ST-HL-AE-1704, Forwards Amended marked-up Pages to 860115 Draft Rev 5 to STS (NUREG-0452).Amends Include Editorial Changes,Revs to Reflect Plant Current Design & Additions to Meet Conditions Identified in SER
| ML20204F396 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 07/29/1986 |
| From: | Dewease J HOUSTON LIGHTING & POWER CO. |
| To: | Thompson H Office of Nuclear Reactor Regulation |
| References | |
| CON-#386-200, RTR-NUREG-0452, RTR-NUREG-452 OL, ST-HL-AE-1704, NUDOCS 8608040175 | |
| Download: ML20204F396 (119) | |
Text
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The Light Company u.,- ,n ugnung u,,m m m>x im u-i.,n.wxamumuw2n July 29, 1986 ST-HL-AE-1704 File No.: G9.6 Mr. Hugh L. Thompson, Jr., Director Division of PWR Licensing - A Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555 South Texas Project Units 1 & 2 Docket Nos. STN 50-498, STN 50-499 Technical Specifications
Dear Mr. Thompson:
My letter to you dated January 15, 1986 (ST-HL-AE-1548) transmitted our South Texas specific markup of Draft Revision 5 to the Standard Technical Specifications (STS) for Westinghouse plants (NUREG-0452).
Attached for your information and review are amended pages to the January 15, 1986 submittal. Changes provided in this amendment include editorial changes, data previously marked "LATER", revisions to reflect STP current design, additions to some specifications to meet conditions identified in the Safety Evaluation Report related to the operation of the South Texas Project (NUREG-0781, April 1986) and corrections of typographical errors. Please remove the affected pages and insert the appropriate pages as indicated in Attachment 1.
HL&P intends to delete in a future amendment the following specifications, tables and/or surveillance requirements in accordance with recently approved technical specification improvements and Generic Letter 86-10:
Section or Table Title Table 3.3-2 Reactor Trip Response Times Table 3.3-5 Engineered Safety Features Actuation System Response Times 3/4.3.3.4 Meteorological Instrumentation 3/4.3.3.8 Fire Detection Instrumentation Table 3.3-11 Fire Detection Instruments 3/4.3.3.9 Loose Part Detection System 1
\
8608040175 860729 PDR ADOCK 05000498 L3/NRC/ai E PDR
_ _ - - - _ _ - _ _ - - _ - 1
Houston Lighting & Power Company ST-HL-AE-1704 File No.: G9.6 Page 2 Section or Table Title 4.4.9.1.2 Reactor Vessel Material Irradiation Surveillance Requirements Table 3.6-1 Containment Isolation Valves 3/4.7.10 Fire Protection Table 3.8-1 Containment Penetration Conductor Overcurrent '
Protection Devices Table 4.11-1 Radioactive Liquid Waste Sampling and Analysis Program 3/4.11.1.2 Radioactive Effluents - Dose 3/4.11.1.3 Liquid Radwaste Treatment System Table 4.11-2 Radioactive Gaseous Waste Sampling and Analysis Program 3/4.11.2.2 Dose - Noble Gases 4 3/4.11.2.3 Dose - Iodine 131, Iodine 133, Tritium and Radioactive Materials in Particulate Form 3/4.11.2.4 Gaseous Waste Processing System 3/4.12 Radiological Environmental Monitoring B3/4.7.10 Fire Protection Systems B3/4.7.11 Fire Rated Assemblies 6.3 Unit Staff Qualifications The future amendment will delineate what sections of the Final Safety Analysis Report, Fire Hazards Analysis Report, operating procedures, Offsite Dose Calculation Manual, and inservice inspection program will be revised to include the requirements previously contained in the technical specifications.
If you should have any questions on this matter, please contact Mr. M. A. McBurnett at (512) 972-8530.
Very truly yours, A^^A W
. G. Dewease Vice President Nuclear Plant Operations FAW/ljm Attachments: 1) Removal and Insertion Directions for Amended Pages
- 2) Amended Pages to January 15, 1986 Submittal L3/NRC/ai
. . 's Houston Lighting & Power Company ST-HL-AE-1704 File No.: G9.6 Page 3 cc:
Robert D. Martin Brian E. Berwick, Esquire Regional Administrator, Region IV Assistant Attorney General for Nuclear Regulatory Commission the State of Texas 611 Ryan Plaza Drive, Suite 1000 P.O. Box 12548, Capitol Station Arlington, TX 76011 Austin, TX 78711 N. Prasad Kadambi, Project Manager Lanny A. Sinkin U. S. Nuclear Regulatory Commission Christic Institute 7920 Norfolk Avenue 1324 North Capitol Street Bethesda, MD 20814 Washington, D.C. 20002 Claude E. Johnson Dreste R. Pirfo, Esquire Senior Resident Inspector /STP Hearing Attorney c/o U.S. Nuclear Regulatory Office of the Executive Legal Director Commission U.S. Nuclear Regulatory Commission P.O. Box 910 Washington, DC 20555 Bay City, TX 77414 Charles Bechhoefer, Esquire M.D. Schwarz, Jr., Esquire Chairman, Atomic Safety &
Baker & Batts Licensing Board One Shell Plaza U.S. Nuclear Regulatory Commission Houston, TX 77002 Washington, DC 20555 J.R. Newman, Esquire Dr. James C. Lamb, III Newman & Holtzinger, P.C. 313 Woodhaven Road 1615 L Street, N.W. Chapel Hill, NC 27514 Washington, DC 20036 Judge Frederick J. Shon Director, Office of Inspection Atomic Safety and Licensing Board and Enforcement U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 Citizens for Equitable Utilities, Inc.
T.V. Shockley/R.L. Range c/o Ms. Peggy Buchorn Central Power & Light Company Route 1, Box 1684 P.O. Box 2121 Brazoria, TX 77422 Corpus Christi, TX 78403 Docketing & Service Section H.L. Peterson/G. Pokorny Office of the Secretary City of Austin U.S. Nuclear Regulatory Commission P.O. Box 1088 Washington, DC 20555 Austin, TX 78767 (3 Copies)
J.B. Poston/A. vonRosenberg Advisory Committee on Reactor Safeguards City Public Service Board U.S. Nuclear Regulatory Commission P.O. Box 1771 1717 H Street San Antonio, TX 78296 Washington, DC 20555 Revised 5/22/86
Attachment 1 ST-HL-AE-1704 Page 1 of 3 Removal and Insertion Direction for Amended Pages Please remove pages from the January 15, 1986 (ST-HL-AE-1548) submittal and insert amended pages from Attachment 2 according to the directions given below:
Remove Page (Front /Back) Insert Page (Front /Back) 1-8/Page No. 9 (12/17/85) 1-8/Page No. 9 (07/15/85) 1-9/- 1-9/-
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Attachment 2 ST-HL-AE-1704 Amended Pages to Draft Revision 5, January 15, 1986 Submittal
c DMR :
1 TA8LE 1.1 FREQUENCY NOTATION leTATION M ~
5 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
0 At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once per 7 days.
M At least once per 31 days.
Q At least once per 92 days.
SA At least once per 184 days. .
R At least once per 18 months.
S/U Prior to each reactor startup.
N.A. Not applicable. g P Completed prior to each release.
)
O -
h i
=
G e
y-sTs 1-s 1/M/86 NUREG 0452/STPEGS COMPARIS0N
DRAFT
, TABLE 1.2
~ '
OPERATIONAL N00E5
\
l REACTIVITY I RATED AVERAGE C001. ANT
. ICOE CONDITION. K gf THERMAL POWER
- TEMPERATURE .
- 1. POWER OPERATION 1 0.99 > 55 1 350*F l
- 2. STARTUP 1 0.99 i 55 1 350*F
- 3. NOT STAN08Y -
< 0.99 0 1 350*F
- 4. HOT SHtJTDOWN < 0.99 0 350*F > T
> 200*F avg
- 5, COLD SHUTDOWN < 0.99 , 0 i 200*F
- 6. REFUELING ** 1 0.95 0 1.tde*F I40 .
. p -
- Excluding decay heat.
. ** Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed. ,
e t
e e
O W-STS 1-9 .
~
07/15/86 NUREG 0452/STPEGS COMPARISON
l Prge No. 9 07/15/86 CouPARIScN oF NUREG 0452 REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE (
............... ...... .................................................. ... Operating Moce- .
DESIGN 2) Table 1.2 Operational Modes. Average Coolant July 1986 TS Amend.
Temperature for refueling changed to <=140 F te comply with standard.
e G
'A SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS DRH '
l
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS
' REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set,consistant with the Trip Setpoint. values shown in Table 2.2-1. -
APPLICABILITY: As shown for each channel in Table 3.3-1.
ACTION:
- a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistant with the Trip Setpoint value.
- b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Value 6 column of Table 2.2-1, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of g.
Table 2.2-1 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 ~
was satisfied for the affected channel, or Declare the channel inoperable and apply the applicable ACTION 2.
statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 , I + R + S $,TA .
Where:
Z = The value from Column I of Table 2.2-1 for the affected channel, R = The "as measured" value (th percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column 5 (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 2.2-1 for the affected channel.
O W-STS 2-4 01/15/86 NUREG 0452/STPEGS COMPARISON
Pcg3 N3. 14 07/15/86 CcMPAoISCN OF NUREG 0452. REV.5. AND STPEGS TECH. SPECS.
NOTE TYPE NC*E a
......................eE5 .......... ..... .. g=E , gg
- 2.0 PAGE: 2.0 2- 5.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS DATA i) Table 2.2-1 Reactor Trip System Instrumentation FSAR TABLE T.2-4 Trip Setpoints. Preliminary STPEGS values OP MODE: 1 234 56 provided. STPEGS specifte loop thermal design flow proviced. e DATA 2) Table 2.2-1 Reactor Trip System Instrumentation July 1986 TS Amend.
Trip Setpoints.
- Loop Design Flow changed to reflect current design.
~
S em 9
O O .
c0 TABLE 2.2-1 i REACTOR TRIP SYSTEM INSTRtMENTATION TRIP SETPOINTS i
U
" SENSOR TOTAL ERROR AtLOWANCE (TA) Z {S) TRIP SETPOINT ALLOWABLE VALUE FilNCTIONAL UNIT
- 1. Manual Reactor Trip N.A. N.A. H.A. N.A. N.A.
I
- 2. Power Range, Neutron Flux
- a. liigh Setpoint D 7. 5}^'. )[4.56[0 109]E of RTP** SQ11.flXofRTP**
- h. Low Setpoint 3 [8. 3] ' }[4.56 p 6 $l25]% of RTP** h27.]%ofRTP** ,
- 3. Power Range, Neutron Flux,- [M011.6 > [ 0. 5 }# 0 h5hofRTP**with ${6.]%ofRTP**with a time constant liigh Positive Rate a time constant 82]'ieconds 32]Teconds of RTPa* with
- 4. Pnwer Range, Neutron Flux, [ tall.6 3[0. 5]# 0 f(5 of RTP** with 1[6 a time constant a ti constant High Negative Rate y 1}2), seconds 32 seconds ,
m -
/ 30.9 t
- 5. Intermediate Range, D17.0]4 -r{8.41}w 0 1[St35 of RTP**
Neutron Flux
$[25{ofIITP**
r
- 6. Source Range, Neutron Flux D17.0]# v[10.0 0 k5
,_ l0 dps hl.4x105[ cps
- 7. Overtemperature AT D6.fl# [ [ See Note 1 See Note 2 See Note 4
- 8. Overpower AT [hi]5.5 hl./](M0.2FSeeNote3
- 9. Pressurizer Pressure-tow [t-813.1 h0.71dll.5]# 1[ fps #1g 1 is O 2398.7 Pressurtrer Pressure-liigh M3.1[ M O. 71]'v-{ 1. 5} 238 M psig 1[f9%]3psle gg 10.
- 11. Pressuriner Water Level-liigh h 5.Ol # of instrument kh .
>{2.18[J1.5}($[92]Espan of instrumenth93.8 span hb m
- 12. Reactor Coolant Flow-Low 32.5[ h j h90 of loop design flow
- h89./
design flow" of loop Y
g
- Loop design flow = w
& 95,4007 hgpm
- **RTP = RATED TilERMAL POWE R
Fage No. 15 07/15/86 COM*ARISDN OF NUREG 0452 REv.5 AND $1 PEGS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Mode--- . . . ..
=* 2.0 PAGE: 2.0 2- 6.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS DATA 1) Table 2.2-1 Reactor Trip System Instrumentation FSAR TBL 7.2-4 Trip Setpoints. Prettminary STPEGS values OP M00E: 1 234 56 providea.
DESIGN- 2) Table 2.2-1 Reactor Trip System Instrumentation FSAR 7.2 Trio Setpoints. Item 14 Steam /Feedwater Flow OP MODE:t 234 56 Mismaten and Note (**) coleted. Not applicable to STP design.
EO 3) Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints. Item 15 - 18 renumbered 14 - 17 OP M00E: 1 234 56 due to coletion of item 14
?
DESIGN 4) Table 2.2 1 Reactor Trip System Instrumentation FSAR 7.2 Trip Setootnts. Item 16.a. Changed to ' Low CP M00E: 1 234 56 Emergency Trip Fluto Pressure
- to reflect STP design.
- 5) Tacle 2.2-1 Reactor Trip System Instrumentation .
Trip Setpoints. Item 16. Turoine Trto values to be OP M00E.1 234 56 provicea LATER.
ED 6) Table 2.2-1 Reactor Trip System Instrumentation Trip Setootnts. Item 17. 'ESF* changea to 'ESFAS' OP M00E:1 234 56 for clarity.
DATA 7) Table 2.2-1 Specific values provided for Turbine July 1986 TS Amend. .
Trips 16a & 16b.
h h
o o t
I
'T TARLE 2.2-1 (Continued) ,
m REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS
- d SENSOR TOTAL- ERROR filNCTIONAL UNIT All0WANCE (TA) Z 15) TRIP SETPOINT Att0WA8tE VALUE
- 13. Steam Generator Water [30:0)15.0 [27:381k l.51 1 M it of narrow t h of narrow level Low-Low II l9 range instrument ranSe instrument span span 147--5 team /feedsatsv=F)our: [16.0}- :[13.24] {1;S] 3[40]Faf-foit- x(42.-S]Ersf-futt-
-Mi smatses - steam floor at-RTWt*==-ilmer st:RTP^^
foincia " "itte
-$ team-Co-.i v Li.ier. -[12.0]. -[9,18]u [1.5] 25{,32;3]Osf_ . -TE . 4 3 --i - ; . -
ieve1-i ;-- Tange-instriumsut:- rangs-instrument-5 Pas so,o2.0 W
- - ' 9+tf 4 o.3 -7.
1/. Undervoltage - Reactor th0110.6 01T2#1 1L483Ghvolts 1[4380] volts Coolant Pumps 4
1% underfrequency - Reactor 07: 5] 3.4- 0.01- kO. ] b 57. z h57.1 z Soolant Pumps 17.' Turbine Trip 2.32. l pt.s.
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- a. Low3 Fluid #$$ Pressure ik e 2A' 22 3f900) ps4 y"-9 -- ; ,- -
tsauR AefEur LATER . LATER
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Page No. 16
- 12/47/85
, CCMDARISON OF NUREG C452 REV.S. AND STAEGS TECH. SPECS.
NOTE TYPE NOTE e NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Moce...
i
== 2.0 PAGE: 2.0 2- 7.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ED 1) Table 2.2-1 Reactor Trip System Instrumentation
( Trio Setpoints. Item 19 - 21 renumoered 18 - 20 CP M00E: 1 234 56 due to coletion of ttom 14, l DATA 2) Taele 2.2-1 Reactor Trip Systems Instrumentation FSAR 7.2 l Trio Setpoints. STPEGS specifte preliminary CP M00E: 1 234 56
) values provided.
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e g3 No. 2o 07/15/86 COMPARISON OF NUREG 0452. REV.5 AND STPEGS TECH. SPECS.
NOTE TYPE NCTE a NOTES FSAR CROSS REFERENCE
............... ...... .................................................. .. 0cerating Mooe---
I DATA 2) Changed K6 to read T 1 T" and for K6=0 T<T". July 1986 TS Amend. l t
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! 8ASES FOR SECTION 2.0 SAFETY LIMITS i
AND LIMITING SAFETY SYSTEM SETTINGS .
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01/15/86 NUREG 0452/STPEGS COMoApisnN
7 l . ._ . - - . --- . . -
2.1 SAFETY LIMITS ~ !
1 BASES -
2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel .
and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented i by restricting fuel operation to within the nucleate boiling regime where the 1 heat transfer coefficient is large and the cladding surface temperature is j slightly above the coolant satur tion temperature. 4Q j Operation above the upper boundary of the nucle its boiling regism coul.d result in excessive cladding tqeperatures because of the onset of departure free nucleate boiling (DNB) anc the resultant sharp l eduction in heat transfer coefficient. DNS is not a directly seasurable params ter during operation and therefore THERMAL PCWER and rea ctor coolant temperatt re and pressure have been related to DNB through the W-3 h correlation. The W-3,i0N8 correlation has been
- l developed to predict the DNB flux and the location of DN8 for axially uniform and nonuniform heat flux distributions. The local DN8 heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DN8 at a particular core location to the local heat flux and is indicative of the margin to DNS. ,
The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is Itaited to 1.30. This value corresponds to a 95% probability at a 95% confidence level that ONB l will not occur and is chosen as an appropriate margin to DN8 for all operating conditions.
l '
The curves of Figure 2.1-1 .2 [2.1-2] show the loci of points of .
[
THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel axit is equal to the enthalpy of saturated liquid.
- 1. 'ri 2 These curves are based on an enthi ipy hot channel factor, F N of1.5/a,nd a reference cosine with a peak of 3::ssr u for axial power shape. AgH,11owanceis a included for an increase in F"g at reduced power based on the expression:
F"g=1.5![1+0. (1-P)]
Where P is the fraction of RATED THERMAL POWER.
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f t(AI) function of the Overtemperature trip. When the axial power imbalance l is not within the tolerance, the axial power imbalance effect on the Over-tamperatureATtripswillreducetheSatpointstoprovideprotectionconsistantl with core Safety Limits, w-STS
- B 2-1 07/15/86 NUREG 0452/STPEGS
SAFETY LIMITS BASES 2.1.2 REACTOR C0OLANT SYSTEN PRESSURE Coolant System (RCS) from overpressurization and thereby prevents the relea of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor vessel, pressurizer, and the RCS piping, valves and fittings are designed to Section III of the #SME Code for Nuclear. Power Plants which pemits a. maximum transient pressure of 110% (2735 psig) of design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.
3107 L' The entire RCS is hydrotested at 125% (3220 3 psig) of design pressuref to demonstrate integrity prior to initial operation.
e e
O y-STS s 2-2 .
07/15/86 NUREG 0452/STPEGE COMPARISON
2.2 LIMITING SAFETY SYSTEM SETTINGS i
BASES l
2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS l The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal .
, values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the core and Reactor Coolant ,
System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in sitigating the consequences of accidents. The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is withinL the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Satpoints can be seasured and calibrated, Allowable Values for the Reactor Trip Setpoints have been specified in
- Table 2.2-1. Operation with Setpoints less conservative than the Trip Set-point but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accosusodate this error. An optional provision ;
has been included for determining the OPERASILITY of a channel when its Trip i Setpoint is found to exceed the Allowable Value. The methodology of this '
option utilizes the "as measured" deviation from the specified calibration I -
point for rack and sensor components in conjunction with a statistical coccin-ation of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equa- l tion 2.2-1, I + R v 5 < TA, the interactive effects of the errors in the rack I and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and
^
rack drif t and the accuracy of their ' measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip. R or Rack Error is the "as measured" devia-tion, in percent span, for the affected channel from the specified Trip Set-point. 5 or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptions. ,Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTA8LE EVENTS.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty sagnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not ,
met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
W-STS 8 2-3 01/15/86 NUREG 0452/STPEGS .
esem
m O
i SECTION 3/4.1 REACTIVITY CONTROLS SYSTEM j
O l
O 01/15/86 NUREG 0452/STPEGS COMPARISON
Prge No. 24 07/15/86 CouPARISON OF NUREG 0452 REv.5. AND STPEGS TECH. SPECS.
NOTE ~ TYPE NOTE a NOTES FSAR CROSS REFERENCE :
.......--.. ...... .........--....-- .....---.----------.... ..-----. ---Operating Moce--- , ~i
- 3/4.1.1 PAGE: 3/4 1- 1.0 REACTIVITY CONTROL.BORATION CONTROL.SHUTOOWN MARGIN - Tavg > 2OOF DATA 1) 3/4.1.1.1 STPEGS specific values providea.
OP M00E: 1 234 --
DESIGN 2) 3.1.1.1 Reference to (n) loop eneration coletec FSAR 7.7.1 3.7.1.o e since STP will not pursue n-1 loop operation. OP MODE.1 234 -.
EO 3)
Delete 4.1.1.1.1.b since it is already covered by July 1986 TS Amend.
rod insertion limits.
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8 3/4.1 REACTIVITY CONTROL SYSTEMS ,
3/4.1.1 80 RATION CONTROL SHUTDOWN MARGIN - wTav_ GREATER THAN 200*F )
l LIMITING CONDITION FOR OPERATION i l I.7S7.
3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to SuGES Ak/k. I I
APPLICA8ILITY: MODES 1, 2", 3, and 4.
ACTION:
I.75 With the SHUTDOWN MARGIN less than (hSR): Ak/k, immediately initiate and .l continue boration at greater than or equal to 30 gpa of a solution j
containing greater than or equal to 7.000 ppe boron or equivalent until the required SHUTDOWN MARGIN is restored.
~
SURVEILLANCE REQUIRENENTS ,.
4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to "1 E] Ak/k: '
l I.757e
- a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod (s) and I at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is inoperable.
If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);
L. When in 00 1 er "00: 2 ith ;' ,7f WM O.K. E q C O 1 M l le.;t ec.;; ,.;r 12 te.re by J.rifying th.t ;;.0r;1 t.at with-:r:::1 f:
.M hir, th. lieft; ef Opecifiee;.iei. 3.1.0.", I
- b. gr. When in MODE 2 with X,ff less than 1, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to l achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6; I
- c. 4 Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.le. below, with the control banks at the maximum inser-tion limit of Specification 3.1.3.6; and "See Special Test Exceptions Specification 3.10.1. l W-STS 3/4 1-1 .
07/15/86 NUREG 0452/STPEGS COMPARIS0N
Page No. , 25 07/15/86 Coo ARISON OF NUREG 0452 REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE ..
............... ...... .................................................. ... Operating Moce... , ,'
ED 2) 4.1.1.1.le changed to d. to reflect deletion of b. July 1986 TS Amend.
