULATED DELTA T & UL DELTA-T 107 -
- \
106 -
f-
/ .
O~ '
j\ ,
\
g 104 - / 1 / 1 I
z 103 - e t / I j
$ 1 \ ,
e . .
m 102 e \ / ,
g h f \ .I
- g,101 - ,
i ,', ,,
\
, I r .
r
$ 100 ' ,, y '
g j
\, /
g 1
. st 99 - I *'
g i
, g
} ., .r 98 - a I s ,,
97 ' ' ' ' '
O 5 10 15 20 25 30 35 Time, Seconds DELTA-T - - UL DELTA-T FIGURE 4
. . ., . .. . .=. ... . . ,
~ .:(p ;.;-
_ _ 3-
__ j. -- (A
- e?
- TVA LOOP 3 OPDT SETPOINT 108.8 -
i 108.6 - 1 m ,
O 108.4 -
h 2
. W o 108.2 -
cr w
Q.
p- 108 -
6 n.
$ 107.8 -
m -
H i
O
- o. 107.6 -
O 107.4 -
107.2 ' ' - - >
0 5 10 15 20 25 30. 35 Time, Seconds 6
FIGURE 5
. - - . _ _ _ _ = - _ = _ - _ - - - . . _ _ .
- ._. - . . - . 1 y ,Y".,
9-TVA LOOP 3 OTDT SETPOINT 118 - '
117 -
m -
O '
< 116 - ,
b
=
@ 115 -
m .
D.
p 114 -
-6 o.
F 113 -
m -
u)
H o 112 -
0 111 -
110 ' ' - - -
0 5 10 15 20 25 30 35 Time,' Seconds FIGURE 6
a n -- _ . , % g.
b .g- -
g"%--
e MARGIN TO OPDT 12 - .
LU O 10 -
h z .
tu O
E 8-tu
. 1 -.:
CL~
Hi F 6-Q CL O
O 4-I-
. _Z_
e .
E g 2-0 '
a 0 5 10 15. 20 25 30 35 Time, Seconds n
FIGURE 7 '
- - . - - ~ . . .
- i, -
o_zFO - 0FOFFEa
- - - :.aW0OWZhoW 1 1 2 2 0 5 0 5 0 5 0
5 '
1 '
0 M -
- A 1 R
- F T 5 G
I I
G i
m N -
U e T R
E O
S' 8 e O c T o D n T d
s 2 '
0 2 '
5 .
3 '
0
~
3 5
?
n
ea - . ;: - 4 ;.
Q SEQUOYAH MARGIN TO OPDT 14 12 - - ---
tu O .
f f l
$ 10 .& --- -
o /j j
?, A E
l/lhr '
O"% V f U
,_ s o O s V, \ j o
F s/
\ l s \
l
.E 4 i
\. l 1 w 2
\/
0 ' ' ' -
0 5 10 15 20 25 30 35' Time, Seconds LEAD / LAG 5/3 - -LEAD / LAG 8/3 - - LEAD / LAG 12/3 FIGURE 9
9.,_ .
D' 5
ATTACIBfEh"T 2 PIE CIIANGES i
l'
- 3. _
- . oEs :sc-ctsE,::c;4: E m : Lass 2
.: .!d i
- ,,i, !
Page-ta f ':2 l
)
P'(PO OTSP) P' =.2235 psig Nominal RCS operating pressure T'(TAVGO FULL T= 1 578.2*F Nominal T at rated thermal ccwer OTSP) )
ATg (Celta T }O- 0 0
indicated AT at rated thermal power (See Appendix A)
S = laplace. transform operator
-T . Average Temperature, 'F P = Pressurizer Pressure, psig 4- f j(AI) = See Item 4 below aT turbine runback setpoint (OT TRB R8K) (C-3)
- AT reactor trip (TB-4110 TB-4210, TB-4310. TB-4410) setpoint - 3%
- Turbine runback time delay relay .on 1.5 sec '
off 28.5 sec .,
Turbine load reference reduction
. rate 2001/ minute-Parameter Four-Loon Oeeration ,
Kj (K1 OTSP) 1.15 (1151)
K2 (K2 OTSP) 0.011 (1.11/'F) 0.00055 (0.0551/ psi)
K3 (K3 OTSP) tj (TAU 1 Lead TAVG) 33 seconds t 4 seconds 2 (TAU 2 LAG TAVG)~
> t 4 (TAU 4 Lead Delta T) bconds 55 (TAUS Lag Delta T) .3 seconds
- 2. Overpower AT trip and turbine runback / Block automatic .l rod withdrawal AT reactor trip setpoint (OP. Trip):
~!