,m
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQb!REMENTS (Continued) cl /.- When in MODE 3 or 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors: ,
_ 1) Reactor Coolant System baron concentration, --. - - - - - - --
- 2) Control rod position,
- 3) Reactor Coolant Systes average temperature, ,,
- 4) Fuel burnup based on gross thennal energy generation,
- 5) Xenon concentration, and
- 6) Samarium concentration. -
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within 11% Ak/k at least once l per 31 Effective Full Pcwer Days (EFPD). his comparison shall consider at least those factors stated in Specificatic 4.1.1.1.1,e. W above. The l credicted reactivity values shall be adjusted (normalized) to correspond to -
the actual core conditions prior to e.xceeding a fuel burnup of 60 EFPD l after each fuel loading.
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i W-STS 3/4 1-2 07/15/86 NtAEG 0452/STPEGS COMPARISON
I Pcge No. 26 12/17/85 COMPARISON OF NUREG 0452. REV.5. AND STPEGS TECH. SDECS.
NOTE TvPE NOTE a NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ..-Operating Moce--.
..s 0
,_ ... i
, .,7 . . . , - , .- ,. . ,, . , , . . - .
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CON 0! TION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be: *
- a. Less positive than 10 Ak/k/*F for the all rods withdraiwn, beginning l of cycle life (80L), hot zero THERMAL POWEA condition; and
- b. Less negative than 10 4 Ak/k/*F for the all rods withdrawn, and of cycle life (EOL), RATED THERMAL POWER condition. -
APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2* only**.
Specification 3.1.1.3b. - MODES 1, 2, and 3 only**.
ACTION:
- a. With the KTC acre positive than the limit of Specification 3.1.1.3,a. ,
above, operation in MODES 1 and 2 may proceed provided: ,
- 1. Control red withdrawal limits are established and maintained O.
sufficient to restore the MTC to less positive than 0 Ak/k/*F l within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; l
- 2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its Itait for the all rods withdrawn condition; and l.
- 3. A Special Report is prepared and submitted to the Commission, pursuant to Specification 6.9.2, within 10 days, describing the value of the seasured MTC, the interim control rod withdrawal
, limits, and the predicted average core burnup necessary for
! restoring the positive MTC to within its limit for the all rods
, withdrawn condition.
l
- b. With the MTC scre negative than the limit of Specification 3.1.1.3b. l above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l l
"With X,ff greater than or equal to 1.
""See Special Test E.xceptions Specification 3.10~.3.
O' W-STS 3/4 1-4 01/15/86 NUREG 0452/STPEGS COMPAP' SON
_ . _ _ _ - - ~ _ _ - - - _ _ _ _ _ _ _ _ _ - _ _ -
Deg3 N3. 28 07/15/86 CoupantSON CF NUREG 0452 REv.5, AND STPEGS TECH SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE
..........----. .----- ---------....-------.....----....----.------------ ---Coerating Moca---
=* 3/4 1.t.3 PAGE: 3/4 1- 5.0 REACTIVITY CCNTROL SYSTEMS. Mc0ERATOR TEMPERATURE COEFFICIENT DATA 1) 4.1.1.3.D STPEG5 Specific values proviced FSAR 4.3 OP MODE: 1 23---
EO 2) 4.1.1.3 eettertal corrections OP MODE.1 23---
DATA 3) 4.1.1.3b STPEGS Specific values provided. July 1986 TS Amend.
O l
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l 9
REACTIVITY CONTROL SYSTEMS DRAFT O') SURVEILLANCE REQUIREMENTS 4.1.1. 3 The MTC shall be determined to be within its limits during each fuel
. cycle as follows: .
- a. The MTC shall be measured and compared to the BOL limit of Specifi-cation 3.1.1.3a. , abova, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading; and .
' b. The MTC shall be measured at an> THERMAL POWER and compared to
-3,l JM 10
- Ak/k/*F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppm. In the event this comparison indicates the MTC is more negative than - x 10 4 Ak/k/*F, the MTC shall be rameasured, and compared to e EOL MTC limit of Specification 3.1.1.3b. , at least once per 14 EFPD during th,e remainder of the fuel cycle.
- 3.1
( . -
9 O
E-STS 3/4 1-5 07/15/86 NUREG 0452/STPEGS e n.. , . - , e n.,
I l
Pag 2 No. 29 l 12/17/85 ,
COMPARISON OF NUREG 0452. REV.5. AND STDEGS TECH. SPECS.
NOTE TvkE NOTE
- NOTES FSAR CROSS REFERENCE t
............... ...... .................................................. ... Operating ucce... . I l ** l 3/4.1.1.4 PAGE: 3/4 1- 6.0 i REACTIVITY CONTROL SYSTEMS, MINIMJM TEMPERATURE FOR CRITICALITV i
l l DATA 1) 3/4.1.1.4 Reactor Coolant System temperature FSAR 4.3.2 1
values provided. OP MODE: 1 2----
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POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High .
Trip Setpoin less than or equal to 55% of RATED THERMAL POWER within next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; and l
- 3. Identify and correct the cause of the out-of-limit condition l prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least.
once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or until verified at 95% or greater
- RATED THERMAL POWER. ,
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 5
4.2.4.1 The QUADRANT POWER TILT RATIO shall be detemined to be within the limit above 50% of RATED THERMAL POWER by: ,
- a. Calculating the ratio at least once per 7 days when the alam is OPERABLE, and l l .
- b. Calculating the ratio at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during steady-state operation when the alam is inoperable. l 4.2.4.2 The QUADRANT POWER TILT RATIO shall be detemined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confim that the nomalized symmetric power distribution, obtained from two sets of four symmetric thirable locations or full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
O -
jf-STS 3/4 2-13 01/15/86 l
NUREG 0452/STPEGS COMPARIS0N
Prge No. 60 07/15/86 COMPARISON OF NUREG 0452. REV.5 AND STPEGS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Mode---
- 3/4.2.5 PAGE: 3/4 2-14.0 power OISTR. LIMITS - DNS PARAMETERS DATA 1) 3.2.5 Reference to Table 3.2-1 deleted and the 11miting values for RCS T(avg) and Pressurtzer OP MODE:t - - - - -
pressure will De provided LATER.
EO 2) 4.2.5 Reference to Tanle 3.2-1 deleted. FSAR 4.4.2 DATA 3) CP MODE:t - - - - -
3.2.Sc RCS total flow rate 389,00Cgom acced. FSAR 4.4.2 STPEGS specific.
CP- MODE : 1 - - - - -
DESIGN 4) 3.2.5 Action a modtfted to reflect action to be FSAR 4.4.2 taken when RCS Tavg. or pressurtzer pressure CP MODE:t - - - - -
exceed the Itatts. STPEGS destgn.
DESIGN 5) 3.2.5 Action b added to reflect action to be taken FSAR 4.4.2 when the RCS flowrate falls cetow the expected OP M00E: 1 - - - - -
value. STPEGS design.
DESIGN 6) 4.2.5 Modiftsd to require vertffcation of the FSAR 4.4.2 stated values prior to power operatton in excess CP MODE:t - - - - -
of TS%.
DESIGN 7) 4.2.6 Acced to require Channel Calibration of RCS FSAR 4.4.2 ~
total flowrate instrumentation at least once per OP MODE:t - - - - -
18 months. ~
DESIGN 8) 4.2.7 Added to require RCS total flowrate to be FSAR 4.4.2 determined once per 18 months by practaton heat OP MODE:t - - - - -
Calance measurements.
DATA 9) 3.2.5.c RCS Total flow rate should be 189, 3 July 1986 TS Amend.
600 gpm, STPEGS Specific.
l l
l .
l 1
POWER DISTRIBUTION LIMITS - wilt L 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNS-related parameters shall be maintained within the F8ae*;4 Iimits: '- - - - - - - ?-li 2.: =
< (LATER) *F a.
Reactor Coolant Systas Ty,4 and
- b. Pre s s uri ze r P re s s ure f 1 (L ATER) P.5IG
- c. RCS ren Feae xarr) 389,4oora.
- APPLICABILITY: MODE 1.
ACTION:
THE REAcroR CooLAMT $1SrEn [gg CA PKi$$URit!R PR653uMT
- a. Withm?f tN i-s ~ i. ; exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
, b. WITH ThE h(S TDrAl- 4 0 W TE hEkW D TS MGP~ASU WIT, WsTntM 1. Ms4 is%Ei;& RIStoRL THE RCS rarau rweser To warnt oss secseraait umr eR U) Raouet Tunnet Pouien to isss Thm 50% or RATED THERMAL ltWER AND Repuct rac fewin RAnst NruTnca Fws-Hy Test Streamr To uss THA4 oR EquAt To SS% or RATEb konetAL. fewig Wrstog Tyf Mxf h Novg, SURVEILLANCE REQUIREMENTS A80VC Y
4.2.5 Each of thegara:netersh;f T: !- 3.2-2.- shall be verified to be within its limits at least once perm 5-hcurs/Mo ms ACS Terat.nowgarr seu er orngm<vto To EL WardoM 875 LIrm7 PRoog re cMtATsed Anevs 757, or AATEb THEF.rnAL F%1R ArrtR Enca TVGL loa 0sHG.
4.2.6 Tus RCS rurn rw~<<rt incarn, susa ac asizerzo re a carssi cAUBRArDN Ar L EA$ r CovCE PER l 8 MM rH3 . TMC M4A$v4fhENT lW$rAWEMMrtW .wAU, SE CwoRAT(O WorntN IDAYS PRoca so suc rsaronnAwt er 78: cs..suswc riow nEAsantxt+ T.
4.2.7 Tnd RCS rarm. nonaArt surs et osuamco er rue,s,aa star sAiAar em;nsonrneur Ar uAst oves na 18 mouro.
Y .init 1 met opplicable during either TEDMA!. POWD ramp :sasivsse in excess pereent RAYD TMDMI. Pol.TR per sinute er a TMDMA1. POWER etap -4msresse:in escess of to percent RATE TunMA!. POWu, O
rSTS 3/4 2-14 07/15/.86 NJREG 0452/STPEGS COMPARI',0N
Page No. 61 12/17/85 COMPARISON OF NUREG 0452. REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE .
............... ...... .................................................. ... Operating Moco--- ,
~
3/4.2.5 PAGE: 3/4 2-15.0 POWER DISTR. LIMITS ON8 PARAMETERS DESIGN- 1) Table 3.2-1 Deleted. Information incorporated into FSAR 4.4.2 LCO 3.2.5 (pg 3/4 2-14) for simplicity. (Note: OP MODE:t - - - - -
No n-1 1000 operatton is planned for STPEGS.)
I t-l l
I O O~ Os .
J
, TA8tE 3.3-2 (Continued)
]
i 3 REACTOR TitlP SYSTEN INSTRtBENTATION M5PONSE TIES -
I l fiBICT10NAL tAIIT RESPONSE TIE .
- 12. Reactor Coolant Flow--Lew I
- a. Sta01e Leap (As. eve P-8) <k1]'second
- b. Two Leops (Above P-7 and below P-8) 7
' [1]'second
\ , . n, l 13. Steam Generater h ter Level--Low-Low $D2] seconds ..
j l 14;- Stese-Generator-idstar ,l 1evel-tow tofacident with j Steam /Feedseter-flourMfoustcer -ftrAs j
l }
j , 14 19. h derveltage - Seacter Coelant Pumps
$31.5[soconds l ,!
s *
[ 15.IS. Underfregesency - Reacter Caelant Pumps <he.6[second l l b 16.17. Turbine Trip ,
d Eawa Thr
- a. Law, Flu e Pressesre fl. A. .E -
- b. Teorhine Step Valve Closesre N.A. ,!
- 17. 1 8. Safety injectlen input from E5FAS N.A. g
\ .: ;
) 18.3% Reacter Trip System Interlocks N.A. .
} 19. f9. Reacter Trip Breeters II. A.
2o.71. Automatic Trip and Interlock Logic N.A. -
1 1 EE 1n ~O C3 E85
$i h - - -
A .
o 5: I-
Fage No. 72 07/15/86 COMPARISON OF NUREG 0452. REV.5. AND STPEGS TECH. SPEC 5.
NOTE TYPE NOTE e NOTES FSAR CROSS REFERENCE
.--........---. ...... --..... --.--------...--..........-------------... ... Operating Moco---
- 3/4.3.1 PAGE: 3/4 3-11.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION
- 1) Table 4.3 1 Item 2a. 2D, 3. 4 5. 6. 7 8 9 10. WCAP 10271 11 12. Analog Channel Operational Test frequency CP N00E: 1 2345-cnanged from 'M' to 'Q*. Per WCAP-10721 and associated SER.
DESIGN 2) Tacle 4.3 1 Item 7 Overtemperature Delta - T .
Channel Calibretton changed from 'R (13)* to 'R*. OP MODE: 1 2.--.
DES!CN 3) Table 4.3-1 Item 6 Source Range Neutron Flux .
Channel Calteratton changed from "R(4.5.12)* to OP M00E: 1 2-.-.
- R(4.5)* Notation 12 to not oppitcante.
DESIGN 4) Taele 4.3-1 Item 2a.- Semtannual (R) Channel WCAP 10271 Calieration Actions encanded to include Action 6. OP MODEST 2.---
Now remos *R(4.5.6)*. This enange results from WCAP.10271 ano associated SER.
DESIGN 5) Table 4.3-1 Item 2a - Semiannual, Item 6 Source July 1986 TS Amend.
Range Neutron Flux and Item 7 - overtemperature -
Delta-T were inadvertently not changed in ,
accordance with Notes 2, 3 & 4 above.
M 8 @
I taste 4.3-1 !
9 ,
U
,,. REACTOR TRIP SYSTEN INSTRtSENTATION SURVEILLANCE KQUIRDENTS TRIP !
ANALOG ACTUATING ISOES FOR CHANNEL DEVICE WHICH .
CHAleEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE l FUNCTIONAL INIIT CHECK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED i M.A. N.A. N.A.
R N.A. 1, 2, 3*, 4*, 5*
- 1. Nanuel Reacter Trip
- 2. Power Range, IIeutron Flux ?
- a. High Setpoint 5 D(2,4), JEQ(85) N.A. . N.A. 1, 2 l N(3,4), !
Q(4, R(4, 6),
5,6 ) g 5 R(4) JI'QOS) N.A. N.A. 1***, 2 '
- b. Low Setpelat 2 -
Power Range, Neutron Flux, N.A. . R(4) N.A. N.A. 1, 2 .:
u 3. JI'QOS) y High Positive Rate Power Range, Ilsetron Flux, M.A. R(4) ff'QOS) N.A. N.A. 1, 2 ',
- 4.
- l High IIegative Rate gus)
Intermediate Range, 5 R(4,5) F'{: .; N.A. N.A. 1***, 2 ,j
- 5. ,
Neutron Flux
- 6. Source Range IIeutean Flux 5 R(4,5,42) 6,ItS)
QO)
N.A. N. A. 2**, 3, 4, 5 l !
5 RBS) fitpus) h.A. N.A. 1, 2 l l
- 7. Overtemperature L. e ,
N.A. N.A. 1, 2
- d. Overpower AT 5 R /QO5)
Pressurizer Pressi..e--Low 5 R ft"gos) N.A. N.A. 1 8ES 9.
4AD Pressurizer Pressiwe--Higfi 5 R ff"QUs) N.A. NJA. 1, 2
- ROR 10.
GRE Pressurl'zer Water ievel--Migh 5 R )('Q05) N.A. N.A. I gg II.
C3 -
m 12. Reacter Coolant Ft.w--tow 5-R pqOD N.A. . N.A. 1 l ,
s y,
Page No. 73 12/17/85 COMPARISON OF NUREG 0452. REV.5. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Mode-.. ,
- 3/4.3.1 PAGE: 3/4 3-12.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATICN DESIGN 1) Table 4.3-1 Item 14 Steam Generator Water Level -
Low deleted. Not applicable to STPEGS design. OP MODE:t 2 - - - -
ED 2) Table 4.3-1 Item 15 - 19 renumbered as Items 14 -
18 due to deletion of Item 14. OP MODEST 2 - - - -
DESIGN 3) Table 4.3-1 Item 13 18a. 180 18c. 18d Analog WCAP 10271 Channel Operational test frecuency changed from OP M00E;1 2 - - - -
"M" to 'Q". Per WCAP-10271 and associated SER.
DESIGN 4) Table 4.3-1 Item 14 & 15 Trip Actuating Device Operational Test frequency changed from 'M' to CP M00E:1 2 - - - -
"Q". Per WCAP-10271 and associated SER.
- 5) Table 4.3-1 Item 16a - Change title from ' Low Fluid 011 Pressure
- to " Low Emergency Trio Flutd OP M00E.1 2----
Pressure" to reflect correct title for STPEGS.
- 6) Table 4.3-1 Item 17 - Safety injection Input -
Identify as coming from "ESFAS" rather than 'ESF" CP 400E: 1 2----
for clarification.
l
(
i l
O,
~
' ~
sO ,
i - .d'
{
s ,
1A$tii'8.3-4 (Contineed) N , s, . .-
,5 I,
,l EIEINEEEE8 SAFETY FEATURES ACTUAT'ON I SYSTEM lif51RtBE'KTATION TRIP SETPOINTS h , ..
TOTM.
, SEftSOR ERROR FUNCTIONAL UNIT ALLOWANCE (TA) 2 (5) TRIP SETieffff ALL0lal8LE VALUE ,,
- 5. Turbine Trip and Feedsater .
Isolation u.
Automatic Actuation LegIC AND N.A. N.A. N.A. N.A. II. A.
- a.
- Actuatten Relays .
9"o 87.0 88.8
- b. Steam Generater water (**} h2.18p,41.5p $$8t *3E of $[94,835 of narrow Level--Nigh-High (P-14) - narrow range range instrtment ]
fastrument span.
span.
0.5 1.9 it 538'F 2 536.4*F Censmo T c-tow . 3.3 C. S 30.9% new 4.Z, I.b 3.0 S 30.0'/ Flow
- d. kronaren Fsen- Hec,n cotMcsOENT 'WW RCS Fsen-Low 2.5 2.1 0.6 L 90.0'Jenew t89.67. new E.9 ? 558'r t 5S6.2* F Tm_ Ee"h 4.0 1.1
- 6. Sanry Innenow SEC IrW f AacWE Fet nu. Sarrtr Inunner Teir Senman nuo nuonnetr 4tas.
4.0 I.I l.9 % S5B*r ?S56.z*F
- f. %. -Low caouesceur wem Rsacron Tasv (P.4)** .
$6. AamsiIfary Fee h ter ,
- an N.A. N.A. N.A.
- a. Mananal InftfatIen N.A. N.A.
U
- b. AutamatIc Actantten Le01c M.A. M.A. N.A. N.A. N.A. (~ O n2 g --
oc.- and Actuetten Relays
.$ l. 3 UD is pseiS3 12,18 genseg Mr.5p . > [s ms3E of 33.0
> tas;w35 ef narrow y 5"gE
- c. Steam senerater water Level--Low-Low narrow ra#0e ran0e instrument .
Le
= ro instrument span.
span.
3o ,.
-(
^
^
O b ^_ _l$^ l - - - - - $_I. -- $ _ $ _ - ( U-- r
- iime= :t . -
__ - =-_ ,
l 1
I I
Pcge N3. 97 07/15/86 CcMPAR!s0N Or NuREG e453. REv.5. AND sTPEG5 TCCH. SOEC5.
NOTE TYPE NOTE s NOTES FSAR CROSS REFEREfdCE
............... ...... .................................................. ... operating Moce---
3/4.3.2 PAGE: 3/4 3-34.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS ED 1) Tanle 3.3-4 Item 6.e renumoored to 6.d.
- 2) CP M00E:1 23456 DESIGN Table 3.3-4 Item 6.f " Loss of Offstte Power *, item 6.g "Teto of All Main Feecwater Pumps *, and Item CP MODE:t 23456 6.n *$uctten Transfer on Low Pressure" coleted.
Not applicable to STPEGS design.
DESIGN 3) Tacle 3.3-4 Item 7.b 'Automatte Switenover to Containment Sump. RWST Level-Low-Low
- total CP M00E: 1 23456 allowance value to ca proviced LATER. Other preitminary valves provided.
DESIGN 4) Table 3.3-4 Item 78
- Containment Sump Level coleted . not applicaule to STPEG5. OP M30E: 1 23456 l
DESIGN 5) Table 3.3-4 Item 8.a and 8.0 - Title enanged from
- 4kv' to *4.16kVESF". Consistent with STPEGS OP M30f:1 23456 nomenclature.
- 6) Table 3.3-4 Item 8 STPEoS prei tminary values .
proviced.
OP k00E: 1 23456
~
DESIGN 7) Table 3.3-4 Items 8.a and 8.b - Less than/equel July 1986 TS Amend.
to signs changed to greater than/ equal to.
Consistent with Trip Setpoint and STPEGS nomenclature.
1 0 >
TABLE 3.3-4 (Contimsed) .
. v 9 . ;
d ENGINEERED SAFETY FEATURES ACTUATION SYSTEN IWSTRUNENTATION TRIP SETPOINTS SENSOR TOTAL ERROR FUNCil0NAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT AtteWAttE VALUE
! 6. Aux 111ary Fee &#ater (Continued)
'I
) d.e. Safety Injection See Ites 1. above for all Safety Injectfon Trip 5etpefnts and Allowable Values.
.L-. - tess-of-ef fsi' - N;Ar. - -- --- N. A. -(4800 , ._ - ]p- -
^
- 3. .- irip of Alt-Mets-T -- M. A; :- = - E A._ _ K L . . N. f.. - - - -- _ _ LA.
l -l l
w h_. 5actioer.Transf . .= t- - - -. N.-A. : r - r- Nd, =r- N; A.-
- -3-[44t} 4G.-- 3 [44}-4 ,
! J.... .-
l na
- 17. Automatic Switchover to .
f Containment Sump l f
l 5 m. Automatic Actuation Logic M.A. N.A. N.A. N.A. N.A.
l and Actuation Relays '
(LATfR) 0.2 l.S 19 UAM)
- h. RWST Level--Loir-Low it A. -88;A. 18;*. };:(283K 3; ,1213K
- . }.
Coincident With taats _ _ 1 - , i___? M.R. --E.R. E.A. f [3Gi f=_ i [^^_ E " .
,g
, show-tesei ^i _1_ _ _ t-- ; rt :
Safety injection See Ites 1. above for all Safety Injection Trip 5etpoints and A11oweble Valces. .
l -
- 8. Loss of Power 4.86 E5F 3807 2 Eh
- a. A akV3 8us Undervoltage k.A. N.A. N.A. l'lS7883 AI f its j ,g with ah A u g (Loss of Voltage) volts with ,
s s.93 -
a @ - fee t ,7f second delay. tjee ggS second ac = D . delay.
g b. 3 kV as Undervoltage N.A. N.A. N.A. Ag volts,g voltsf gg (Grid Degraded Voltage) with e - p ; with ;-_ n
- s M6 '
g Two Taas g we v,.w. ,i
- 4elnue. Jake % dela s2eas z .,
B marh(or4;r c. Nearm c o e e. -
vs w& .sx')so ree tr;p w.ll,st)
Cor b ;p # ss see c.e 4-;g,.
2
Pege No. 98 12/1T/85 COMPARISON OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE e NOTES FSAR CROSS REFERENCE
------ -------------------------------------------------- ---Operating Mode---
3/4.3.2 PAGE: 3/4 3-35.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS ED 1) Table 3.3-4 Item 9.d 'Steem Generator Water Level, P-14" deleted. This is covered by Item 5.D. CP MODE: 1 23456
- 2) Table 3.3-4 Items 9.a and 9.D - Values proviced and vertfied for STPEGS. OP MODE: 1 23456 0
DE3!GN 3) Taolo 3.3-4 Item 9d acced - P reflects STPEGS cesign. Preitminary values provided. OP MODE: 1 23456 9
1 .
. c' e
1
I TA8LE 3.3 4 (Continued) .
TA8LE NOTATION $
8 Time constants, utilized in the lead-lag controller for Steas Line Pressure-Low are t g $507 sounds and 12 b^{53' seconds. CHAMMEL CALIBRATION shall ensure that these time constants are adjusted to these values.
A~
- The time constant' utilized in the rate-lag controller for Steam Line Pressure-Megativa Rate-Mighth less than or epual tam 50fseconds. CHANNEL CALIBRATION shall ensure thatVthts time constant kVWusted to this value, n<ssE ARC
- ** TMr riots cwrurs vroutro su T8t La t LAG con nrottEn Fen Cwe~sarno Ta,., set ~, t. I2 stw~n MD %g & 3 3ECDN95. (HMN[L ($LIBRhTION Hu EWuRC THAT ThCST Dm! consTwT3 s%' ADMSIED To THESE VAwis.
1 o
O 01/15/86 PSTS 3/4 3-36 NUREG 0452/STPEGS COMPARIS0N
Page No. 101 07/15/86 c0MPARISON OF NUREG 0452. REV.5 AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Mooa--- '
1 3/4.3.2 PAGE: 3/4 3-37.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DESIGN 1) Table 3.3-5 Item 1
- Manual Inttistion" - delete Item 1.d
- Phase 'B' ! solation", Item 1.e " Purge CP M00E: 123456 and Exhaust Isolation *, Item 1.1
- Reactor Tetp*
and Item 1.m " Start Diesel Generator
- to reflect STPEGS design.
DESIGN 2) Table 3.3-5 Revised to reflect STPEGS assign and nomenclature. Items renumoered as required. OP M00E: 1 23456 STPEGS response times to De provided later.
DESIGN 3) Table 3.3-5 Revised to reflect STPEGS design and July 1986 TS Amend.
nomenclature. Items renumbered as required.
STPEGS final response times will be provided later.
~
.g .
f l
r
- .. . ._. . ._ L i
s M
=lI TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIME 5 INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 1. Manual Initiation
- a. Safety Injection M Pus 4PS N.A. ,
Containment Spray M.A.
raesr b.
- c. CefhaseIsolation f, "A" M.A.
NN .et;*.
y F. CewL.hri: Arsfel4 Tis Mafss- A. 6.
e f. Steam Line Isolation '
M.A.
f, g. Feedwater Isolation N.A.
p *, Auxiliary,Feedwater pumps M.A.
- t. .$ . E'ssential$fd$$eer Water SysTu- N.A. .
. i t$. R.=saovContainment " . ; Fand'Ca2sR.i --
M.A.
f Qv ilfTse,4
- f. o om == = :;- =- Ek.ERCECY STAOP t
N. k . . n. . . . . :~ _.. : . . ' *.' a , ,, . a ,
--E D'r- G==- :tr-. .et;:4, i
- 3. W :t'bfesa{21 Osweeghc.1 fY. /t. M / D23(4 Centainment Pressure--High-1 2.
- a. Safaty :njection A Fvy.rj g)N [n;'" W
- [6J r
! E.M Reactor Trip Craom 5.I.) -
- - w 4 S L* 3 Feeewater Isolation m- 'm . . 7 _ M-*) #"ADb.
C,.
- d. S)t) , f *P"hNe^* Isolation A i '\n 6.4) *args:Cyca.wc r vrurita f ovsac-4xttmast ,Isol2 71 gg ation J g p/j N s}'E ( W" )
^-
- f. 99 Auxiliary Feedwater Fumps 4'tzsy*(M
. F:2.
/ tud *it.tc3 C c.uc --
$.4) _ Essential demo 6eegWater Systre, e ini-enei >i r ,3-Q-"
t<rs.: r-a / 3 uswA K411/M 4.29 3Containment : _ * ' -; Fanf Casas < r ss v -' /r ao P-'
. Vru _
iL tiri')to/ tgo y d <. - 3
.99 Control Room a smEMerCY STAttvr N.A.
,-)
^
{^_E t E ' _ : {- 1_-..._-
! _ _ _ ^_ : ,,[10] f D6}
- g. ,,.m_ .__
(g Sh.* he.i 6ewedors I
t o'
07/15/86 i W-STS 3/4 3-37 NUREG 0452/STPEGS
- COMPARISON
Pag] No. 102 07/15/86 CouPARISON OF NUREG 0452. REV.5 AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES
- FSAR CROSS REFERENCE
............... ...... .................................................. ..-Operating Mode...
3/4.3.2 PAGE: 3/4 3 38.0 .
INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM }NS
- 1) Table 3.3-5 Revised to reflect STPECS nomenclature and design. Items renumbered as required. STPEGS OP MODE:t 2 3 4 5 6 response time values to be provided later.
- 2) Table 3.3-5 Item 3.a.9) delete " Start Diesel-Generators
- since the diesel start and loading CP MODE: 1 23456 times will be included in the values provided as indicated by the notes.
- 3) Ta01e 3.3 5 Add new item 3.j " Steam Line Isolation
- and new Item 3.k
DESIGN 4) Table 3.3-5 Add new Item 4.1
- Component Cooling Water". Required by STPEGS design. OP M00Ett 234 56 DESIGN 5) Table 3.3-5 Revised to reflect current STPEGS July 1986 TS Amend.
Design. .
b t
O,
. I
- * * . . . . ~ . - .
TABLE 3.3-5 (Continued) .
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS
- 3. Pressurizer Pressure--Low
- a. Safaty Injection :46045): Puers 1 (273( }/[12)($}
b.M Reactor Trip-(FRoa S.T.) 1 (2] .
c.M C Feedwater Isolation -- " : : 72-1 ~< (7[3)
- '~
d.St .'* f Phase"A" Isolmtion . ur Vrvruanov -< (17](2)/(27](1) 0.4
^
.-- -----j E ' --t 3 1 solation 1 [25](1)/[10](2)
- f. S) Auxiliary Feedwater hans ~< [60]
C 9.49' Essential % eeu~eWater 3 Sy3Trm 1 (47](1)/[32](2) h., 4
. oNtainment " a-.,
- Fen [CootrAs -(55](1)/(40p2)
. w .a. - .
w.65 fontrol .Roomf.tsoteados Emresta Sre.r:W N.A.
1y -;;; , , ,- a theseWeedes 2::* _.
4 t.io3 '
f c _ _ _1 r _ __ : _u -t __
- r' '*': * ._
- 9. Y'plbh. ,'t.?.W- iWI-a? ' _ _ _ _ _ :_
+-n
. Y&'4.M'd% WQir. sto2h/02]
b sr s$nal) N m 4[G '
bNRe.4teFNip)N*I cNaNcN dn; c pemev*
m.se a* .
a
- fb.m.....m M-bcl*/ n'[D) l , s. af-.< .rs . _,n ,,,, m ,,, ;n; m, , ,m , I.2 I-I
=#--. r _-.;- :- 2 -- : :::::_: - -g; -
8.,,C.6tM M* 0 k V,* M' N: '0 n .*fs. k- N o h _
~,, p
_ - _ = - , .., ;
!M 1 .
- D N L _.**M. [ - - .fM _ -- M 6[32 f
{ ' E ss e'W6' fGeK= Wit.v- N_T._._ ' " ') m-[' 'm
-- ( n _~_ ]_ _
E) Ac=tw c uca. h:t c;;d=2.o% ::
Er - ste30')/ ao3,C
- c : : . . . ._ __:_.:.s m g g-y'cg.a w w iu%nem p >,~p
- .=_----- = = = = _ - - - . . . . .
< t i .1 Ce m - r>>. n -i sure--
%=a:t=une.rs. w -
uma.
8 4f-6; 4 Steam Line Pre)s. s ow io n
- a. Safety Injection (9004) Pumrs 1 (12p5)/[22](#)
6.4 Reactor Trip (FAdm f.I.) { [2]
C. E Feedwater Isolation s ;-,- -- 1 (7)(3) l d. S$ r$bI" Isolation 1(17[2)/[27](1)
^
Ce mraereoa 8.8$ . . , . . . . . L:. _m.umnu_ a ls lation 1 (25](1)/[10](2)
- f. St Auxiliary Feedwater Pe..3 1 (60]
- g. H Essential se=4ckf,,Ta~$er Systm ,i(32p2)fg47pl) g+ L.-24 f%c*Nainment icwy&my Fan [Cco ras 1 (55)(1)/[40],(2)
" a. C : =c ~ % _. .: w e 5:r:= .
2 .n.flz.9 1
07/15/86 W-STS
~
3/4 3-38 NUREG 0452/STPEGS COMPARISON l
Pcg3 Ns. 103 07/15/86 COMPARISDN OF NUREG 0452 REV.S. AND STPEGS TECH. SPECS.
NOTE # NOTES NOTE TYPE ...... ...................,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, FSAR CROSS REFERENCE 3/4.3.2 PAGE: 3/4 3-39.0 .
INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM I'ks
- 1) Table 3.3-5 Aevised to reflect STPEGS nomenclature and destgn. Items renumeered as required. STPEGS OP M00E:1 2 3 4 5 6 response time values to be provided later.
- 2) Table 3.3-5 Item 6
- Containment Pressure-High 3" -
change to Item 5 and delete Item 5.0
- Phase 'B' OP M00E:1 2 3 4 5 6 Isolation
- since it is not applicable to STPEGS.
- 3) Table 3.3-5 Item 7 " Containment Pressure-High 2*
and Item 4
- Steam Flow in two Steam Lines - High" OP MODE:1 23456 delete as they are not appitcable to STPEGS.
- 4) Tabl e 3. 3-5 I tem 12 "'Jndervol tage RCP*. Item 13
- loss of Offstte Power" and Itsa 14 " Trip of all CP M00E:1 2 3 4 5 6 Main Feedwater Pumps *. Delete as they are not appitcante to STPEGS design.
DESIGN 5) Table 3.3-5 Revised to reflect current STPEGS July 1986 TS Amend.
Design. ,
t 1
I l -
l
' s 4
l l
l .
l l
e
.~.~......:.~"'". .~~~.. *~~;".
~
- ~ ~ ~ . ~~T .. . . .. ,
TABLE 3.3-5 (Continued) .
~
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN $ECONOS XC omp e**<-%d 5' f,aSteamLinePressure--Low (Continued) kns o r aU .gi:W M.A.
Control Roosg _,_jw^' Wad ?.*_se\.1secewho. EestAstE
. es.
- y,) ..
ttM. Steam Line Isolation 1 [9]C3) 4
[p.:7.*ContainmentPressure--High-3
. A? Containment Spray 1 [45)(2)/[57]C1)
_-- :=. _:"
~
re- M:r 1$
~ ' ~
~
hntaimEeM PrusN-- Nih ~'d
[4 {-:- ^-i- : ^_ "r==:_ n
- ,%% bt .W#o;n.E
. _ . _- #M b
- - - _ _z= ;;=--._. _ ,., <__
g y= . w - r_. e
'="
. , _ _ , _. _ _ . , . : ; ~ ~ ~ ' _ . T ' ' '
- -_- - e_> ._- r. = a_ _ r_ s _ -
g , ,=,-
?
, M
- f. [# Steam Line Pressure - Negative Rate--High Steam Line Isolation 1 [9]C3) f,,le'.SteamGeneratorwaterLevel--Hign-High '
- a. Turbine Trip i [2.5]
- b. Feeewater Isolation 1 [7)(3)
~
/d. . L.S SteamGeneratorWaterLevel--Low-Cow
-(, b -C- ' _-- Auxiliarf Feeawater Pumps 1 [60]
l
- d. _ _ _ . - ; i t ;; 2. :
N _ U^- 2
_~
< === > -^e =-
__... . . .-__... .__5 2 _ _ _. . 7 -2
^
E : 7 _ + -
- ti A w.ili: j ;;- L;- "_ ,. .eb=b.
W-STS 3/4 3-39 07/15/86 NUREG 0452/STPEGS COMPARISON
Pcg3 N3. 104 07/15/86 COMPAntScN OF NUREG 0452..REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES
............... ...... FSAR CROSS REFERENCE
.................................................. ...Coeest,ng uoa....
- 3/4.3.2 PAGE: 3/4 3-40.0 s INSTRUMENTATION
. . . ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INS '
- 1) Table 3.3-5 Revised to reflect STPEGS design and nomenclature. Items renumbord as required. OP MODE:1 2 3 4 5 6 STPEGS response time values to be provided later.
2)
New Item 10.c acced "4.16kV ESF Bus undervoltage (Sustained Gegraced Voltage)". This reflects OP M00E: 1 23456 STPEGS cenign. Response time to be proviced LATER.
DESIGN 3) Table 3.3-5 New Item 11 acced ' Compensated T_ cold Low. Low" acced to reflect $7)EGS cesign. OP MODE:1 234 56 CESIGN 4) Table 3.3-5 Revised to reflect STPEGS Current July 1986 TS Amend.
Design.
1 8
l
TA8LE 3.3-5 (Continued) .
gayiq i ENGINEERED $AFETY FEATURES RESPONSE TIMES INITIATING $!GNAL AND FUNCTION RESPONSE TIME IN SECONOS w_-- __- u .- _ _ , _ __ _ _ _ . _ . ....
- -_ } _ ; g .--- _:(---
,-- - a- .
- g. $% RWST Level-Low-Low C&m6ded MM% M*b** ..
Automatic Switchover to Containment ila50- Q t.
Se ,w-A d@53 b)
^ -
^
E: ! - _ : _' _ . _ "_z --
L. .J. _ _ : - j ; ;-[
p.....__ m r a ' .-- .. ....,--_-. --
,.s 4dEp4 , ,- s,
_si_nov_ -e/r nun. -
t$.E e
[2,, Lossg{,
- a. 4,kVA us Undervoltage 1 (10]
(Loss of Voltage) 4.I6 ES~
_ b. 8 3kV A -- ._ _ - Bus i gic). f3
- g. '
Uncerveitage ($ste (TOLERABLE
- Degraded Voltage) c.m&uft se b4% Lyb c,. 4.16 d,ESF BoS UNDERVOLTAGE _ s, s (SusinmCD DEGRADED VCl.TAGE) l .
____--_-_._.ol.
7 _ _
-...t.> "'*"
.. -ras s-feg ;ecu r -ere-l
________ . . . , .'.2. - 0I) -EE
_-- -- - _ . q u/
l _ e e--- r t
a :. ; . . ' , M
-..s-,,,,-
-)
'h --, , . . . ,
_ . _{ U ,, M
- . ,,i.;-- - f _ _ '_:^;.v : 1 6 - vo ---- -- -
__ _...: :_._ :_ _ .: _ _ . _ . r .--
580:0
_-- crnr .m- -,. . . - - _---~-n _ _ ger A5E/
,, .. -- M AO8 81 M
.a ,A g A% O I I19 OI #
4 i, ... 72,., .meisng --r-r system
Paig) No. 105 07/15/86 COMPARISON OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE
------ -------------------------------------------------- ---Operating Moca--- .,
3/4.3.2 PAGE: 3/4 3-40.5 .
INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM YNS DESIGN 1) Table 3.3-5 New Items 11 - 17 added. Addition of these items required by STPEGS costgn. OP MODE: 1 23456 DESIGN 2) Teole 3.3-5 New Items 11 -
17 Control Room Makeup Air Radioactivity - High and Spent Fuel Pool DP MODE:t 23456 Exhaust Raatosctivity - High, response times to be provided LATER.
DESIGN 3) Table 3.3-5 Revised to reflect current STPEGS July 1986 TS Amend.
Design.
e 5
e o
S
TABLE 3.3-5 (cmtiroed)
INCINmm SAFETT FIATURES RESPONSE TIMES Initiatina Signal and Functies Response Time In seconds
- - - p -r)
~
col._
2
, J. L r;;n: C;eli;; '.'.;. 0,.;.a ;;7.0 /;;.0
- h. f ::::1 E;n ";;;il;;i;; "- rg;;;, 0;..;., ;;;,0 /;3,0 2)
- 1. ::;__ :.1. !;;1e: er. ere
=. Iv1 bins inP (later)
Containment Airbo ne Radioactivity - Eigh
.p Containment Ventilation Isolation (later)
$ Lew Compensated T;old --Lotu
- a. Turbine Trip 52.5
- b. Feedwater Isolation 87.0 I)
Eigh Feedvater Flow - Ifigh Coincident with 2 of 4 loops Raving Either Reactor Coolant Flow - Lov or T - Low
- a. Turbine Trip - Reactor Trip 52.5
- b. Feedwater Isolation $7,9(
T - low Coincident with Reactor Trip l -ev; I
l - b :- Feedwater Isolation 57.0(3)
. Control Roon Iurar Air Radioactivity - Einh 2C- Control Roon Venin 4raes Energency Startup (later) l
. Spent Fuel Fool Exhaust Radiosetivity - Einh l
4:. FEB EVAC Energency Startup (uter) l O
3/4 3 -40A 07/15/86 l NUREG 0452/STPEGS COMPARIS0N l
l
P393 No. 106 07/15/86 COMPARISON OF NUREG 0452. REV.5 AND STPEGS TECH. SPECS.
NOTE TYPE NOTE 8 NOTES FSAR Cross REFERENCE
............... ...... .................................................. ... Operating Moce...
- 3/4.3.2 PAGE: 3/4 3-41.0 's INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATICN SYSTEM INS DESIGN 1) Table 3.3-5 Table Notations. Items 4 & 5. Change
'RHR pumOs* to " Low-head Safety injection pumps". OP MODE:t 2 3 4 5 6 STPEGS coes not utttize RHR for low head safety tnjection.
OESIGN 2) Table 3.3-5 Table Notation 3. Re'/ised to July 1986 TS Amend.
identify applicable air-operated valves.
~,
0 .o 4
1
- - - - ~ ~ ~ *- **~ ~ * -
7~ ....:......_._-~~...~~..~... . . . . .
5 g TA8LE 3.3-5 (Continued) whR
~
~
TA8LE NOTATIONS
, (1) Diesel generator starting and sequence loading delays included. .
(2) Diesel generator starting --f ::;_ _-__ ::: !:., delay ,n,,o,t include DC' ^ ^ "
- Offsite power available. '
(3)h*/tr-operated valves. -
[
Low HEso SmrvIroter <w
,(4) Diesel generator starting and sequence loading delay included.4 m pumps gLt, included.
g (5) Diesel generator starting ::: :: ;:::: ::: !:-
.itikbuss not included. offsite p;ower avas4wfe. delays not include * ** '
Les nn; .5m*YIMEcnou I"Uen hp s* vdves wren * '84Id* vdve.s , smain de% 'ned O
l l
l l
l i
l l
l l
l O
W-575 3/4 3-41 07 /86 NU 0452/STPEGS COMPARISOf4
I l
l Page No. IC 7 12/17/85 COMPARISON OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
------ -------------------------------------------------- ---Operating Mode---
3/4.3.2 PAGE: 3/4 3-42.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS l
i ED 1) Table 4.3-2 Item 1. Changed to
- Safety Injection l (Reactor Trip. Feedwater Isolation. Control Room CP M00E:1 23456 ;
Emergency Ventilation. Start Stancey Diesel 1 Generator. Reactor Containment Fan Coolers, l l Essential Cooling Water)' to reflect STPEGS '
titles.
( DESIGN- 2) Table 4.3-2 1.e. ' Differential Pressure Between l Steam Lines - High' coleted. Not applicaole to OP MODE: 1 23456 STPEGS design.
1 EO 3) Table 4.3-2 1.f. Changed to 1.e ' Low Compensated l Steam Line Pressure'. Editorial. OP MODE: 1 234 56 DESIGN 4) Ta01e 4.3-2 1.f. Low-Low Compensated T(cold) -
Low-Low acced to reflect STPEGS design. OP MCOE: 1 234 56 l
l l
1 l
l nu;
O F CS .
TABLE 3.3-6
,7
- u. RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS ,
I
'NINIMUM
' ' \ .. '
N ,CllANNELS CllANNELS NAPPLICAhtE Al b / TRIP hlNCTIONkUNIT 10. TRIP /AtARM DPERABLE HbOES N , SETPettiT T10N
- 1. Containme'at N, ', , s l
- a. Containee Atmosph'ere 1 2 All .i [2] slR/h\,
2 2 **
Radioactivi ty 'High s i ,, i b.RCSleakageDetechen N
iI 10 g _
- 1) Particulats Radio ivity . N.A. I 1, 2, h . 4 N.' . \ 29 'N M
' .2) Gaseous Radioactivity 'N ' N.A. 1 1, 2, 3, 4. N.A.
\ ,)
'x w
) 2. Purge' and Exhaust, Ventilation -
y .
a.PartiblateRadi ettwity l s., ,
2 \ All '
- 26 b. Gaseous. R activity s
'1 ,. 2 - All
- 26 E hs. Fuel Storage Poo eas s. 'N N,'
x Ity-High s g -
- l
'x N h, Radioacti ,
N s
/h D fl! :
s Gaseous Radi ctivity 1 2 N ** N, < [2] 27 ps b,Crlthality-Radttionlevel 'N' 1 2 a i 15 mR/h 28 o : l
' \ , ' '. , N L. :
I
- 4. Contral Room \ . 's fs g gg o a. Air IN ake-R ation Leve 1/in'take 2'/ Intake All N< [2] mR/h 7 .
cn ' b. Cohtrol Ro Atmosp re 1 -s 2 All _
] mR/h 27 e l
,N N & I g o, co N 'Radiation-H1
@$ \ - T i n -
t .,
g -
> .l 8
m . i s-t
s Peg) No. 117 07/15/86 COMPARISON OF NUREG 0452. REV.5. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE A
3/4.3.3.1 PAGE: 3/4 3-48.5" INSTRUMENTATION - MONITORING INSTR RADIATION MONITORING DESIGN 1) Table 3.3-6 Replaces STSRev5 page 3/4 3-48 to reflect STPEGS design. OP M00E: 1 234 56 ED 2) For clarity, the following are the ractation monitors provideo for each of these functions: ta. OP MODE: 1 23456 A1RA-RE/RT-8012. C1RA-RE/RT-8013 A1RA-RI-S0128 C1RA-RI-80138 ED 2A) 1b. N1RA-RE-8011C. N1RA-RT-8011 tc. N1RA-RE-8011A. OP MG3F: 1 23456 N1RA-RT-8011A 2a. AinA-RE/RT-8035. C1RA-RE/RT-8036 A1RA-RI-80358 C1RA-RI-8036B ED 28) 3a. A1RA-RE/RT-8033 C1RA-RE/RT-8034 A1RA-RI-80338 OP MODE: 1 234 56 DATA 3) Table 3.3 STPEGS Specific Values provided July 1986 TS 5end for items la, 2a and 3a.