(TB-411G, TB-421G, TB-431G, T8-441G) l l
I+T 5 T5 3 Overpower AT (j ,34 ) 1 AT, (K4-K5 (1 +5 3 ) T -6K l
(T - T") - f CAII) 2 8864e:10/080191 l
4; j j'- '
4 57:NGacLSE ;CP;;I 2D ~353 2 ' 55 - l E aeviscn 1: I Page 15 of 102 where, l
4 1+t 2S
- , g = Lead-Lag compensator on measured at T .= Average temperature, *F -
(TAVG1 Full OPSP: T" - Indicated average temperature at rated thermal power (calibration temperature for aT S = sf m o er r (Delta TO) 00 indicated af at rated thermal power (TAU 3 Rate TAVG) t 10 seconds 3
(K4 OPSP) K 4
- 1.087 (108.7%)
-> (TAU 4 Lead Delta T) t g .hskconds (TAUS Lag Delta T) t 5 = 3 seconds (K5 GPSP) K 5
- 0.02/*F for increasing T 0.0/*F for decreasing T (K5 OPSP) K 6
= 0.0011/*F for i > T" 0.0/*F for T 1 T" f 2(AI) = 0 for all AI AT turbine runback setpoint (OP TRB R8K) (C-4)
. AT reactor trip t (TB-411H, TB-421H, TB 431H, TB-441H) setpoint -3%
- 3. Nuclear calibration for AT trips During plant startup tests, all eight calibrated current signals from the power range nuclear channels are to be calibrated from core power distribution measurements such that the same signal (defined at 100% normal current) is obtained for the reference flat power' condition. The reference flat power condition is defined as rated core power with nominal plant conditions and equal power in the top and bottom halves of the core (qtop - 50%.
" bottom - 501).
8864e:10/080191
a w a ., a>- ,m - 4 . -.. __ a m d
ATTACEDENT 3 TECH SPEC CHANGES 1
4 4
3 1
, - y -
7
$ I I. I. . =
- c, c
It v gy, 3
, II 2
- g. S. _. =,
% m sg
-, t i s as -
s: 2- 12 -Z 1 =s 2
- 12._,g l = ei e
= ~
. .s 2.y ,g 2 Um -
g ._ _ 3 _s _ ,
~
~
E i 1 i
a s a e
t_i .-
.l_ =. . 33 : ,- .- .. -
i
,g.
=
i e_~
i i i i- =1 =l a u a =l -
=.
sI i
e 4
I ,=-,: 1
- ,.1 ,.1 s .1 a
i n x e
.3 n ~i.
- =
.4 .3 a .3 i
- = =
a -
g:a
= :
_8 :
n <
1 6 g '
,m t
.i s
'j- *
,j2 w
-i .
g-l
! I3
_]
c.
vi w
a :] el 1 -
3 tu .-
-*]jisaa -
- I
!I , _'3 8
.,s s1.I 5
II !}3 I 1 l 1*a 3 - - ' -
. 5- - .2 er li}
as 1.1 t
. 1a
=
8
_I L j3 i
3
" 3 lja 8
w
.3 } }
! g-
- i. .!a,I 3
l!.$j
. 1I t 3 a .!1.I - -
J e.
z xn a SEQUCYAH - UNIT 1 27 Amenesent No. 19. 114, 141
- - - - - MAY1o'id0
- s TA8tE 2.2-1 (Continuedl
{
Y x REAC10R 1 RIP SYSTEM INS 1RiseM1AT10N 1 RIP SEIPOINTS --,
NOIATION (Continued)
@ N01E 1: (Continued)
'T
~ l+rSg
= I 3,2g The function generated by the lead-lag controlter for I, dynamic compensation
> nio
=
t y, & 12 cens s u H ire in e I secs. as con mHer for T,,g, r 3 3 s u s.,
T = Average temperature *F T' 5 578.2*F (Nominal T,,, at EATED THEmeAL POW R) y K = 0.00055 an 3 P =
Pressurizer pewssere, psig
- P' =
2235 psig (Nominal ACS eperating pressure)
S
= Laplace transform operater (sec-I) and fy(AI) is a function of the ladicated difference between top and bottom detectors of the power range nuclear ion chambers; with gains to be selected based on measured instrsament response during plant startup tests such that:
(1) for tq ~\ -
Percent W + 5 penent3 f (al) = e (he g t and g gg are percent RATED THENGAL POWR la the top and bottom halves of the core respectively, d= and q + is total TWWGAL POWER in percent of RAIES TMNGAL POWR).
-< g t
- h t9
,=;. o .i w
e
. o. -
J E IABLE 2.2-1 (Continued) g REACIOR ItIP SYSTEN INSI888ENIAil0011 RIP stipe 3mgs h
e N0IAII0et E
Q
= NDIE 1: 5 $
Seertemperature AT (I * '4 ) $ AT, (K g -E 2 I I*155 * *I )II - I'l * "3(P-P') - 8 1(AI)1 1*v5 a t 81 2 %
1*v54 ubere: 3,35
- I'8 ** 9 "***"*' ** ***' d AI 5 h..
4 ,v' 3
= Tlas constants utilized in the lead-lag controller for AI, r 6 secs, Se> -- AD 4 -
AT, = ladicated AT at RAIES THEst4L POWER ys
.e g nza g i 1.15
, R
- s'nll 2 '
1*i5I
= 1he function generated by the lead-las controller for Ier52 I,,, dynamic campensaties at :1 u,&1 = Tlas constants utillaed la the lead-lag centroller for g 2 I,,', ag @l33 secs,,
g in-e===-@ \
a 1 = ne.r e t oratore r CC Cg -
P. 1* lmi st
$ 578.2*F (Iteminal 1,,, at WEs Mm NR)
CD ! E = 0.00065
, 3
- ; ." p =- Pressurizer pressure, psig
.u 3 pi = 2235 psig (Nominal RCS operating pressure)
. - - - - - _ - - - _ _ - - - . - . .