9 S
9
N k k k 0
$ $ =
0 ll 9
1
. 3 3 5 lll lll i nll a ~ 4
_ s s
2e 4-i w '!> ? Y II d
!l al 1
1, J 4
4
- ?
g 4
$n[
s-
. r, i
dr 41 I a%!E d 3 ), if
- g. -
lp ,* ;! s.4 21 ,4-
'u n y H a~
- M x0 m 2, I b d 4 % N YQ 3 c- i i t ~,
la i 1 id i li : 4 lj.
gi gjd I,l h. s .
i e.
!yIl* . l1!.II E: -
4 A C 3.'4 3 48 A NU 52/STPEGS COMPARISON 1
Page No. 118 12/17/85 COMPARISON OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE '
------ -------------------------------------------------- ---Operating Mode--- \,.
3/4.3.3.1 PAGE: 3/4 3-49.0 INSTRUMENTATION - MONITORING INSTR RADIATION MONITORING ED 1) Table 3.3-6 Table Notations - Delete
- and ***
statements and change ** to
- wnich reads ". . fuel CP MODE: 1 234 56 in the spent fuel pool". Tnts reflects requirements for STPEGS.
ED 2) Table 3.3-6 Actions renumbered because of the addition of Action 26 in th) ESFAS Specification. OP MODE: 1 23456 DESIGN 3) Table 3.3-6 Action 27 - Change to read ". FSAR 7.5.1 containment purge tsolatton valves are maintained op MODE: 1 234 56 closed". This reflects STPEGS desiJn and nomenclature and is consistent witt. ESFAS Action 17.
DESIGN 4) Table 3.3-6 Action 28 Changed to identify the FSAR 7.5.1 STPEGS system as " Control Room Envelope HVAC CP MODE: 1 234 56 System in the emergency mode". Tnts reflects STPEGS cesign.
DESIGN 5) Teole 3.3-6 Action 29 changed to provtoe an Action FSAR 7.5.1
. reflective of the safety function of the FMB CP M00E: 1 234 56 exhaust filtration. The revised Action 29 is consistent with Action 27 for the Control Room HVAC and Callaway TS.
m W
. - - - - ~ ~ ' '~
. l- _ .. . ... . . . . . . . . . ...~.-.-.T'".'
.T INSTRUMENTATION .l
~
METEOROLOGICAt. INSTRUMENTATION LIMITING CONDITION FOR OPERATION l
l 3.3.3.4 The meteorological monitoring instrumentation channels shown in Table . l 3.3-8 shall be OPERABLE.
l 1
APPLICABILITY: At all times.
. ACTION:
- a. With one or more required meteorological monitoring channels inoperatie for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the salfunction and the plans for restoring the channel (s) to OPERA 8LE status.
- b. The provisions of Spec'ifications 3.0.3 and 3.0.4 are not applicaele.
SURVEILLANCE REQUIRD8ENTS t
4.3.3.4 Each of the above seteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK anc CHANNEL CALIBRATION at the frequencies shown in Table 4.3-5. ,
l W-575
- 3/4 3-55 01/15/86 NUREG 0452/STPEGS
. COMPARIS0N
Page No. 125 07/15/86 COMPARISDN OF NUREG C452. REV.5. AND STPEGS TECH. SPECS.
. NOTE TYPE NOTE 8 NOTES FSAR CROSS REFERENCE
............... ...... .......................................----------- ---Cperating Mode.-.
- 3/4.3.3.4 PAGE: 3/4 3-56.0 INSTRUMENTATION - METEOROLOGICAL INSTRUMENTATION CATA 1) Table 3.3-8 Meteorological Monitoring Instrumentation Instruments and locations OP M00E: 1 234 56 provided.
DESIGN 2) Taole 3.3-8 Note (*) If tne primary tower detta temperature enannel is out of service the CP M00E: 1 234 56 atmosonerte stan111ty is calculated by tne metnod presented in tne esses Section.
DATA 3) Table 3.3-8 changed elevation level to 197 ft. July 1986 TS Amend.
for Air Temperature AT to reflect actual tower elevation.
W i
O 1
I O
1 O
l
~
TA8LE 3.3-8 NETEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATTON CPERABLE
- 1. Find j$ eed !
rRMARY CK BAUUP '
m ' Mtrteemo6ut mm Neatns1 E1ev. b'b FEET 1
- N -t:
- 2. Wind Direction PRsmut CR hnckUP l
- t. MrTreet es reat Tcwen Noetnai E1ev. 3 5 FRET 1
% M -k.
- 3. Air Temperature - AT
- l
~
's. Pmov Mf7FRoe6icALTeure Nominal Elev. 63-19 FEr7 1*
- h t=
l
- The delta temperature channel may be caken out of service for up to eight hours provided a wind direction channel is gjerab1_e to allev calculation of atmospheric stability. ,
7 07/15/86 y-sis 3/4 3-56 NyREG 0452/STPEGS COMPARIS0N
Page N3, 126 07/15/86 COMPARISON or NuREG o452. REv.5. AND STPEGS TECH. SPECS.
1 NOTS TYPE NOTE # NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ...oper, ting uooe... .
- 3/4.3.3.4 PAGE: 3/4 3-57.0 l INSTRUMENTATICN - METEOROLOGICAL INSTRUMENTATION l
- 1) Table 4.3-5 Values proviced consistent with those presented in Table 3.3-8. OP MODE:1 2 3 4 56 DATA 2) Table 4.3-5 changea elevation level to 197 ft. July 1986 TS Amend.
O 9
e W
e 9
- O .
e
\
- \
l
. . . . . .. c.... .. . .. , . . . . . .. - .. .. C . .. _ .. : " .
TABLE 4.3-5 .
METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
- 1. Wind Speed l s Nominal Elev. 33 rser o gar (
e==moedaa5=Etec. s0= "$fF
- 2. Wind Direction
- z. Nominal Elev. ,
33 FEET o 4k k 1- .
'nel El- _ *.~ -sps: .
- 3. Air Temperature - AT
'l
._ e. Nominal Elev. 33 - 19/ FEZT p .gfr k
" '..;' L .i * $4:
I I
l i
~
l . .
~
O W-STS 3/4 3-57 07/15/86 NURES 0452/STPEGS COMPARISON
Page No. 127 12/17/85 COMPARISDN OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
------ -------------------------------------------------- ---Operating Moco---
3/4.3.3.5 PAGE: 3/4 3-58.0 INSTRUMENTATION - REMOTE SHUTOOWN SYSTEM DESIGN 1) 3.3.3.5 Delete " transfer switches, power, controls and". The Spec is then consistent with Callaway. OP M00E:1 23456 design 2) 3.3.3.5 Action b. Deleted. See Item 1 aoove. FSAR 7.4.1 OP N00E: 1 23456
- 3) 3.3.3.5 Action c renumbered b. due to coletion of Action D. OP MODE: 1 234 56 DESIGN 4) 4.3.3.5.2 Survet11ance deleted. See Item 1 acove.
OP M00E: 1 23456 ED 5) 4.3.3.5.1 Delete *.1* - because of deletion of 4.3.3.5.2. OP M00E:1 23456 i
l l
I l
~
ei
g, f -
) - .
-,; (..
j . ,3 2 ;
i g.
- TABLE 3.3-10 -
ic ; ,
' 5 ACCIDENT NOMITORING INSTRt#ENTAT'I3N +
h ' ,.
s
\'
. REQUIRED '
s
-talAt: M1tilensi '
' - NO. OF CHANNELS s!Nsims*NT fn:- : :.:- --
- 7 :4 CHANNELS OPERABLE ' ACTION i I
- 1. Containment Pressure f3 1 1 sysrca t
- 2. 7I/t. cop 3 Reactor Coolant * **='-Temperature - TH0T (Wide Range) l/ LOOP l .5 Y.5 TEN!
- 3. Reactor Coolant g1Miar Temperature - TCOLO (Wide Range) :S l/teoP 1/toor 3 .
- 4. Reactor Coolant Pressure - Wide Range /ErrENDEO R4W6C 43 . 1 i #
1 5. Pressurtrer Water Level $3 ,
1 . I i
- 6. Steam f.ine Pressure 32/ steam generater* 1/ steam generator I
{ '
- 7. Steam Generator Water Level - Narrow Range 31/ steam generator" 1/ steam generator /
- 8. Steam Generater Water Level - Wide Range 1/ steam generator 1/ steam generator 3 ,
- 9. Refueling Water Storage' Tank Water Level 73 1 I l
, 1^. E--- iv :- ? ^ ? _ ^ -? =^ " . ^ l --2=
- . 22:
l0 AutlLs ARY YEEDWArte Sroose TMK,lYATER LRCL 3 1 $
l
- 11. Auulliary feedwater Flow Amts lf/ steam generator 1/ steam generator 3
- 12. Reactor Coolant System Si&cealing Margin Monitor 2 1 2, U. L -,' 7. 8 4. " ; ih :$/valwe: :tfestws:
- ,ttut0RV_ Btack-01 _.7..itteur-tadicator
- -2/velve- Lh!^^-
I n2o gg 15.--Saf ety. h;.-_;..it4=r-#=*er=*=r- t/vatve :1/welwer I
EW Iff.3 Containment Water Level (Harrow Range) 2 1 2.
3y ,
17f Containment Water level (Wide Range) -73 1, ;
.I G .
p b
- To rns. %een or CawEts us 4- _
b l ,
page No. 134 07/15/86 COMPARISON OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ...Coerating Mooe---
illustration only. not requirements.)
DATA 2) Table 3.3-10 ' Action' required added to table as FSAR TABLE T.5-1 page 3/4 3-634. OP M00E: 1 23--.
DESIGN 3) 3.3-11 Delete Ifne items 19 - Unit vent - High Range Nonle Gas Monttor and item 23 - Reactor CP N00E: 1 234 56 Coolant Raatation Level Monitor and renumoer. Not applicaele to STPEGS.
DESIGN 4) 3.3-11 A5d line Iteme 19 . Neutron Flux (Extenced Range). 20 Containment hydrogen Concentration. 21 OP M00E: 1 23456 Containment Pressure (Extended Range). 22-Steam Generator. 81owcown Radiation Level and 23-Neutron Flux. Startup Range. This reflects STPEGS casign.
DESIGN 5) Table 3.3-10 changed Action on Item 18 to July 1986 TS hinend.
' Action' "2" to reflect actual design. -
Action "4" was inadvertently incorporated previously.
1 S
e 0
O O' . . . . . .O -
) .
TABLE 3.3-10 (Continued)
'[3 ACCIDENT DelliORING INSTRtNNTATICII Rent)l RED 44,* MINisess .
NO. OF CHANNELS
} IN51RtM NT '!?*_:'
^'
_; " ;-)
CitANNELS OPfRABLE ACT10N 1 isCorhbroocouples 4/ core quadrant /reene 2/ core speadrant/FRAf4 I .
i -- - - - -
-- - == -.=tes: -EA: =4=
1 16.p. steam ; "--? ": _' - Radiation Monitor ** I/stran uarc 1/ steam line 3
l _
i g 123. Containment Atmosphere - High Rap 0e Radiation Monf ter St;N. h 1- 3 i
lass. Reactor vessel Water Level (RVWL) 2* - 8 2, $
r : = CM - - =t " - " _ 3 " - " " -- -st:t -2:
u 5 % I %
I9. Nsurnow rwn (rxrenoso Ranoat)'
2 ; g b 20. Cournownwr Hmosta couesargarioa .
2 ) y,
- 21. Covra,wment Pierssone (extruor, anecs) j 3
- /aiownen=r zaur 'Isne.oo.or uur i I
Z2, Steam Gswename Beo.ucown Raoraroow % ron '
! 2, I 2,
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I
page No. 135 s 12/17/85 COMPARISON OF NOREG 0452. REv.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
---Operating Moco---
~
3/4.3.3.6 PAGE: 3/4 3-63.5 INSTRUMENTATION - ACCIDENT MONITORING INSTRUMENTATION DATA 1) Table 3.3-10 ACTION STATEMENTS added to Table. FSAR TABLE 7.5-1 OP MODE: 1 23---
9 8
'il '
s INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LINITING CONDITION FOR OPERATION ~-
3.3.3.10 .The radioactive liquid affluent monitoring instroentation channels shown in Table 3.3-12 shall be OPERA 8LE with their Alaru/ Trip Setpoints set to .
ensure that the limits of specification 3.11.1.1 are not exceeded. The Alarm /
Trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the 0FFSITE 005E CALCULATION MANUAL (00CM). .
! APPLICABILITY: At all times. .
ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel Alars/ Trip 5etpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.
- b. Wit'h less than the minimum number of radioactive Ifquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown O;' in Table 3.3-12. Restore the inoperable instrumentation to CPEAABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.A7why this inoperability was not correctsa within the time specified.
- c. The provisions of Specifications 3.0.3 and 3.0.4, are not applicable.
SURVEILLANCE REQUIREWENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECK. SOURCE CHECK, CHANNEL CALIBRA ION, and ANALOGJGMANNEL OPERATIONAL TESTYat the ' " " "
frequencies shown in Table 4.3-8 as Amsc4w (
l l
l l
0 w.5TS 3/4 3-70 01/15/86 l ,
NUBEG 0452/STPEGS COMPARISON
l Pcg3 No. 153 COMPARISON OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS. !
l NOTE #
NOTE TYPE NOTES FSAR CROSS REFERENCE
.. 0,er tin,Mooe... O; 3/4.3.3.10 PAGE: 3/4 3-T1.0 INSTRUMENTATION - RADIDACTIVE LIQUID EFFLUENT MONITORING INSTR.
DESIGN t) Table 3.3-12 Table revised. those iteme not applicable to STPEGS have been deleted and the OP N00E:1 23456 Action Statements renumbered as the unnecessary Action Statements have been coleted. There are no changes in the requirements.
EO 2) Table 3.3-12 Note: Since Actions in 3.3.3.6 have not been estantished, numeers for actions here CP M00E: 1 23456 revised to x.y. pending resolution 3.3.3.6.
DESIGN 3) Table 3.3-12 Table revised for item la) to show July 1986 TS Amend.
Action X; Action 28 was inadvertently incorporated previously.
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Page No. 153 12/17/85 COMPARISON OF NUREG 0452. REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Moco---
- 3/4.3.3.10 PAGE: 3/4 3-72.0 INSTRUMENTATION - RA010 ACTIVE LIQUIO EFFLUENT MONITORING INSTR.
DESIGN. 1) Table 3.3-12 Item 5 Radioactivity Recorcers deleted. Not applicable to STPEGS cesign. OP M00E: 1 234 56 l
1 I
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I
REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant Systes' Leakage Detection Systems shall be OPERABLE:
- a. The Containment Atmosphereh Gaseous ;r b rLF ici:2 Radioactivity Monitoring System, N w.at
- b. The Containment sockst35 ump Level and Flow Monitoring Systers, and i
- c. G' U= EMcontaie_ W-sooler e-2;nete ;- . n;] : :
Tst Containment Atmosphere (G______ ;r Particulateyadioactivity Monitoring System.
APSLICABILITY: MOCES 1, 2, 3, and 4.
l ACTION:
l '
With only two of the above recuired Leakage Detection Systems CPERABLE, l operation may continue for up to 30 days provided grab samples of the contain-sent atmeschere are obtained anc analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required Gaseous or Particulate Radioactive Monitoring System is inoperable; l otherwise, be in At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREVEN'S 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERA 8LE by:
- a. Containment Atmosphere Gaseous and Particulate Monitoring Systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3,
- b. Containment Pocket Sump Level and Flow Monitoring System performance l of CHANNEL CALIBRATION at least once per 18 months, and
+ , - [5p+dif wc.griate-surveill.re 2- - _, 2d1 1 5x r-e t-pe '
testags Ostect. ion Systam selectett l 0
3/' 21 01/15/86 y-STS NUBEG 0452/STPEGS COMPARIS0N
_ _ . _ - _ . _ _ . _ - . _ , _ , - _ ,, ,, __-._.__-___..-,,_,,...-3, _ _ _ _ - , - - _ _ . _ _ _ _ . - - . _ . . - . -
Pcge N3. 195 07/15/86 COMPARISON OF NUREG C152 REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Moce..-
DESIGN 2) 3.4.6.2.e Deleted. Not applicable to STPEGS FSAR 5.2.5 design. CP MODE: 1 234 . .
DESIGN 3) 3.4.6.2.f and Note (*) It is not possible to test FSAR 5.2.5 valves at full system pressure prior to entry into CP M00E: 1 234 . .
Moce 4 since full system pressure cannot be attained untti Moce 4.
DESIGN 4) 3.4.6.2.f 5gom acced. Tnts issue a result of Tech Spec approved imorovement for tne vogtle Plant and OP M00E: 1 234 - -
ts appitcable to the STPEGS cesign.
DATA 5) 3.4.6.2e Changed leakage rate to 0.5 gpm to July 1986 TS Amend.
reflect STPEGS specific design.
~
W 9
o S
e
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION
( 3.4.,6.2 Reactor Coolant Systes leakage shall be limited to: ,
- s. No PRESSURE BOUNDARY LEAKAGE,
- b. 1 gps UNIDENTIFIED LEAKAGE, l
- c. 1 gpa total reactor-to-secondary leakage through all steas generators . _ ' . .__' _'__f '. __. r _ " .__ _. ^ [
__!_... ^,____ and
^
j t1500fga11ons per day through any one stans generator,moe=eedete*
- d. 10 gpa IDENTIFIED LEAKAGE from the Reactor Coolant Systes, Ano l g '
.. . . . _ ^ ^ '_? "." ' l f ." : I ".r_^ _t_____f__._
- _ __ - r_ _
l Ai"'er4 n Au suen e r vaurc . sore IP TO A Mansaa tuaA e6 . fin PM 6,. &. Om[TgpmleakageiYaa Reactor Coolant Systes pressure of 2235 1 from any Reactor --c Coolant Systes Pressure Isolation Valve ,. - ^;eu n ,nn . s., -
z u ?_ _ . = = , .- : . w_ ,
c a :- a _ .; ,,._....,, ,m,m _,1 -_ :: , .rmo<<*o w rAa c. J.+-//
APPLICABILI*Y: MODES 1, 2, 3, and 4.
ACTION:
- a. With any PRESSURE BOUNDARY. LEAKAGE, be in at least HOT STAND 8Y
, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
! . b. With any Reactor Coolant Systes leakage greater than any one of t.6c i above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leaka'ge
' rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY l within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- c. With any Reactor Coolant Systes Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of :
the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Tssr entss0RES LESS TnAa 2255 PstG hvT GwaiR YnAM 150 r,ssG ARE Allowro, OBSERVED LEAKAGE .sHAu SC ACJu1TED FCR. THC ACTVAL TEST PRG550RE Vf TO AISS rsac MsumeG Thi LE%e6E TD bE DIMR Y PAWbRrtoM TD PAfssu!J DWFERfWr m
. To Tur M - a cJ PerVfR ,
PSTS 3/4 4-22 07/15/86 NUREG 0452/STPEGS COMPARISON L
Page No. 196 12/17/85 COMPARISON OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Moce--- s
- v
- 3/4.4.6.2 PAGE: 3/4 4-23.0 RCS - OPERATIONAL LEAKAGE ED 1) 4.4.6.2.1.b
- pocket' sumo changed to ' normal and secondary
DESIGN 2) 4.4.6.2.1.c Deleted. CVCS is not assotcated with FSAR 5.2.5 SI for STPEGS design. Conststent with 3.4.6.2.e. OP MODE: 1 234 - -
- 3) 4.4.6.2.1
- containment atmosonere radioactivity monitor" should read " Containment Atmosonere CP MODE:
Radtation Monitor. Gaseous and Particulate Channels
- to reflect STPEGS nomenclature.
9
- 9' d
}M REACTOR COOLANT SYSTEM w V ]E 3/4.4.10 STRUCTURAL INTEGRITY
.i. :
~
LIMITING CONDITION FOR OPERATIO'N 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components ,
shall be maintained in accordance with Specification 4.4.10.
APPLICABILITY: All MODES.
. ACTION:
- a. With the structural integrity of any ASME Code Class 1 component (s) not conforising to the above requirements, restore the s*.ructural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NOT considerations.
- b. With the structural integrity of any ASME Code Class 2 component (s) not confoming to the above requirements, restore the structural i
integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F. *
- c. With the structural integrity of any ASME Code Class 3 component (s) not conformir.g to the above requirements, restore the structural
, integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.
- d. The provisions of Specification 3.0.4 are not applicable. ,
SURVEILLANCE REOUIREWENTS l
[
4.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.
O W-STS 3/4 4-40 01/15/86 NUREG 0452/STPEGS COMPARIS0N
I Ptge N3 215 07/15/86 COMDARISON OF NUREG 0452 REv.5, AND STPEGS TECH. SPECS. i NOTE a
. NOTE TfDE NOTES FSAR CROSS REFERE'.;E
...Operatteg Moce--- e!
- 3/4.4.11 PAGE: 3/4 4-41.0 RCS - REACTOR COOLANT SYSTEM VENTS DESIGN 1) 3/4.4.11 Change " Reactor Coolant System vent path" to " Reactor vessel head vent path" to reflect CP M00E 1 234 - -
STPEGS nomenclature.
- 2) 3.4.11.a. b & c Delete to reflect STPEGS cesign.
OP MODE:
- 3) 3.4.11.1 Deleted, anc 4.4.11.2 modified since LCO FSAR 5.4.15 reQutres valves to De closed. OP M00E:
EO 4) 3.4.ll.b Deleted "...two or more...". July 1986 TS Amend.
W O
1 I
J
.3 L REACTOR COOLANT SYSTEM x
3/4.4.11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION ho VEsSEs Hfho Ar L. CAST su m s 3.4.11 **=4 West =wwgeactord..:.... .,.. _ vent path} consisting ofdtwo;l vent . _
valve,y,sR = ,..,-a e...- .;i-; powered from emergency busses shall bTOPERABLE and closed,.'. _;^ . ::._ ?;'.';_-:.., ':: i -_;. .
- . -C ..
- '.. .....' '.__ G , -
e ? . . . . . . . . . . . L_ ; _ _ = ] ; = sad:
-_. - P---_ = ^ 2r i e:: t I ,-. E-- ? ,:. ,. . . '.2 APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:"
vrssn MAc
- a. With one of the above Re.a.:torJ_;i... 5 7."..- vent paths inoperable, STARTUP and/or POWER OPERATION may' continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the evnt=w*Wes=upe=stast. valves in the inoperaole vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in NOT STANOBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
eer- vrsst. ne4s l b. Witna & =t Reactor,Costanc=5=eesar vent paths inoperable; maintain the inocerable vent path closed with power removed from l the valve actuators of all the
- A :1 :'.;-:# valves in the
! inoperable vent paths, and restore at least $casijof the vent CAE patns to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ano in COLD SHUTDOWN within tne following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REOUIREWENTS
?. .1.1 sd ^2s-t-- Ca tant =$,.^.a ..... r..., .'...L ..'. ... ..- . -_ ;;
i sw ; ..__...;-: -; ; E. E E p j = ;
^
gy.E^*-!^{4. ; ;., _....,.h;i" y - - : _ - - -- i _ -> -r --~ ,
"-w -'----r--- -z - - - ---
4.4.11.2'1 Each Reactor,-
Vr= h HW'
- Iy-t=- vent path shall be demonstrated OPERABLE at least once per 18 months by
- a. Verifying all manual isolation valves in each vent path are locked in the open position.
l b. Cycling each vent valve through at least one complete cycle of full travel from the control room, and vsssn wrac
- c. Verifying flow througn the Reactor4 "
- ; Sy-n --- vent paths during venting.
W-STS 3/4 4-41 1 -
07/15/86 NUREG 0452/STPEGS COMPARISON
.a A Q
SECTION 3/4.5 EMERGENCY CORE COOLING SYSTEM O .
m O
01/15/86 NUREG 0452/STPEGS COMPARISON
Page No. 216 07/15'86 /
COMPARISON OF NUREG 0452. REV.5. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
------ -------------------------------------------------- ---Operating Moca---
- 3/4.5.1 PAGE: 3/4 5- 1.0 EMERGENCY CORE COOLING SYSTEMS - ACCUMULATORS DATA ED 1) 3/4.5.1 ' Reactor Coolant System accumulator
- name changed to ' Safety Injection accumulator'. STPEGS OP M00E: 1 23---
specific values provided.
D A'T A 2) 3.5.1 Minimum coron concentratton to be provided LATER. OP MODEST 23---
DESIGN 3) 3.5.1 Actiers: Action statements realigned to FSAR 6.3.2 match the LCO's. This reflects the significance of OP M00E: 1 23---
deviation from each LCO.
DESIGN 4) 3.5.1 Action a: 'immediately" changed to "within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
- in order to take credit for level C valve OP M00E: 1 23- -
testing.
DESIGN 5) 3.5.1 Action O & c: The time periods are cased on allowing sufficient time to return the vartaole to GP M00E: 1 23- -
within the required limits.
DESIGN 6) 4.5.1.1.a.1 Add " indications or Dy". FSAR 6.3.2
. OP M00E: 1 23- -
DATA 7) 3.5.lb and d: STP Specific values provided. July 1986 TS Amend.
O_
O
g I 3/4.5 EMERGENCY CORE COOLING SYSTEMS l 3/4.5.1 ACCUMULATORS ,
} LIMITING CONDITION FOR OPERATION h.e- SArcry 1Auscnod (SIS) 4$ 3.5.1 Each tuastus:Costant a System 18CS)3 accumulator shall be OPERABLE with:
- a. The isolation valve open, 9300 9100
)f b.
- isse ecza -
, A contained borated water volume of between3(4306) and348800) gallons,'
y .
2600 .
- c. A boron concentration of between 1900 andafft0DPppe,and k' =3 I :586 590 423 /s70 l
$@$ d. A nitrogen cover pressure of between 4 M and 3 M psig. l Y
g APPLICABILITY: MODES 1, 2, and 3*.
Q3
- t: y ACTION:
on,,e w manso ta=>rs.wwnr w me soMTLD wann vowmt wiso warra vow adn etwov Mwn yugg Mvlon minoww NLuxe
- 46. Withpone accumulator i..,,,... .;,%ro A rx nowuo wurs Mrme t, mas g . . . . . . . . . . . . . . , . . . . . . . . . . . . . . , _ , . . _ . . . . . , , , , , , , . . . _
Ev k stsess:N or be in at least HOT STANDBY within the.next 6z 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig i CI within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
R*w OelT[ )a. With one accumulator inoperable due to the isolation valve being [Urrms closed, either pgs ___._. O open the isolation valveior be in at N oa least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
,, gjn SURVEILLANCE REQUIREMENTS gr3 wgg 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
$$ a. At Isast once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:
$* IN0ccAfioN5 oR BY
- 1) Verifying, byathe absence of alarns, the contained borated l
' wh water volume and nitrogen cover pressure in the tanks, and w
6=
- 2) Verifying that each accumulator isolation valve is open. l E
- b. At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of greater than or equal tohl% of tank volume by
. E verifying the boron concentration of the accumulator solution; and l 3Rb d l
. v
" Pressurizer pressure above 1000 psig.
y-STS 3/4 5-1A 07/15/86 NUREG 0452/STPEGS COMPARISON
Dage No. 217 07/15/86 COMPARISON OF NUREG 0452. REV.5. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
------ -------------------------------------------------- ---Operating Moco---
- 3/4.5.1 PAGE: 3/4 5- 2.0 EMERGENCY CORE COOLING SYSTEMS - ACCUMULATORS DATA 1) 4.5.1.1.c., d. STPEGS values provided (brackets removea). OP MODE:1 23---
DESIGN 2) 4.5.1.1.c. 'otsconnected by removal of the FSAR 6.3.2 breakers from the circuit' changed to ' removed' to OP M00E: 1 23---
reflect STPEGS cesign.
DESIGN 2) 4.5.1.2a Deleted to be consistent with recently July 1986 TS Amend.
licensed plants utilizing Rev. 5.
l
(~
l l
l l
l
1 DRTI EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- c. At least once per 31 days when the ACS pressure is abovekl000 psig[
by verifying that power to the isolation valve operator isVdiscon= , ;
- . :' :' r : M:r: ' :- " c' 7_2* REM OVED.
2 - :- t ,-
M d. At least once per 18 months by verifying that each acetanulator isola-tion valve opens automatically under each of the following conditions: l
- 1) When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) Setpoint, and
- 2) Upon receipt of a Safety Injection test signal. K -
4.5.1.2 Each accumulator water level and pressure channel shall be demon-strated OPERABLE 4j ,
- . ^t 'eert e-ee pe- 11 day = hy m r e -fa - e e e f e . '"*1 M Cu a ' T L-C?EMT!0hAL TOT, ..,4 l
p J-At least once per 18 months by the performance of a CHANNEL CALIBRATION.
O W-STS 3/4 5-2A 07./15/86 NUREG 0452/STPEGS COMPARISON
I P:ge No. 218 12/17/85 COMPARISDN OF NUREG 0452. REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Moca---
- 3/4.5.1 PAGE: 3/4 5- 2.1 EMERGENCY CORE COOLING SYSTEMS - ACCUMULATORS COLD LEG INJECTION DESIGN- 1) 3/4.5.1 Section coleted. Alternate section on FSAR 6.3.2 pages 1A and 2A used for STPEGS. CP MODE: 1 23--.
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O. CONTAINMENT SYSTEMS A PPLlCABI E SURVEILLANCE REQUIRENENTS 4.6. Each C tainee Spray stas all be nstra d OPE E: l
- a. A east once per days veri ng that ch'valv (manual *
. r-o 'ated, o aut ic) in a flow p h that not lo kad, sealed r othe ist ser'ured in sition, ,s in its arrect ositio ; .l
. By rifying/that o reci N ation fl , , each p devel sa diAcharge pressure Af grea than or ual to psig en tasted
, rsuant o Speci ication .0.5; / l c ' At leas't once er 18 no ths, duri shu , by: !
- 1) Verifying that ch au ic valve n the flew path a ustes l to iAs correc position na test s gnal, aryd
/
- 2) Varifying pat each s ay ptamp rts au tically/on a
/
- d. At leaste per ong//est 5 years by per signa'i.
raing an ir or s eflowdest O~ hrough e Ah spray h der and v ifying a unobstrupted. hspray7czzle1[ -
O 01/15/86 PATNOSPHERIC 3/4 6-21A NUREG 0452/STPEGS COMPARISON
I Pcge No. 253 07/15/86 COMPARISON OF NUREG 0452 REv.5 AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Hoce--.
ecuctors' to reflect STPEGS design and comoined as Item a.
Date 2) 3.6.2.2.a Spray adottive tank volumes and concentrations provided. FSAR 6.2.2. 0312.12 0312.13 DATA 3) OP MODE:t 234 . .
4.6.2.2.c STPECS spectftc test signals provided. FSAR 6.2.2 DESIGN 4) CP M00E: 1 234 . -
4.6.2.2.d Revised to cover survet11ance of eductor FSAR 6.2.2 flow rate and Itne flow vertftcation to reflect CP MODE:t 234 - -
capactisties and Itmttations of STPEGS design.
03ston 5) 4.6.2.2.b Modifted to define survet11ance FSAR 6.2.2 regatrements for the Spray Acatttve Suesystem and CP MODE:t 234 - -
ttems 2). 3) and 4 ) acced to e to define the requirements for the Auxillary Sump Acattive Suosystem.
0:sfon 6) 3.6.2.2 Item o acced to reflect the incorocratton FSAR 6.2.2. 0120.20
~
of a Austitary Sumo Acattive Suesytem into STPEGS OP MODE:t 234 - -
cesign. The specirte values of tank volume and NaOH concentratton to De provtcod later.
DESIGN 7) 3.6.2.2 Changed to reflect deletion of July 1986 TS Amend.
Auxiliary Sump Additive Subsystem.
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CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM :(4854024t9:
LIMITING CONDITION FOR OPERATION T4=Ks 3.6.2.2 The Spray Additive ' System shall be OPERA 8LI with:
Tnace SmAv Aootnv l
- a. 4 /f ^^ e ="t%:qasis, essa, sacs Cat n..M'n' , : volume cowrm,q of between 106/ and 1342 gallons of between 30 and 32, X by weight NaOH solution, and ass
- b. he.
h.
srs Gust spray additive eductor ensk capable of adding NaOH solution -sm OPERABLE 1Ar@ MA
. riow.. . e _. .additive m . _ _, tank m . _. to siContainment. 5 pray Systes pumpnamel _.-
y, m mm , - ,- - - . -- - . - - --,3..v_ ,-
3 _- --
,',;f- ,_
~~~'#
! APPLICA8!LITY: MODES 1, 2, 3, and 4. T ",_* 7 171 E'.T,'."~~'
, , , m , , . . . _ . . . . -
l ACTION: , , _ _ , _ _,,__m, , , , _
g&t TEAW of hl
$st B.. WithAthe Spray Additivefystem3 noperable, i restore the system to OPERABLE l status.within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; .;
l restore the Spray Additive System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> ororn be rwa in COLD 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> ka.
rsamsSHUT c= ne 00WN 5 mar witNin the followinlawnasts, ina Seesysrem s sau o w or rM '"* * ^B .
! b.r- w$wy Accorta .EnesHims n OPEMSf wirs wirm 2A men.s cA sf rw Ar MAsr HOT STMDSY ,
l*
i jSURVEILLANCEREQUIREMENTS
- '# N "" # # # ## ## #
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[4.6.2.2 The Spray Additive Systas shall be demonstrated OPERA 8LE: l l !!t !)
g a. At least once per 31 days by verifying that each valve (manual, F ! ui power-operated, or aittamatic) in the flow path that is not locked,
[f g 5
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- a, sealed, or otherwise secured in position, is in its correct position;l l [ ]'
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4
- b. At least once per 6 monthsApe sw rwt Smf Ammvt 86843TE% By!
i * .y4Y 1) Verifying the contained, solution volume in the tanks, and ;
kl Ih h ,
- 2) Verifying the concentration of the Na0H solution by chemical analysis, l b } h..} O h Verifying that each
,N('c .
'. lt I
\ At least vah automatic once in the per flow 18 athmonths actuatas to duringit shutdown, by;l corre 4
D a Ce,vn.i.u-v M . test signal and fue Assaues erm .sasuat.ra sipf(A rwt&A4r tl t V U %b Accortes Tm avn on TrW WS l.*w Law um rur .susenti l
!o p id ht least once per 5 years by verifying: f _ r _ ^ ^ = -E_: ' _- -
j, [ d
^
-- -- .^ U.,-
--! 1-- -- l ^ .
b -- - , - ! - 7_ h ' - i . ., 7 - ...
I, -. _,____.
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- 1) [0iei= 'i :eti=] _ - , f l i
?) [0r;in 1 1 -- 1 --; t ! =- ] -
l
&: )l y{" h i l) sovcson ca v Aart 11 anuren rou a e Wsr As Tw rest .sovner innernte nr ro rw 4Artt*)m mecrog so7,w ro (Lanh osina m- !
p' IrG LIGT1, hrrdh' rni SPthi MoITitt Tkom AND THE EOcrM.S Att Nor SLDCWZD pr ventria kw.
W-ATMOSPHERIC
. 8 Te er pr o m uro 3/4 6-22A kriss B wyn n s. TssriwG. 07/IS/86 NUREG 0452/STPEGS COMPARISON
~
p ge No. 254 12/17/85 COMPARISON OF NUREG 0452 REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES ,
FSAR CROSS REFERENCE -
............... ...... .................................................. ... operating Moce-..
=e 3/4.6.2.3 PAGE: 3/4 6-23.0 CONTAINMENT SYSTEMS . CONTAINMENT COOLING SYSTEM DESIGN 1)
, correct nomenclature including a note that one fan OP M00E: 1 234 . .
may be out of service incefinitely.
DESIGN 2) 3.6.2.3 Action b. c. Revised to cover i and 2 FSAR 6.2.2 groups of RCFCs plus one Containment Spray System CP M00E: 1 234 . .
inoperable to reflect STPEGS design. CSS and RCFC are not redundant as CSS has no post.tnjection heat removal capoeiltty.
OFSIGN 3) 4.6.2.3.a.1 Modtfted to require 1) each FSAR 6.2.2.2 non. operating fan group to De started from control OP M00E: 1 234 - .
room and 2) to be operated for 15 minutes.
0: sign 4) 4.6.2.3.a.: Deleted since the reactor containment FSAR 6.2.2.2 cnilled water ts normally in service and OP MODE: 1 234 . .
monitored.
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. L CONTAINMENT SYSTEMS J/4.6.4 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6./ The containment isolation valves specified in Table 3.6-1 shall be -
OPERABLE with isolation times as shown in Table 3.6-1.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one or more of the isolation valve (s) specified in Table 3.6-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration
.- that is open and: l
- a. Restore the inoperable. valve (s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, .
or
- b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, ;
or
- c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange, or
- d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD
'5HUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .
SURVEILLANCE REQUIREMENTS 4.6. 1 ThbsoIaTon valves specified in Table 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control
. or power circuit by performance of a cycling test, and verification of isola-tion time.
4 01/15/86 W-ATNOSPHERIC 3/4 6-27A NUREG 0452/STPEGS COMPARISON
I Pcg) No. 259 07/15/86 COMPARISON OF NUREG 0452. REv.5, AND STPECS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Moco--.
- 3/4.6.4 PAGE: 3/4 6-28.0 CONTAINMENT SYSTEMS - CONTAINMENT ISOLATION VALVES
- 1) 4.6.4.2 (4.6.3.2) Colete item D. STPEGS has no ecutoment actuated by the pnase 'B' tsolation OP MODE:t 234 . .
signal.
EO 2) 4.6.4.2 renumeered 4.6.3.2. 4.6.4.2.c renumeered 4.6.3.2.0, 4.6.4.3 renumeered 4.6.3.3. 09 M00E:t 234 - .
EO 3) 4.6.3.2.0 Correct nomenclature used for STPEGS is Contatnment venttitation Isoletton Test and OP MODE: 1 234 . .
Containment ventilation Isolation valves.
ED 4) 4.6.3.3 changed "Within its" to "less than July 1986 TS Arnend.
or equal to its".
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CONTAINMENT SYSTEMS BRU SURVEILLANCE REQUIREMENTS (Continued) 3 CoerAidter 4.6./.2 Each4 isolation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHLITDOWN or REFUELING MODE at least once per 18 months by:
~
- a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;
' ^^ " " * ' * * ^^
_- _-_ $ 5_1 t ?__*_!'____ !__! _ ___ _-_*- _ _ - -
'.:'n ' ...
'; <- ' 7. = . ;.. 7.. . -_. .-_ g __-_ _ _ _g7.._.. __
. . . . _ _ - . - . - ..... ,__7._. ___
YifHATCW b.w Verifying that on a Containmenta. _. .. ..._ ~ :.___ ^. Isolatiort test _
signal, eachV,_. ,- _ 1- _:^_ valve actuates to its isolation position. Caramtvr WNrnAnov MnoN I
- 4. 6. (3 The isolation time of each power-operated or automatic valve of Table 3.6-1 shall be determined to be 5ith - itt limit when tested pursuant to
- Specification 4.0.5. (m *" ~ *@# '#.-
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'l PATMOSPHERIC 3/4 6-28A -
07/15/86 NUREG 0452/STFEGS COMPARISON
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/
4 Page No. 260 12/17/85 COMPARISON OF NUREG 0452 REV.5. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE #
............... ...... .................................................. ... Operating Moce.--
3/4.6.4 PAGE: 3/4 6-29.0 CONTAINMENT SYSTEMS - CONTAINMENT ISOLATION VALVES
- 1) Table 3.6-1 Conta tnrnent I w1ation Valves numbers provided. OP MODE: 1 234 - .
F a!.'
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TABLE 3.6-1 (Continued)
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CONTAINMENT ISOLATION VALVES MAXlMdM
$7ROAf TIME FENETRATION VAL'VE NO. FUNCTION _(SECONDS)
M&NDAL AED WICC VALVES (Continued)
M-10 ISI0005C Bigh Bead SI Discharge N.A.
M-19 ISI0030A Low Bead $1 DischarSe N.A.
M-15 1S10030E Low Bead SI Discharge N.A.
.M Il ISIOO30C Low Bead SI Discharge N.A.
M-68 S10058 Accumulater Nitrogen Supply N.A.
OIEER AUTOMATIC VALVES (SI Actuated Valves)
M-25 CC0059 RCFC Chilled Water Supply 10 M-26 CC0070 RCFC Chilled Water Return 10 M-27 CC0137 RCFC Chilled Water Supply 10 M-28 CC0149 RCFC Chilled Water Return 10 M-24 CC0199 RCFC Chilled Water Supply 10 .
M-23 CCO209 RCFC Chilled Water Return 10 l
O!EER AUTCMATIC VALVES (Steam Line Isolation Actuated Valves)
M-2 FV-7412 Mai'n Steam Isolation Bypass 10 M-3 FV-7422 Main Steam Isolation Bypass 10 l
M-4 TV-7432 Main Steam Isolation Bypass 10 M-1 Fv-7442 Main Steam Isolation Bypass 10
. M-2 FSV-7414 Main Steam Isolation 5 M-3 FSv-7424 Main Steam Isolation 5 3/4 6-294 (11) 07/15/86 NUREG 0452/STPEGS COMPARIS0N
Pcg3 No. 271 07/15/86 COMPARISON OF NUREG C452. REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE e NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ...Coerating Moce...
- 3/4.6.4 PAGE: 3/4 G.29.9 CONTAINMENT SYSTEMS . CONTAINMENT !$0LATION VALVES CATA 1) Tante 3.6 1 Contatnment Isolation Valves numoors FSAR 6.2.4 Figure provicoc. 6.2.4 1 OP MODE: 1 234 ..
DATA 2) Table 3.6-1 Other Automatic valves (Auxiliary July 1986 TS Amend.
Feedwater Initiation Actuated Valves) numbers provided.
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TABLE 3.6-1 (Continued)
D CONTAINMENT ISOIATION VALVES max gggg STOW TsmE FENETRATION VALVE No. FUNCTION (SECONDS) l OTEIK AImBl& TIC VALVES (Continued)
(S. sam Line Isolation Actuated Valves)
M-4 FSV-7434 Main Staan Isolation 5 M-1 FSV-7444 Main Ste'an Isolation 5 M-2 FV-7900A /4 AIM STEAM LINE DRAIN 5 M-3 FV-7901A MA,N SrsAm 4,vr dea /N 5 M-4 FV-7902A M%N .5 TEAM Laut D44tu 5 M-1 FV-1903A MAta Strnm Lawr DRA>W 5 M-6 FV-7141 Feedwater Isolation 5 M 63 Fi-7I92 FaowArzn Tsatarsca s M-7 FV-7142 Feedvater Isolation 5 0" M 64 M-8 FV 719; FV-7143 FgowArrR Isetarsov Feedvater Isolation S
5 M-94 h 7i&9 FErWATER hosArsed 5 M-5 FV-7144 Feedvater Isolation 5 y-95 Ft 21So FEtowATER Iso L ATION $
M-5 FV-7145A Feedwater Isolation 5 Valve Bypass .
M-8 FV-7146A Feedwater Isolation 5 Valve Bypass M-7 FV-7147A Feedvater Isolation 5 Valve Bypass M-6 FV-714BA Feedvater Isolation 5 Valve Bypass OTHER AUIUMATIC VALVES (Auxiliary Feedwater Initiation Actuated Valves)
M-62 FV-4150 Steam Generator Blowdown 35' Jg;3l;,
M-65 FV-4151 Steam Generator Blowdown 36 C O
i 3/4 6-29A (12) 07/15/86 NUREG 0452/STPEGS COMPARISON
P:g) No. 272
! ! COMPARISON OF NUREG 0452 REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE _
............... ...... .................................................. ... Operating Mode--- -
- 3/4.6.4 PAGE: 3/4 6-29.9 CONTAINMENT SYSTEMS - CONTAINMENT ISCLATION VALVES DATA t) Table 3.6-1 Containment Isolation valves numeers FSAR 6.2.4 Figure provided. 6.2.4-1 OP MODE:t 234 --
DATA 2) Table 3.6-1 Other Automatic valves (Auxiliary July 1986 TS Amend.
Feedwater provided. Initiation Actuated Valves) numbers e
9
TABLE 3.6-1 (Continued)
CONTAINMENT ISOLATION VALVES AfAxp106 STRDYi TIME PENETRATION VALVE NO. FUNCTION (SECONDS)
OTEIR AUTW ATIC VALVES (Continued)
(Auxiliary Feedvater Initiation Actuated Yalves)
M-64 FV-4152 Steam Generator Blowdown 35" 4 M-63 FV-4153 Steam Generator Blowdown 36ZG M-86 FV-4186 Steam Generator Sample 6 D.
M-86 FV-4187 Steam Generator Sample 6 3:A M-86 FV-4188 Steam Generator Sample ( CL M-86 FV-4189 Steam Generator Sample f 101.
FV-4186A Steam Generator Sample f Aht, M-86 M-86 FV-4187A Steam Generator Sample f JCL M-86 Fv-4188A Steam Generator Sample 6 Kli.
M-86 FV-4189A Stena Generator Sample s ER O
4 07/15/86 3/4 6-29A 03) NUREG 0452/STPEGS COMPARISON
I l
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)
P:ge No. 273 12/17/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE 's'
... Operating Moce..-
reflect correct nomenclature. ,
DE3IGN 3) 4.6.5.1 (4.6.4.1) STPEGS hycrogen analyzer sample FSAR 6.2.5 calibratton gas wt11 es four percent hydrogen with OP MODE: 1 2----
the balance nitrogen. Delete survet11ance a ano move b to body of statement.
- . .s B
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l PLANT SYSTEMS SURVEILLANCE REQUIREMENT 3 (Continued)
- b. AT LEAST CWCE PEA 31 DAYS BY' j)M Verifying that each non-automatic valve in the flow path that !
is not locked, sealed, or otherwise secured in position is in its correct position; and l 2% Verifying that each automatic valve in the flow path is in the f'MFR
? ? ?, ,-.. position whenever the Auxiliary Feedwater System is placed in automatic control or when above 105 RATED THERMAL POWER.
Cr. k At least once per 18 months during shutdown by:
- 1) Verifying that each automatic valve in the flow path actuates l to its correct position upon receipt of an Auxiliary Feedwater Actuation test signal, and
- l
- 2) Verifying that each auxiliary feedwater pump starts as designed l automatically upon receipt of an Auxiliary Feedwater Actuation m test signal. .
4.7.1.2.2 2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying normal flow to each steam generator.
f3) VKRtFYtM Te AT EMn lwu ARY EErowATER Sy.s TEM reorce orgaartp passp D&WO vs.sE u%TS THE now re EM STfA M GlWERATM To BED +M 550 Mc G750m.
Ilb
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5 - i ; N . M ,} .."..-
,v,
.-v.
Nif7 I"-
'";""=
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O W-STS 3/4 7-5 01/15/86 -
NUREG 0452/STPEGS COMPARISON
Page No. 286
- 07/15/86 CcMPantSON OF NuREG e4s2, REv.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ... Operating Mode-.-
- =
3/4.7.1.3 PAGE: 3/4 7- 6.0 PLANT SYSTEMS - AUXILIARY FEEDWATER SYSTEM DESIGN 1) 3/4.7.1.3 ' Aux 111ary Feedwater Storage Tank FSAR 10.4.9 (AFST)* to correct STPEGS nomenclature. OP M00E: 123---
DATA 2) 3/4.7.1.3 STPEGS specific values provided, eachtel calc MC-5865 (3-14-85) FSAR 10.4.9 CP M00E: 1 23---
DESIGN 3) AFST is the seismic category 1 source of primary NRC 0440.39 feedwater for the STPEGS cesign. OP MODE: 1 234- -
DATA 4) STPEGS AFST volume provided. July 1986 TS Amend.
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pdL' f4 pc&. '1 PLANT SYSTEM:L .- ,
CONGENSATE STORAGE TANK .
LIMITING CONDITION FOR OPERATION -
phhdry feMT gg ,
3.7.1.3 The/sondeneste storage tangM shall be OPERA 8LE with a contained-water volume of at least pggpgallons of waterth c cres%s to an AFAT l"Id Md.
SI9,000 APPLICA8ILITY: MODES 1, 2, and 3.
ACTION:
With the 1 hin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either inoperable,w)I IV4 A: /estore the 19T to CPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,se
~
-- P. Dnnoist i
at th 'PERASILI th ite to a te e) l a b :k sp1 to a 1a y a er s ti :S t CP RAI L sta ihn d s be i a l
$"N ) th" th next6 ou and in SHUTV0WN e f 11 ir 6 hourt.
O SURVEILLANCE REOUIRENENTS 4.7.1.3.1 The bFSCshallbedemonstratedOPERABLEatleastonceper12hoursbyl verifying the contained water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps. .
7 - ' 1.3.2
. me Letternate-water scurcej sne!' te erstratec 0=EF35LE et
- 9c: : Sy - -II: {--I -jpp--}-1---} [: ir- :i : j c- - -) -r---
} --} T}
{ --
= :-p-- = : - === - ,_ __
--?---}-
-TueP*=
0 07/15/86 NUREG 0452/STPEGS COMPARISON
Page No. 287 12/17/85 COMPARISON OF NUREG 0452 REV.S. AND STPEGS TECH. SPECS. -
NOTE TYPE NOTE a NOTES FSAR CROSS REFEGENCE
............... ...... .................................................. ... Operating Moco---
- 3/4.7.1.4 PAGE: 3/4 7- 7.0 PLANT SYSTEMS - SPECIFIC ACTIVITY NONE 1) 3/4.7.1.4 No changes.
OP M00E:1 2 3 - - .
l .
l PLANT SYSTEMS 3 /4. 7. 3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR CPERATION THREE .
3.7.3 h independent 4 component cooling water loops shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
Two a.. With only see s component cooling water loop 50PERA8LE, restore - THREE loops to CPERA8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REOUIREMENTS Tunce 4.7.3 h componenta cooling water loops shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that m is not locked, sealed, or otherwise secured in position is in its '
correct position; and l
- b. verif that:
At least once per 18 months during shutdown, ce sg bgas yg mE f,in
, y[RSWE7F
- 1) Eachautomaticvalveservicingsafety-refaiedequipmentactuates 3 to its correct position on a sarrry ^ 2 .
_ u'. , Z INJECT 104 CA LOSS on Orr$ RTE PbWit CR 40eV SdMEPW TE5T.StGNM, AS"APPucALEj ANC
- 2) Each Component Cooling Water System pump starts automatically -
on a bwr tux <tatest signal on es er corsort r0WEA TEC ssGNAr.
i b) h CHMNL CALIBRATsod TEST IS FER*tMGD OV ME SURGE TANK LEVEL n N5 n e < N M Too k w ek.M Pterters normaric ibuArtow Dr THE tbt NGClEAR bhPETY RELATCp PbRTscw of THE Sy5TEr
- b. WITH ONLY OWE COMPsNENT CAKING hv4TER Loor OPERABLE, RESTDRE AT LEAST Two Lacts To CPERAS E srnrus wirgiw 24 noens on ar er wasr ix HOT STANDBY withik THE NEXT fo noeRS Ano ou COLD SHUTDOWN waraiN ruc FOLLOWiNG 30 MegRS, i
lo l W-STS 3/4 7-11
- 01/15/86 -
- NUREG 0452/STPEGS l COMPARISON
Pcg3 No. 292 07/15/86 CouPARISON OF NUREG 0452 REV.5. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES
............... ...... FSAR CROSS REFERENCE
.............-.....--.---.... -..--------- - ..--- --.Operattrg Moca---
.* 3/4.7.4 PAGE: 3/4 7-12.0 PLANT SYSTEMS - SERVICE WATER SYSTEM DESIGN 1) 3/4.7.4 Renameo section 'Essenttal Cooling Water FSAR 9.2.1.2 System
- for correct STPEGS nomenclature. OP MODE: 1 234 - -
DESIGN 2) 3/4.7.4 3 trains must ce operable to satisfy FSAR 9.2.1.2 single fatture cetteria since worst case coston OP MODE:t 234 - -
Dasts events requires 2 trains. Action 0 added to provice for case where only one ECW toop is operacle.
DESIGN 3) 4.7.4.1 Expanded to incluce ECW pumo start. FSAR 9.2.1.2 screen ecoster pump start and essenttal cnt11er CP MODE:t 234 - -
start signals.
DESIGN 4) 4.7.4.b.2 Added vertftcation of ECW pump start on FSAR 9.2.1.~2 either a SI or LOOP test signal. OP M00E: 1 234 - -
DESIGN 5) 4.7.4.0.3 Added to vertfy start of ECW pump. FSAR 9.2.1.2 screen wash cooster pump and traveling screen on 09 M00E:1 234 - -
SI test signal.
DESIGN 6) 4.7.4.c Acced to verify flow rates to eacn heat FSAR 9.2.1.2 ".-
exchanger because of the concern for potential OP MODE:t 234 - -
Asiattc Clam Out1 dup. -
DESIGN 7) 3.7.4 Deleted word " independent" to reflect July 1986 TS Amend.
STPEGS design.
PLANT SYSTEMS EsrEAlTMt. cood%6 3/4.7.4 asseustis, WATER SYSTEM n
LIMITING CONDITION FOR OPERATION Tatt EsstunAi Ca em 3.7.4 h W assesses 3 ter loops shall be OPERA 8LE. .
APPLICA8ILITY: M00E5 1, 2, 3, and 4. -
)
ACTION-Two Essrvr,At Ocuug MET l
, a.. With onlyg% pater loops 0PERABLE, restore at least sus 3 oops 1 to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least NOT STANOBY.within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,
- b. Warn om ent Esst,,ru.. Cain kAn w OTRA&E,arsroer At irrst Two m ro OPERABLE srara wurow 24 nove a se su Ar aasr HOT STANDBY waik rw ext 6 nouns swa osv CO'D SHUToomrwa rnt ronowixo 30 muas. .
SURVEILLANCE REQUIREMENTS THREr Esser.rist Cocuxc .
4.7.4 At least 3 & fater loops shall be demonstrated OPERA 8LE:
O a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; ano l
- b. At least once per 18 months during shutdown, by verifying that:
- 1) Each automatic valve servicing safety-related equipment actuates i to its ruose correct position s er7Y L _ '. . . z iNmrio% l mz,, ECW .srast, stscow on wrsn$ru amHe cAtr Aw iss._'. ,wim cwaux siner.StGkks As
- 2) Eachferm9uw Water tysteer pump starts automatically on a WASEj w SArrry purer,.< test signal g as A tess er arms,rr phase rasr 4*4t
- 40
- 3) Escu assurist emum mren rme, scasu usa aasrex roer an ruc TR.Mauds .sewv .srAtt AerevelocAu.y ok A SArrryZauscitoW TEST.swAnt.
Co. AT L. fast CHCE PER 3l MYS By yergpYiNG TW TUW MA7Es rb acy g6(7 EXCHANGER.
O 07/15/86 w-sTS 3/4 7-12 NUREG 0452/STPEGS COMPARIS0N
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Pcg3 No. 293 12/17/85 COMPARISON OF NUREG 0452. REV.5 AND STPEGS TECH. SPECS.
NOTE TYPF NOTE 8 NOTES FSAR CRCSS REFERENCE
............... ...... .................................................. ... Operating Moce-.-
3/4.7.5 PAGE: 3/4 7-13.0 PLANT SYSTEMS - ULTIMATE HEAT SINK OATA 1) 3/4.7.5 STPEGS spectftc values provided for FSAR 9.2.5 minimum water level. Maximum ECW inlet water CP MODE: 1234 --
temperature provided.
OESIGN 2) 3.7.5.b and 4.7.5 Average water temperature FSAR 9.2.5 changed to ECW intake water temperature since CP M00E: 1 234--
average temperature is not the appropetate parameter for a cooling pond.
OTHER 3) 3.7.5. Action statement consistent with recently approved Calloway Tech Spec. OP MODE:t 2 3 4 --
e
1 I l PLANT SYSTEMS l.
6 a 3/4.7.7 CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM I
LIMITING CONDITION FOR OPERATION 5 Tgti 3.7.7 lus3 1ndependent Control Room Emergency Air Cleanup Systems shall be OPERA 8LE. .
APPLICABILITY: All MODES.
ACTION: ;
MODES 1, 2, 3 and 4:
- a. With one Control Room Emergency Air Cleanup System inoperable, restore the inoperable system to CPERA8LE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
< b. Warer Twc Carrm.. Em EmtaCnENet Asn Ctssuur Srsrums suorumste, assroar ar irrsr no
'" * *^3
MODES 5 *nd 6- -
LE45T HO~ .5$ANfY WITN/W (o THE MWR$ NfXTsr5 AWD sW CDI.D s to SHuTDodN OPERABLE srmras w w.ims rnc Fo.cowius .so uooas,
- a. With one Control Roos Emergency Air Cleanup System inoperable, l restore the inoperable system to OPERA 8LE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control I I
y Room Emergency Air Cleanup System in the regr,cglgijn}m,ogggy TWO oR Mons
- b. WithposhControlRoomEmergencyAirCleanupSystemsinoperable,
, ,,A' or with the OPERABLE Control Room imergency Air Cleanup System, l required to be in the recirculation / mode by ACTION a. , not capable 4,8 ra.ranov/ of being powered by an OPERABLE emergency power source, suspena all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REOUIREMENTS
- 4. 7.7'6 Each Control Room Emergency Air Cleanup System shall be demonstrated l OPERABLE:
- a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control room air temperature is less than or equal tov 'F; g At least once per 31 days on a STAGGERED TEST BASIS by initiating,
%,Tggy,,pi_from the control room, flow through the HEPA filters and charcoal l
AIR FITTER / adsorbersikand Verifying that the systes operates Yor at Itast i tiusr5 / 10 continuous hours with the heaters V operating;
- nct se h&rse avor O
W-STS
~
3/4 7-15 hE 52/STPEGS COMPARISON i
L_._._.._....-___.______..-___ . . , . __. _ _ _ _ . _ _ _ . _ , _ . _ _ _ _ _ . _ ___ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ _ .
Pcge N3. 296 07/15/86 COMPARISON OF NUREG 0452. REV.5 AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE
.....--.--..... ...... ...------------................--. ...--....---... ..-Operating Moce---
3/4.7.7 PAGE: 3/4 7-16.0 PLANT SYSTEMS - CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM DATA 1) 3/4.7.7 Renumberec 3/4.7.6 cue to coletion of FSAR 9.4.1 3/4.7.6. STPEGS spectfte values provicec. OP MODE: 1 234 56 DESIGN 2) 4.7.6 Revisec to reflect STPEGS cesign. FSAR 9.4.1 OP MODE: 1 23456 DESIGN 3) 4.7.6 STPEGS specific values anc nomenclature FSAR 9.4.1 provicec. OP MODE:t 23456 DESIGN /ED 4) 4.7.6 Revised to reflect latest guidance from July 1986 TS Amend.
the Effluent Treatment Systems Branch of the Office of Nuclear Reactor Regulation which recommends using 1980 revisions of ANSI N509 and N510 to the greatest extent possible while still satisfying Table 2 of RG 1.52.
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PLANT SYSTEMS .
SURVEILLANCE REQUIREMENTS (Continued)
- c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in arqy ventilation zone communicating with the system by: -
- 1) VerifyingthatthM1eanupsystemsatisfiestheinplace -
penetration and bypass leakage testing acceptance criteria of less than (f35 and uses the test procedure guidance in Reguia g tory Position C.5.a C.5.c, and C.5.d of Regulatory Guide 1.52, Revisions 2 March 1978, and the system flow rate is
>- 60D0 cfm t 1 fon rne ca.sAwr v417s AND I,ooo CM 28 0?* YO' D'E mp umis' 2)' Verifying, within 31 days after removal, that a laboratory 8 analysis of a representative carbon sample obtained in accor g dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2,. March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.527 Revi- .
l**n sion 2, March 1978, for a methyl iodide penetration of less thandSG%; and , y y,, , y p ,og g y y y ygg,7,
- 3) Verifying a system flow rate of 6,000 cfm + 10%gduring systes operation when tested in accordance with XNSI M510-19580. l; t
, d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, l within 31 days after removal, that a laboratory analysis of a repre-sentative carcon sample obtained in accordance with Regulatory PositionC.6.bofRegulatoryGuide1.527 Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52? Revision 2, March 1978, for a methyl iodide penetration of less tha ;
- e. At least once per 18 months by: ,
gog me w egr N 1) Verifying that the pressure drop across the combined HEPA 4 l unir Mo d.o iec \ filters and charcoal adsorber banks is less than l lugrrA6w ree M ] Water Gaugefwhile operating the system at a flow $ 1nches rate of 7 MMW **r / cfm + 10Q pra ca.caxar uker Mo 4000 evn so% Pee Mnx*vP l UN 2 Verifying that on a g_ i; _ . _ ___
=~ I_-?;i._. - - -
.m %I MD MAdvPb ) ' ^
__ _.._._, test signal, the system automatically switches l AR Fu.rRAT'ON/ into a recirculationAmode of operation with flow through the -
HEPA filters and charcoal adsorber bankger The c4sner Amo Md8vPv"'T5. l
- 3) Verifying that the system maintains the control receYt .
positive pressure of greater than or equal tom 1/8 hinch Water l The svameawo w;- Gauge at less than or equal to a pressurization flow of 2. coo g cfm relative tofsdicoont areas during system operation;
- 4) Verifying that the heaters dissipata 4.5 KW + 10% ter when l tested in accordance with ANSI M510-19 g and J50AArts ruh5) Verifying that on a High W oxic Gas test signal ,
courAa. Aam M ry: E - automaticallyR :--i _- _ _ - __ ; -_-- _1: 1 ---- <
m9m NO -- I r .-- L - - -- :i? " .-- E^ _ _ ,
li-- Y^ i ^ ' ? i;-: :- - - - - - - - -
l
' * ** n ~ G ^ , . % .= -2 .
07/15/86 NUREG 0452/STPEGS W-STS 3/4 7-16 COMPARIS0N
Psoe N3. 297 07/15/86 COMPARISON OF NUREG 0452, REV.5. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ...Operatino Moce---
g-
- 3) OP MODE: 1 234 56 4.7.6 Footnotes (*) and (**) Deleted. Footnotes FSAR 9.4.1 are Instructions for implementing STS to a OP MODE: 1 234 56 spectftc design.
DESIGN 4) 4.7.6 Revised to reflect current guidance from July 1986 TS Amend.
Effluent Treatment Systems Branch - NRR.
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PLANT SYSTEMS
( SURVEILLANCE REQUIREMENTS (Continued)
r,truajaur verifying that thef:!._.., .,. -_ satisfies the in place penetration a,o
, and bypass leakage testing acc iptancecriteriaoflessthan34%in accordance with ANSI N510-19 f5i or a 00P test aerosol while operating the system at a flow rate of 6.000 cfm t 10%;=und rea r>E cumwr umrs .
Ano 1.w tso?.*cen ret rat Mater units.
- g. After each complete or partial replacement of a charcaal adsorber /cuace bank, by verifying that the : :r. :, ^ isatisfies the in placa (AoS*8afR
- t.
- penetration and bypass leakage testing acceptance criteria of less \"
thanjlf@EinaccordancewithANSIM510-195 for a halogenated
'*" hydrocarbon refrigerant test gas while operating the system at a flow rate of 6.000 cts t 10% 504 rm utkro wts Mc1.M cfm tICie $bt Tnt M NGi!P WHITS.
96
% ANSI NSo9-1780 Noe khodo "teri~n d d so b b 0"\ " fi M
ensr t4 slo -1990 he %oc. te.sTn o.te u. sed 'sn cop (Ji% R. G. - /. 5 2. in Me.u. d AN N 603 -I9 N ANSI N&to - 17 75', respciike .
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W-STS 3/4 7-17 07/15/86 NUREG 0452/STPEGS COMPMISON 1
l
I P ge No. 298 07/15/86 CCMPARISON OF NUREG C452 REV.5 AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES l
FS AR CROSS REFERENCE
..... ... -- -------------------------------------------------- ---Operating Moce--- .e
- 3/4.7.8 PAGE: 3/4 7-18.0 I
PLANT SYSTEMS - ECCS PUMP ROOM EXHAUST AIR CLEANUP SYSTEM l
! ED 1) 3/4.7.8 Renumeered to 3/4.7.7 due to deletion of FSAR 9.4.2 3/4.7.6. OP M00E: 1234 - -
1
- since CP M00E: 1 234 - -
l ECCS pump room exhaust is part of FHS exhaust in l STPEGS cesign.
l DESIGN 3) 4.7.7 - Revised to reflect current guidance from July 1986 TS Amend.
Effluent Treatment Systems Branch - NRR.
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9
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~s S- PLANT SYSTEMS a .
g$ 7 FUEL HANDLING BUILDI!G (FHB) 3/4.7.# 4CC8:8W e=#008*, EXHAUST AIR 900MIWh SYSTEM y,3 '
t-*
W k d k, LIMITING CONDITION FOR OPERATION
? E s2 7 THE FHB houst Aun Svstm comentsro or rut" rotwwws emmturs fiE( QPERABLE:
3.7./ .-
^ ---- -
, -_ _-; ^M -
^
3 shall be l a.. Tm suouwotur umsi Alt FILrtR TR4lus s
$, APPLICA8ILITY: MODES 1, 2, 3, and 4)b. c,. Tuari TetREE ,MDENN06NT suornworar M4w DMust EMAOST FANS, Aup h0 STER FANS k y d. AssounTED Dwas, g g C ACTION: u 5 4 h Ti m ___ m . _ .___ _ m .. . . . .
jrestorethe l l f k E inoperable system to OPERABLE status within 7 days or be in at least HOT g g$ TAN 08Ywithinthenext6hoursandinCOLDSUUTDOWNwithinthefollowing30 g g {g hours.
S
- %y
$ SURVEILLANCE REQUIREMENTS
$ &eo5 Tur FHB '
y :.
- 4. 7. 7'7 2:- 7 ._ , Exhaust Air **uenup System shall be demonstratec l OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operate for at least 10 continuous hours with the heaters operatin ware, rwo er r* Aire Exuurr E rm wM mwMg pws rmg To mogrw AMGuMTE
- b. / y rj s y t Two cff. least once per 18 months or (1) after any, structural m on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zcne communicating with the system by:
- 1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of
- f. 0 less thanA'l28% and uses the test procedure guidance in Regula g tory Positions C.S.d, C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 37,000 cfm 2 10%;
- 2) Verifying, within 31 days after removal, that a laboratory i analysis of a representative carbon sample obtained in accor g dance with Regulatory Position C.6.b of Regulatory Guide 1.52,
, Revision 2, March 1978, meets the laboratory testing riteria of Regulatory Position C.6.a of Regulatory Guide 1.5 , Revi-i ,,g_ sion 2, March 1978, for a methyl iodice penetration of less 4,.;,.
than($S"-;%; and 07/15/86 NUREG 0452/STPEGS W-STS
- 3/4 7-18 COMPARISON
P;go N3. 299 07/15/86 COMPARISON OF NUREG 0452. REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
............... ...... ...........--............................-........ ... Operating Moco---
3/4.7.8 PAGE: 3/4 7 19.0 PLANT SYSTEMS - ECCS PUMP ROOM EXHAUST AIR CLEANUP SYSTEM 's-DESIGN 1) 3/4.7.7 Text revised to reflect STPEGS design and FSAR 9.4.2 FSAR commitments. OP M00E: 1 234--
DESIGN 2) 4.7.7.b.3 Inserted "witn two supply fans and 2 of FSAR 9.2.2 the 3 booster exhaust and main exhaust fans OP MODE:t 234--
operating
- to cefine the value of 37.000 cfm.
Added requirement to test all 2 out of 3 fan comninations.
DESIGN 3) 4.7.7.d.2 Changed system starting on SI Signal to FSAR 9.4.2
- verifying that the FHB HVAC emergency test signal OP MODE: 1 234- -
starts the system and directs its exhaust flow through the HEPA filters and cnarcoal adsorbers".
This statement is consistant with Callaway T.S.
EO 4) 4.7.7.d.4 Change 'of greater tha'n or equal to 1/8 inch Water Gauge" to "sitghtly negative" OP MODE:t 234--
clarification.
DESIGN 5) 4.7.7.d.5 Delete - there are no filter coo 11eg bypass valves in STPEGS design. OP M00E: 1 23&--
EO ' 6) Note (*) and (**) Delete - These notes are used to
- dertve the values presented and then deleted. OP MODE: 123I--
DESIGO 7) 4.7.7 Revised to reflect current guidance from July 1986 TS Amend' Effluent Treatment Systems Branch - NRR.
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PLANT SYSTEMS
%fSURVEILLANCEREQUIREMENTS(Continued) bE g* 3) Verifying a system flow rate of 37.ddo cfm + 10% during system l 4 y' operation han tested in accordance with XNSI N510-19 E M m a Au.er osrorruers acesssa ano cao.sar rna ce,asanew susu at reswo.
3
- After every 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> cf charcoal adsorber operation, by verifying,
(
Q$]c. within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained i i
dg g0 Position C.6.b of Regulatory Guide 1.52haccordance Revision 2, March with1978, Regulatory
- w w seets the laboratory testing criteria of Regulatory Position C.6.a
- - of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl
- iodide penetration of less than ;
f i" w d. At least once per 18 months by:
- 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsortier banks is less thanM671nches Water Gauge while operating the system at a flow rate of 31000 ,
cfm + 10%* l FHB HVA C EnrRatsct resi s\GNAL srws THi sys7EM Awo
- 2) Veri fying that ther :;:^ : j__t;_ _:__: i
= . - - ' , os uc rs ars g m s: 7__. f u^;r m ;. 5y 7,re U_^'"g55, farm vuo _
l c w ccal AcsonBERS FHB
- 3) Verifying that the "systes maintains the N Aat a
.suGfm.y negative pressure e' ; __^__. _ _. .. __' a [U^3 '--- =i=:
Whangs: relative to the outside atmosphere, 9 U=i;:,; n ='_ E; Ii;. ___:'.., .,,.... ;I--_- _:= i _.__ ?,
N !
[ 4)# Verifying that the heaters dissipate 50 xW + lo7e *W when l tested in accordance with ANSI N510-1 6 7 g e. After each complete or partial replacement of a HEPA filter bank, gg pm by verifying that theistemnas: systes satisfies the in place pene-tration and bypass leakage testing acceptance criteria of less than 1.0 lei in accordance with ANSI N510-195 for a DOP test aerosol while operating the system at a flow rate f3700 cfm i 10%; and CHMCoAkf*, After each complete or partial replacement of a charcoal adsorber
/;p p 3 g g bank,byverifyingthatthe/cleanupsystemsatisfiestheinplace Sw(/j*g Penetration and bypass leakage testing acceptance criteria of less thanl M in accordance with ANSI M510-1954 for a halogenated hydrocarbonrefrigeranttestgaswhileopeytingthesystemata flow rate of 37,000 cfm i 10%. ,o
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- e. W% are. us,e.h NUREG 0 52/STPEGS
.Asod r o.nd Arvsr ra sto -MM (or r% -
l 4 ce 9 ve_% o g e % 0.b .-l.5 L A Ge.u. eT et c set, g MPARIS0N
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P:ge No. 300 12/17/85 COMPARISON OF NUREG 0452 REV.5 AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES
------ FSAR CROSS REFERENCE
---Operating Mode---
- 2) OP MODE: 1 23456 OATA 3/4.7.8 STPEGS specific values provided. FSAR 3.9.3.4 DESIGN 3) OP M00E: 1 23456 4.7.8.b "All snutbers* Qualified as 'all FSAR 3.9.3.4 mechanical and hydraulic snubbers' . OP M00E: 1 23456
\...
ELECTRICAL POWER SYSTEMS Unf[ l SURVEILLANCE REQUIREMENTS (Continued) leeds withinh10deconds, energizes the auto-connected I shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the -E5F steady-state voltage and_ frequency of the mpusses 3.0 shall be maintained ati4160p1 teselfvolts anc N 416 l
,f 60gMy09 Hz during this test a Surn luJKcTwv
- 5) Verifying that on3 test signal, without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 M 6 _sinutes. The generator voltage and frequenev shall be -3.0 l
l J4160h!)_ *203 volts ane(607 it:eliHz witnin,410hseconds after thf auto-start signal; the steady-state generator voltage and frequency shall be maintained within these limits during '
l this test; ,
A Sartry Jwnr#
l
- 6) Simulating a loss-of-offsite power in conjunction gwith ": !
h test signal, and:
ESF a) Verifying deenergization of the asungsacyxbusses anc load I sheddingfromtheimasummasygsses; .
t [T D b) Verifying the diesel starts on the auto-start signal,4 l
_ _-,izu: O: _ --.;u y ;;;__ 1i.x , w =;-- . - . . :;- ,, _ : m stes:
i ESF Amen withinM10Fseconds, energizes the auto-connected I
! emoegency!(accident) loads through the load sequencar and operates for greater than or equal to 5 minutes while its generator is loaded with the m % s. Aft h ESF ES enargization, the steady-state volt:ge and frequency of FT S m busses shall be maintained ath a160 F; g ,o 416Ed20) volts and/60h+ lhel[A's during this test; anc c) Verifying that all automatic diesel generator trips, ,$N,(
l except engine overspeed g ens generator cifferential automatically bypassed upon loss of voltage on the ESF esecgency bus concurrent with a Safety Injection Actuation signal.
sianost
- 7) Verifying the, diesel generator operates for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l During the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of this test, the diesel generator shall be loaded to greater than or equal to O .1__. . i ., gc5936 kW l and during the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of this test, the diesel generator shall be loaded to greater than or equal to @:aut
- 5500 r-_-__ .r_.43 kW. The generator voltage and frecuency sha11 5.0 41670Q4160h,1,(idtel volts andg60V! l339jHz withinJ10Aseconds after the start signal; the steacy-state generator voltage and
~. -
W-STS 3/4 8-5 07/15/86 NUREG 0452/STPEGS COMPAD.ISON i
PCQ2 N3. 334 07/15/86 COMPARISON OF NUREG 0452. REV.5. AND STPEGS TECH. SPECS.
NOTE TVPE NOTE a NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ...Coerattrg Moce---
- 3/4.8.1.1 PAGE: 3/4 8- 6.o
- ELECTRICA,L POWER SYSTEMS - A.C. SOURCES. CPERATING DATA 1) 4.8.1.1.2.f STPEGS specific values and FSAR 8.3.1 nomenclature proviced. CP M00E:1 234 --
DESIGN '2) 4.8.1.1.2.f.11 Deleted. Same justification as FSAR 8.3.1
- 4. 8.1.1. 2. c (pg 3/4 8-3). Items 12 to
- 14 CP MODE: 1 234 - -
renumoered.
- 3) 4.8.1.1.2.f.12.a Deleted. Turntng gear engaged is FSAR 8.3.1.1.4.6.
not a lock out feature in tne emergency moce for IEEE 338-1971 RG STPEGS. 1.22. Callaway TS CP M00E: 1 234 --
- 4) 4.8.1.1.2.f.13 Air pressure for diesel generator GDC 17 RG 1.1C8 air starter provtcod. OP M00E: 1 234 --
DESIGN 5) 4.8.1.1.2.f.13 Acd "for eacn start
is to comonstrate 5 consecuttve starts, eacn r.eacntng speed and voltage witntn 10... seconcs.
. . . . . ...... .,o DESIGN 6) 4.8.1.1.2(e)7 Add requirement to comply with SER 8.3.1 SER Section 8.3.1 for STPEGS July 1986 TS Amend.
4 e
e o
t 1
I .
ff Ca-lure. l pac / c o n N w .s egCeecl 'fbe- 2000 l?ou r f h E % cheseI yMr*&S {S95S k& , -fbe$)c'/;ese] emn les l W/ l h@c
- 0d 46 # H1e ,
CTRICAL POWER SYSTEMN z.6our. fd;>,f [/,o g O SURVEILLANCE REQUIREMENTS (Continued)
>J
- 1 frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24-hour test, i l
perfom Specification 4.8.1.1.2e.6)b);" 3 4 -
1-STAMDJY - '
' 8) Verifying that the auto-connected loads to each4diesel -
1 generator do not exceed the 2000-hour rating of gg35 kW; *
{
srm m :
- 9) Verifying the4 diesel generator's capability to: l a) Synchronize with the offsite power source while the__ ggp generator is loaded with its smoogensifficads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Se restored to its standby status, sim esi i i
- 10) Verifying that with the diesel i generator operating in a test I mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator [
to standby operation, . int-(2) automatically energizing the ]
ES emeegensylloads with offsite power; I l
g= : _ ;:-- __; _: ;;y ;::. _..,:.._ _ _.._;; __.. ;; _=;.
7 =3: ,
- :: :: r: m:u:::: ;- : : -- - .___2.- . _ _ ,
l
- 11) 42 % Verifying that the automatic load sequence timer is OPERABLE l with the interval between each load block within + 107, of its design interval; ~
I sTMDBY ENRGENCY MF 12.) 1:1) Verifying 5 that the ?=!!- ::-;3dieselgeneratorAlockoutfeaturedl prevent diesel generator starting;: ', ^1- _,_.
. _] AN D
-) .% ; .- - - = - .---C. I i-) L . .. . . ., . . _ , : . I sT4=cer 13)2*9 Verifying that with alladiesel generator air start receivers pressurized to less thangequal to 240 psig and the _37aupey cosoressors s_ecured, the4 diesel generar.or starts at least sNe#]T M5Ltimes y froa)asetent conditions (and accelerates to at least 690th] rpm in less than or equal toJ101 seconds r.oA KAcH 5rART, f "If Specification 4.8.1.1.2e.6)b) is not satisfactorily completed, it is nol.-sT4xosY necessary to repeat the preceding 24-hour test. Instead, theAdiesel generator may be operated at-G =:_: ____ . _ _ ' q]V k W for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or until operating temper-E ature has stabilized. gg W-STS 3/4 8-6
- 07/15/86 NUREG 0452/STPEGS COMPARIS0N
i Pcg3 N3. 335 12/17/85
[ l' COMPARISON OF NUREG 0452 REV.S. AND STPEGS TECH. SPECS.
l i NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE
(
i
............... ...... .................................................. ... Operating Mode---
l
- - 3/4.8.1.1 PAGE: 3/4 8- 7.0 ELECTRICAL POWER SYSTEMS - A.C. SOURCES. OPERATING OATA 1) 4.8.1.1.2 STPEGS spectfte values and nomenclature F$AR 8.3.1 provided. OP M00E: 1 234 - -
- 2) 4.8.1.1.2.14 Add requirement to demonstrate the bypass and toad reinstatement features of tne lead CP Mc0E: 1 234 - -
sGQuences.
-CESIGN 2) 4.8.1.1.2.e.14 added to provide veriftcation of FSAR 8.3.1 the sequencer reset per the STPEGS destgn. OP M00E: 1 234 - -
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f i
1 i
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$Ir TS j ELECTRICAL POWER SYSTEMS A.C. SOURCES SHUTDOWN - OFFSITE TO ONSITE DISTRIBUTION LIMITING CONDITION FOR OPERATION 3.8.1.4 As a minimum, the circuits from two 13.8kV standby Busses to the associated Auxiliary ESF Transformer and ESF Bus shall be OPERABLE.
APPLICABILITY: MODES 5 and 6 ACTION:
With less than the above minimum required A.C. circuits OPERABLE, innsedi-ately suspend all operations involving CORE ALTERATIONS, positive reac-tivity changes, movement of irrafisted fuel, or crane operation with loads over the spent fuel pool.a'An addition, unen in nuuE 5 with t Reactor Coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, issnediately initiate cor-rective action to restore the required circuit to OPERABLE status as soon I as possible. W,mw 8 Hours,DEPRE150R12E Avc var tw REAc ComavrSystrM Mme A VENT CAPA&ut cP REutV/N6 13to ft M AT SM P.536 SURVEILLANCE REQUIRD(ENTS 4.8.1.4 The above required A.C. circuits shall be de-onstrated OPERABLE at
- least once per 7 dayt by verifying correct breaker alignment and indicated v'oltage on the ZSF busses.
l l
South Texas Project 3/4 8-9 B 01
!:"/15/QQ::'S a e -. FEGS COMPARISON
i Page No. 340 07/15/86 COMPARISON OF NUREG 0452. REv.5. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES
............... ...... FSAR CROSS DEFERENCE
.................................................. ... Operating Mode---
l
- 3/4.8.2.1 PAGE: 3/4 8-10.0 ELECTRICAL power SYSTEMS - D.C. SOURCES. OPERATING l
DATA 1) 3/4.8.2.1 STPEGS specific values provided. FSAR 8.3.2 CP MODE: 1 234 - -
DESIGN
DESIGN
- 3) 3.8.2.1 Ref erence to ' full-capaci ty charger
- FSAR 8.3.2 deleted and addressed separately in added action OP M00E: 1 234 - -
D.
l OES!GN 4) 3.8.2.1 Action O. Added to reflect STPEGS FSAR 8.3.2 design. OP MODE: 1 234 - -
DESIGN 5) 3.8.2.1 - Battery chargers are rated for 300 amps. July 1986 TS Amend.
Since the load on the batteries is greater than 300 amps, more than 1 battery charger is required for operability for Channels I & IV.
W O
S O
ELECTRICAL POWER SYSTEMS 3/4.8.2 0.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a sinimus, the following D.C. electrical sources shall be OPERABLE:
CownI fiA11 asr h
- a. 34tidW125_AI-volt Battery BankAs=:1, and its4 associated ^=: t -_ -_7 2_
charger, and l CnAwa H _A E4DH ME -
- b. jW125J volt Battery Banka10s;=t, and 4*s3 associated ?_:' ,__:b charger.
C,. Cmvult III 125 v0'.7 BArrrey bn G181V J
410 ani AssacsATED CMUtGER.
APPLICABILITY: MODES 1, 2, 3, and 4. d. C,wiutt lT s25.ntxt BarrrRyB4NX(fiCii)
ANO qad Ass 0ClaTED CHARGER.
. ACTION: two a,. With one of the required battery banks _ _ '--- -A' --
_,__! _, Jm inoperable, restore the inoperable battery bank _ .1'u !_ ' __,__:'., r _ - to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
?_!?? -Wi~l%v??'?l??'l-(!]? n " ltl_i" 1! ~;::_-L{
w:L:b:c ;;;_::: : ;_ : --
e SURVEILLANCE REQUIREMENTS f
4.8.2.1 Each(*SC/125 volt battery bank and charger shall be demonstrated CPERABLE:
- a. At least once per 7 days by verifying that:
- 1) The parameters in Table 4.8-2 meet the Category A Ifmits, and [
- 2) The total battery terminal. voltage is greater than or equal to t$be/129 M olts on float charge.
[
/ b. With the required charger on any one train inoperable, demonstrate the OPEPMILITT of the associated battery bank by performing Surveillance Requirement 4.8.2.1.a.1 within eau hour, and at least
(
once per a hours thereafter. If any Category A limit in Table 4.8-2 Se not set, declare the battery inoperable.
07/15/86 pSTS 3/4 8-10 NUREG 0452/STPEGS COMPARIS0N l
l .. - . _ _ _ . _ - . - - _ . _ - . _ - _ - _ _ _ - - . - . . - . -. - _ _ _ _ . . _ . ,
T I
P&g3 N3. 341 12/17/85 COMPARISCN OF NUREG 0452 REV.5 AND STPEGS TECH. SPECS.
NOTE TYPC NOTE s NOTES FSAR CQOSS REFEEENCE
............... ...... .................................................. ... operating Moce--- -
DESIGN 2) 4.8.2.1.c.4 Revtsed to reflect STPEGS cesign. FSAR 8.3.2 OP M00E: 1 234 - -
4
TABLE 4.8-2
. BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A II) CATEGORY B II)
ALLOWABLE I3)
LIMITS FOR EACH LIMITS FOR EACH l PARAMETER DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL Electrolyte >Minimus level >Minism level Above top of Level indication mark, indication mark, plates, and < %" above and < %" above and not maxiom level maximum level overflowing indication mark indication mark lFloatvoltage > 2.13 volts
, > 2.13 volts (6) > 2.07 volts l Not more than 0.020 below the I O -
i iSpecific > 1.200(5)
> 1.195
~
8"*9' "I 'll connected calls
, Gravity (4)
- l. Average of all Average of all I
connectec cells connected cells
> 1.205
> 1.195(5) l TABLE NOTATIONS l
(1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered CPERABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the Category B measurements are taken and found to be within their allowable values, and provided all ' Category A and 8 par.ameter(s) are restored to within limits within the next 6 days.
(2) For any Category B parameter (s) outside the Itait(s) shown, the battery may be considered OPERABLE provided that the Category B parameters are within their allowable values and provided the Category 8 parameter (s) are restored to within limits within 7 days.
(3) Any Category 8 parameter not within its allowable value indicates an inoperable battery.
(4) Corrected for electrolyte temperature and level.
O (5) /. BEOrbatterychargingcurrentislessthanh2[ampswhenoncharge.
(6)f orrected for average electrolyte temperature.
3 3/4 8-12 01/15/86 PSTS NUREG 0452/STPEGS COMPARIS0N
Pcg) No. 343 07/15/86 COMPAR: SON Dr NuREG e452. REv.5. AND STPEGS TECs - SPECS.
NOTE TYpt NOTE e NOTES FSAR CROSS REFERENCE
............ .- ---.-- --------------.-----------....-------------------- ---Operating Moca---
DESIGN 2) 3/4.8.2.2 Reference to ' full-caoacity chargers
- FSAR 8.3.2 deleted and accressac separately in Action b. OP MODE:- - --56 DESIGN 3) 3.8.2.2 ACtton O. Acced to reflect STPEGS cesign. FSAR 8.3.2 CP MODE:- - - - 5 6
- 4) 3.8.2.2 Action a venting capactitty of 1310 gom at 570 psig accec. COMS Analysts. OP MODE:- - - - 56 DESIGN 5) 3.8.2.2 Two Battery chargers are needed to handle July 1986 TS Amend.
the load.
1 1
0 1
l
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ELECTRICAL POWER SYSTEMS .
l D.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION
-tIL e (cmwwrt I en Cua=wEL.DO ame- .
3.8.2.2 As a sinfaum, one3 02fc/125 8sk:k=
- ;- _ t- chargers shall be OPERA 8LE.golt battery bank and Ms, associated :
APPLICABILITY: MODES 5 and 6.
ACTION:
- a. With the required battery bank -------- - - - - - - - - - - - - - - - - - inoperable, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel; initiate corrective action to restore the required battery bank _ i;!'- ,2 ;,- .. . . . to OPERABLE status as soon as possible, and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, depressurize and vent the Reactor Coolant System through a&==3 ;,=: . '= i vents camad C#
Rsusnus 83to sm nr 610 F$nG.
, SURVEILLANCE REQUIREMENTS 4.8,.2.2 The above required (450/1257fvolt battery bank ... ^ .. ...., .
ehessee shall be demonstrated OPERA 8LE in accordance with Specification l 4.8.2.11 I / *
- b. With the required charger (s) inoperable, demonstrate the CPIPMILITY of its . associated battery bank by perfen.ing Surveil-lance Requirement 4.8.2.1.s.1 within one hour, and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. If any Category A limit in Table 4.8-2 is not met, declare the bettery inoperable.
O' W-STS 3/4 8-13 07/15/86 tiUREG 0452/STPEGS COMPARIS0N
P*ge No. 344 07/15/86 COMPARISON OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ..-Operating Moce--- .
'i
- 3/4.8.3.1 PAGE: 3/4 8-14.0 .
ELECTRICAL POWER SYSTEMS - ONSITE POWER O!STRIBUTION - OPERATING DESIGN 1) 3.8.3.1 Delete reference to ' tie breakers Detween FSAR 8.3 j redundant busses * - there are none at STPEGS. OP MODE: 1 234 - -
DESIGN + 2) 3.8.3.1 Revised to reflect STPEGS design and FSAR 8.3 l nomenclature. OP M00E:1 234 - -
ED 3) 3.8.3.1 Renuncered items to incorporate additional data. OP M00E: 1 234 - -
CESIGN 4) 3.8.3.1 Note (*) revised consistent with Catawba FSAR 8.3 Tech Spec. OP M00E: 1 234 - -
CESIGN 4) 3.8.3.1 Note (*) revised to reflect STPEGS FSAR 8.3 design, and consistent with recently approved OP MODE:t 234 - -
Catawea Teen Specs. .
DATA 5) 3.8.3.1.g Revised to reflect STPEGS design. July 1986 TS Amend.
\
1 l
d a
1
M 3/4.8.3 ONSITE POWER OISTRIBUTION ] I OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electric.al bu,sses shall be energized:'- -_^_.. ,_.:^:..
.--mu _
__ ____ . . . ._. .. _ .. m iT~~~~~~~~~~~'~~~~~~~~~"
~ ~ ~ ~ ~ ~ ~
. F.~_~.'_T___;~ ~ D ~ - =
_ X '~_;- _ .
g,, s. 44,$w$osp@t 4 A.C. ~ _ ._ -aBusses consisting of:
y43 1) ~'T4160 - olt ' r.z.,Nus # EiA
, and I
$gg 2) 4807 oltvImseysusy ur Bus # Eje.L(g1A2, e usacrar zoao curx mwsqas
@WQ b. O!-!;N2fA.C.' -- _-._husses consisting of:
{ 1)>f4160%olt ._
hus # E18 , and I
> I}
gh 1* p-.d./"120[ Volt3 A.C. Vita 1 M8[.221 ene@ zed fro a l
I t*" * ""'o$ M 0 C. os gus ag u n ,
C
- u. y { g\ $$ eJ120Nolt A.C. VitalateWEif Dgenergized from its associated M e l
- u. ggg u inverter connected to 0.C. Bus a EiDii",
ossrRssvTibV PANNEL ws ys.f.J120N-n olt A.C. Vita 13Bsi;=W DPi203 energized free its associated yg y inverter connectegeg Bus ggigi,i", gg l j
... 34 dC.c/120]koltA.C.
- Vita 3 18ss:S anergized from its associated l
's inverter connected to 0.C. Bus # 8C118,
'J N" " ETAii
[cCC7L!5]-Volt 0.C. Bus #Eenergized fros,Sattery sanaA and ITs ASSDuATED d t h.
11Dii iTs pssociarro I
15 t . [25Cf125]-volt 0.C. Bus #4 energized from Battery Banx,*F. l APPLICABILIiY: MODES 1, 2, 3, its and 4. 9 125 assoc,arro varsanic, Barrsey D.C.Ano 8us"E1Bi1 surAG V. 12.5-var D.C. Sas *E1C11 susmtzro rw 3CT10N: its Assoc ATs0 BarTEAV san.
TR Ams
- ESr
- a. With one of the required,; - 4_ of A.C. messysucy3 busses not l TKAW ' fully energized, reenergize thel =! - :: within S hours or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. With one A.C. vital bus either not energized from its associated I inverter, or with the inverter not connected to its associated 0.C.
bus: (1) reenergize the A.C. vital bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at l 1 east HOT STANOBY Within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; and (2) reenergize the A.C. vital bus froe its associated inverter connected to its associated 0.C.
bus within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Tsc assotraTro wara our chin Ts "Ime Ainvertersa nay De disconnected from toestr30.C. bus for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,ma l m for the purpose of performing an equalizing charge on tess**Ad aT3 ted g
battery cant providea: (1) their vital buss (es)are energized, and (2) theivital 7 bus s e s ::::; i : L;d -! = d ;t- ; 2:i =- ,- 9 are energized from their l
~
associated inverters anc connected to their associated 0.C. busses.
07/15/86 w-sTs 3/4 8-14 NUREG 0452/STPEGS COMPARIS0N
Page No. 345 12/1T/85 CCMPARISON OF NUREG 0452 REV.5, AND STPEGS TECH. SDECS.
NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ...Operatteg Moca---
- 3/4.8.3.1 PAGE: 3/4 8-15.0 ELECTRICAL POWER SYSTEMS - ONSITE PCWER DISTRIBUTION - CPERATING NONE 1) 3/4.8.3.1 No changes.
OP MODE: 1 234 - -
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ONSITE POWER DISTRIBUTION .
LIMITING CONDITION FOR OPERATION ACTION (Continued)
- c. With one O.C. bus not energized from its' associated battery bank, reenergize the D.C. bus from its associated battery bank within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least NOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be detemined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.
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'O W-STS 3/4 8-15 kRE b52/STPEGS COMPARIS0N
Peg) No. 346 07/15/86 COMPARISON OF NUREG 0452. REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE a NOTES FSAR CROSS REFEREN"E
............... ...... ------...............-------- ..-----------------. ---Operating Moce---
- 3/4,8.3.2 PAGE: 3/4 8-16.0 ELECTRICAL power Sv5TEMS - ONSITE power DISTRIBUTION - Sr4UT00WN DESIGN 1) 3.8.3.2 Item a. Changed to reflect STPEGS oesign. FSAR 8.3 OP MCOE:- - - - 5 6 DESIGN 2) 3.8.3.2 Action Venting capactitty of 1310 gom at FSAR 5.4 570 gom accec. OP MODE - - - - 5 6 DESIGN 3) 3.8.3.2.b - Revised to reflect STPEGS designs. July 1986 TS Amend.
O O
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N kJ O ELECTRICAL POWER SYSTEMS U 'l ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION '
3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:
Erratg TMAw A e< Tasau C E
- a. $-- d - i e i -- ' A.C. unsegencySFbussesconsistingofonsh4160]Jvolt E5F aus bPool4 bPlus pHeneM480Fvoit A.C. e ESF Leam crursas, or bPoog ed '
- b. harm dim.bddow H ars ts @f-m.:.:x1....:E1204 volt A.C. vital henses energized fros3 astr assoc bpstoy 3 t inverter Aconnected to their* respective D.C. bussee, and
'*3 Tue(p,14 C=4ws. I e( Cet.5)
- c. 3... [EEG/125 h olt 0.C. bus energized from its associated battery bank. l APPLICABILITY MODES 5 and 6.
ACTION:
{
f3 V With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel; initiate corrective action to energize the required electrical busses in the specified manner as soon as .
possible, and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, depressurize and vent the RCS through E 'eee^- l l a + = = - -; . . . ' +3 y e n t CAPASA of MLstvwc Ar uAss 13100c+s Ar S70 PSIG. '
SURVEILLANCE REOUIREMENTS 4.8.3.2 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.
~
O N RE 0452/STPEG5 PSTS 3/4 8-16 COMPARISON
Pcg:) No. 347 12/17/85 COMPARISON OF NUREG 0452 REV.S. ANO STPEGS TECH. SPECS.
NOTE TYPE NOTE
- NOTES FSAR CRCSS REFERENCE
............... ...... .................................................. ...Cperating Moce---
i
- 3/4.8.4.1 PAGE: 3/4 8-17.0 ELECT. POWER SYSTEMS - AC CIRCUITS INSIDE PRIMARY CONTAINMENT CESIGN- 1) 3/4.8.4.1 Section Deleted. STPEGS design coes not reQutre tnis control. CP MODE: 1 23---
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I DRPL REFUELING OPERATIONS 3/4.9.5 COPMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct cosununications shall be maintained between the control room and .
. personnel at the refueling station.
APPLICA8ILITY: During CORE ALTERATIONS.
ACTION:
When direct comunications between the control roca and pe'rsonnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS.
I SURVEILLANCE REQUIREMENTS 4.9.5 Direct comunications between the control room and personnel at the refueling station shall be demonstrated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.
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! y-STS 3/4 9-5 01/15/86 NUREG 0452/STFEG5 COMPARISON
Pa3e No. 359 07/15/86 COMPARISON OF NUREG 0452 REV.S. AND STpEGS TECH. SPECS. .
NOTE TVDE NOTE a NOTES
............... FSAR CROSS REFERENCE
...... .................................................. ... Operating Moce---
==
3/4.9.6 PAGE: 3/4 9- 6.0 REFUELING CPERATIONS - MANIPULATOR CRANE DESIGN t) 3/4.9.6 Section renamed ' Refueling Machine'. for FSAR 9,1.4 proper terminology, equipment used is not a crane. OP MODE:. - - - - 6 DESIGN 2) 3/4.9.6 Revised to reflect STPEGS design. Callaway TS FSAR Consistent with Callaway Tech Spec. 9.1.4 OP MODE:- - - - - 6 OTHER 3) 3.9.6 Action. 'The provtstons of 3.0.3 are not appitcante.' Acceo for clartftcation. (Since 3.0.3 OP MODE:- - - - - 6 appites to Moces 1,2,3 and plant must be in Mode 6 for fuel movement in the reactor vessel, it coes not apply here.)
DESIGN 4) 3.9.6.a.3) - STPEGS does not have an automatic July 1986 TS Amend.
load reduction trip; this was inadvertently previously incorporated.
DATA 5) 3.9.6.a.2)a) & b) changed pounds primary and July 1986 TS Amend.
secondary automatic overload cutoff.
J e
i
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RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE [I;-1___ ;;g:--d L; - O ;1;:--J e -!gi ,_ --,-----:-j
- LIMITNG CONDITION FOR OPERATION Gaseous ^^ sis Pao
- sswa 3.11.2.5 The concentration of ';"- ;-- - oxygen in ther,W888:44M80tsvP SYSTEM shall be limited to less than or equal to 4% by volume. .
1 APPLICABILITY: At all times.
ACTION:
GAsto:s WAs7E Pnoessswa
- a. With the concentration of ',f -
,_ - - - oxygen in thef, N SYSTEM exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />,
- b. The provisiens of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREVENTS Gerea ur t Ibessswa
,g 4.11.2.5 The concentration of ';"-- ;-- - oxygen in the ris?E eis =^ E:J /,5YSTEM g g*
0ishallbedeterminedtobewithinthe_abovelimitsbycontinuouslymonitori b SYSTEM with the 7--_-
annmuseegases in tReiWASTE 44t*0tSt*h Specification 3 3 3 4; : =-- oxyge required OPERABLE by Table 3.3-13 c ... .
to idCT l O
p 5TS 3/4 11-15b 01/15/86 NURYG 045?/STPEGS
! COMPARISON
Page No. 390 07/15/86 CCMPARISON OF NUREG 0452 REV.5. AND STPEGS TECH. SPECS.
NOTE Type NOTE a NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ...Cperating Moce...
== 3/4.11.2.6 PAGE: 3/4 11 16.0 RADICACTIVE EFFLUENTS . GAS STORAGE TAN 4S DESIGN 1) 3.11.2.6 STPEGS specific value for neole gas to be FSAR 11.3 activity provided Later. STPEGS nomenclature OP MODE: 1 234 56 provided.
ED 2) 3.11.2.6 Action a. 6.9.1.4 renumbered 6.9.1.7 to reflect STPEGS numeering. OP M00E: 1 234 56 DESIGN 3) 3.11.2.4 RCS Vacuum degassing system acced to Grand Gulf TS FSAR reflect STPEGS coston. 11.3 OP M00E: 1 234 56 DATA 4) 3.11.2.6 Quantity of radioactivity contained in July 1986 TS Amend.
each gas storage tank is provided.
W E
r 9
.. - - _ _ ._. . . - . - .. - ~ . . . - _ . .. . _ _ _ - _ _ _ . . . . - = . - . . - . .-. - ._.
Pag) No. 360 12/17/85 COMPARISON OF NUREG 0452. REV.5. AND STPEGS TECH. SPECS.
j NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE 'I!
............... ...... .................................................. .. 0,er,t,ng . co...
- 3/4.9.7 PAGE: 3/4 9- 7.0 REFUELING OPERATIONS - CRANE TRAVEL - SPENT FUEL STORAGE AREAS a
ED 1) 3/4.9.7 Renamed ' Crane Travel . Fuel Handling Butiding'. OP MODE:1 234 5 6' DESIGN 3) 3/4.9.7 *Except when carrtec by the main hoist of FSAR 9.1.4 the FHB crans' acced to reflect STPEGS assign OP MODE:1 2 3 4 56*
exception to STS.
i 9
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REFUELING OPERATIONS i 3/4.9.6 dMGMfb450>0AM@ REF'UELING MACHINE LIMITING CONDITION FOR OPERATION ArMLlw MatWIMr numaLE PLUEs 3.9.6 The r -- 'W- : n and auxiliary hoist shall be used for movement ofa .
M or fuel assemblies and'shallcow besee OPERA waarnam, ao 8LE with: -
asvusuw Maupt
- a. The3amn6psteessionens used for movement of fuel assesblies having:
- 1) A minimum capacity ofa h pounds, and l Ave **f* wrw mr
- 2) An4 overload cutoff di. _ psmus sa7Ptwrs: _: 1-- ;: ::: _1 l
a) Paman<- arenosmarrw He swas Aant rar amems summars war sea wr l 350 - ce,vomeus anM awwas asset at sugwemp av.srsNors wrteer Abt Off Caetriedt, b.) Secewoont-uss inax A!!6 Naas asce rw %snr ovemoso curveh Me 15'O
- 3) luv na;=: u:.;; =;nm ^i,
, F m a n ,, 6,:- ,, or L.;&r =_; C.; ~r
%s ^=.;= = ; .. .- - ^ ..: u, , m.; .'= W ~ ~'" : ^ i
- == ^ ^*
- .
- -
- .--- my e-mn .
- b. The auxiliary hoist used for latching and unlatching drive rods Amor*A
'_' , Tumur rws HAnams cruanons awaa:
3000
.l
{' 1) A minimum capacity of [423] pounds, and A
Ipoonwne neworan
- 2) '41oad indicator which shall be used to4poemene liftsap loads
_-:::: :' :!?!? ;: -f: FoA per opvaarsovs, APPLICABILITY: During movement of drive rods or fuel assemblies within
- the reactor vessel. l t ACTION:
With the requirements.for crane and/or hoist OPERA 8ILITY not satisfied, suspend
[.useofanyinoperablei_-';i'_: _ . - and/or auxiliary hoist from operations g involving the movement of drive rods and fuel assemblies within the reactor vessel. T c emesiews .a Sarc,marw 3.o.5 mer mr muass.
SURVEILLANCE REQUIREMENTS nernuus usenowc 4.9.6.1 Each3_ O _ ' _1. _. z : used for movement of fuel asseabifes within the reactor vessel shall be demonstrated OPERA 8LE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by perfonsing a load test of at least ($99thysendei
_..__. 125 nacsvr on tw sermemar Anwec ov<nume e,rari 4.,a er camersons au l "E-^;d_
Avwur tow career me.sv rat tono aiutas rwr sermwrs er Seconcarw 19.6.a. 2).
4.9.6.2 Each auxiliary hoist and associated load indicator used for movement .
of drive rods within the reactor vessel shall be demonstrated OPERA 8LE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by performing a load test of at least f6499V pounds. l 12.50 M-STS 3/4 9 6 kRE 52/STPEGS COMPARISON
._--..--..s
, RADI0 ACTIVE EFFLUENTS I'
GAS STORAGE TANKS
~~
LIMITING CONDITION FOR OPERATION l or rw RCS Vv.vp !se.ss.iSorm 3.11.2.6 The quantity of radioactivity contained in each gas storage tanW
, shall be limited to less than or equal to ~ Curies of noble gases (considered as Xe-133 equivalent). - -
4 y APPLICABILITY: At all times. [p. sXIO ACTION:
- a. With the quantity of rad'ioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive saterial to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuaottoSpecification6.9.1.f
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
i SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive saterial contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being added to the tank.
O ri! 3/4 11-16 07/15/86 NUREG 0452/STPEGS COMPARISON
Pag 3 No. 391 12/17/85 COMPARISON OF NUREG 0452 REV.S. AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES
............... --.... ...................-........................--.--- FSAR CROSS REF!RENCE
... Operating Moce...
se 3/4.11.3 PAGE: 3/4 11+17.0 RADICACTIVE EFFLUENTS - SOLID RADIDACTIVE WASTES NONE 1) 3/4.11.3 No changes.
CP MODE:t 234 56 l . ,
i 1
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REACTIVITY CONTROL SYSTEMS 8ASES M00ERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the EC associated with a core .
condition of all rods inserted (most positive W C) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with i temperature at RATED THERMAL POWER co ions./ This value of the M C was then ,
transfoneed into the limiting MTC val us;*IAx 10-' Ak/k/"F. The NTC value ,o off -3.07 T 10 4 Ak/k/'F represents a conservative value (with corrections for l burnup and soluble boron) at a core condition of 300 ppe equilibrium boron concentration and is obtained by making these corrections ,to the limiting NTC value -3:*l,x 10
- Ak/k/*F. l 4.0 4- -
The Surveillance Requirements for measurement of the NTC at the beginning
- and near the end of the fuel cycle are adequate to confire that the MTC remains [
within its limits since this coefficient changes slowly due principally to the i
reduction in RCS boron concentration associated with fuel burnup. .
3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY l
This specification ensures that the reactor will not be made critical 5GI j -
with the Reactor Coolant System average temperature less than4M* F . This
- limitation is required to ensure
- (1) the moderator temperature coefficient i
is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, 9) r_ ^ 12 2-d:- - - -
-_ . i: =ws **F55Ep5*N5i (3) bubble, andC87 the pressurizer is capable of being in an**"P"*"' OPERABLE status with a st the reactor vessel is above its minisua RTE T l
3/4.1.2 B0 RATION SYSTEMS .
The Boron Injection System ensures that negative reactivity control is i available during each mode of facility operation. The components required to l perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (f4) boric acid transfer pumps, #)=usses6eesMest-i ...:..,",rn generators. _, and ( @5)'an
( emergency power supply from OPERA 8LE diesel l , ,
350 With the RCS average temperature above3300'F, a minimum of two baron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN i
i i
W-STS
- 8 3/4 1-2 i 07/15/86 NUREG 0452/STPEGS COMPARISON
REACTIVITY CONTROL SYSTEMS BASES I
80 RATION SYSTEMS (Continued) 8.'IS 7e '.
MARGIN from expected operating conditions Aof h4'Ak/k after xenon decay
- and cooldown to 200*F. The maxima expected boration capability requirement occurs at E0L from full power equ111brita xenon conditions and requires
- M9Wgiga11ons of,d7000Mpa borated water from the sboric acidwp,2.Eh , , ,
$a Mr,710 sm5s d555rtM antATtD WATik k WK R{V$f h ((((
~
RieUnRawrr AWD ss Mosa Tww ADatpuArt NR rnn' NOVIRs0 B0MTNMt CAPAtVurY, With the RCS temperature belowY300*F, one Baron Injection Systes is 350 acceptable without single ra11ure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection Systes becomes inoperable.
The Ifmitation for a maximum of one centrifugal charging pump to be OPERA 8LE and the Surveillance Requirement to verify all charging pumps except 350WFequireo i]PERABLE pump to be Inoperable below3235%'F provides assurance l
that a mass addition pressure transient can be relieved by the operation of a single PORV.
The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN.of 1% ak/k after xenon decay and cooldown frog 200'F to 150'-6/M Mote. This condition requires either 3,3odga11ons oTM7000Ppps borated water .
from the boric acid storage 3 tunas orso,coagallons ofV2000 ppe borated water i from the RWST. sva m 2/ 00
..es The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.
y*g . The limits on contained water volume and boron concentration of the RWST_./0.0 also ensure a pH value of betweenge=N and [449thror the solution recirculated l within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.
The OPERA 8ILITY of one Baron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.
T--
= :i 3ic::i:- ef --i- ---tie- -- u-- 9:o G) :11 -. - - + - ; _-- - e: + ;-
^2;;i:- ^irit: :-- _-i : i .;;, G';^ 1Er ri-i _- 559d-- - ;-H- i: ri 1:4- c, ::
$b.N '/ nib ").?((-d2 . N --_-;.__r__: - Li~': -----
- --.-g".;;
- ;--
~-
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--~
=-
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~
T_'__ _. ,',. '_
._ [~~ ~' . _ C_ . _. _ ', '. _' _ ". ;_ ;. .~. .~~. _ _ ' I '.~_
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._~ _
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._-'.'; ~~.~. . .'
_ .T
- ; I'M
'_ _'_': ~. _ _ .
,;n;:L;;;r :g. .._ m ;n in- __ ;..___ p;;;.; n ::;n'
- ; ;;;,-- :; / , er, c, W-STS B 3/4 1-3 07/15/86 NUREG 0452/STPEGS COMPARISON
DRAFT POWir. DISTRIBUTION LIMITS i
SASE1 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) ,
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
- d. The axial power distribution, expressed in terms of AXIAL FLUX OIFFERENCE, is maintained within the limits.
b taine wi in ts tai p i di on a. th gh 4 //i above,are ai d.
Aynot o F1 re ,R f1 r te nd g y j a A' trade,d off again o a th (1 ., se u R ow te s
! ac j i'fp[e seas g s al o 1 e ure ca ul ed NB)t' tabletAefalo the des gn NBR al I as
- y 11
/
. ti o f ctfin op5NLPOWERadow[c nges n e r di r . fo al pe is (ble tod i sd 1)mit !
' / R as cai'culated 2 ig .2 , a apusedi 5 cif cation ount[l fo'rF 1 s than re al 1 9. h A value u di v iou acc ent
- p analyses where i luences/ parameters tha DN , .g. peak cla hfe damparaturI, nd th$s tbe sadmum "as'maasu [ue 11 '
/
/ uel' ' l
/ /[/ /
rod bowing reduced the value pf D / / //ra o. C t
/d.aval able/
Js to offs [et Ahis recuctio 'in ghe poneric pargip' T 'e gener)'c matgins/ to 9/1% DNBR pompletel of set Ainy , cod how 'nal es/ This margin neludes the
/ /
. e n1 i t, BR 1.30 vs .2 ,
- b. Sfid acing ).cf C0 046 s .059 ,
I p/ /*/ /
- c. Thermat Di us%n C ffi en f 0.0 vs .059]
i
/.
DN,BR Multipi r s'f [0 Iitchr eti , .
0.
/
3, nd
/
I pp le lu of od al sa ref encea n th FaA O
POWEP DISTRIBUTION LIMITS PASES HEAT TLUX HOT CHANNEL FACTOR.-RCS FLOWRATE AND NUCLEAR ENTHALPY RISE NOT CHANNEL FACTOR (Continued)
Tgp v'111 be maintained within its limits provided conditions a through d above are maintained. The cogbination of the RCS flow requirement ($$$$00
$7,4,g36$!as gpm) analysis andvill the be requirement met. on axation FThereiHguaranteeghat of F as a functionthe DNBR used of THERMAL POWERin the safety allows changes in the radial power shape fN all permissible rod insertion 11 nits.
The F requirement of 1.52 includgs a 2% uncertainty in design and a 4%
ugeertainkonthemeasuredvalueofF Therefore, the naasured value of F shouldbeincreasedby4%beforebNn.g compared with the required value of
- l 132.
Sti,/,00 eepee 3
,. The flow requirement $$$$003 3pm already includes a measurement uncertaint
'* of 3.5%. Therefore no adjustment of the measured flow-value is necessary j before comparing assinst the flow requirement. f 10 Fuel rod bowing r offsetthisreductionfhegenericmargin.ucesthevalueoftheDNBratio.[
t The South Texas Project generic margins, totaling 3.3% DNBR, completely offset any rod bow penalties. This margin includes the following:
- 1. Design limit DNBR of_1.30 vs 1.28. g$
- 2. Grid Spacing Esr@ 0.059 vs 0.066. o,ogy
- 3. There,a1 dif fusion coefficient of GHIBSE vs 0.061.
h4er use.M modde.kyp*cerksk. t-)
~= . H e, I e.. g, k skredU. i re,epsined 4 F UlM. 9 h K V. 2. 2. d~
When an nF measurement is taken, an allowance for both experimental error and manufactuYing tolerance must be made. An allowance of 5% is appropriate l for a full-core map taken with the incore detector flux mapping system, and a 31 allowance is appropriate for manufacturing tolerance, i The 12-hour periodic surveillance of indicated RCS flow is sufficient to l detect entyr fle.i degradation which could lead to operation outside the =ee-p*.
==' 1 = ; = ,1 = = i = y= = t 1= = - ? =--- i -- Q- L L: :- THE MQvMMENT5 0F $NcofotA rnod M. 3. 2,. 3 '
l l
l B 3/4 2 4 A 07/15/86 l NUREG 0452/STPEGS COMPARISON
. . ~-
. P-IS w.,c., r-, ha ,, ,.ai,,.o (p.4 ,a w.m seus rw mn ns.m twrnov rax .stre.sv.- P IS n ntswr Mo wes's SmiY ZAW AC.7JA144 C ts ks % g MD Aucws FEES %+ATER liMTM ^*D r [ ,'y Taewc Tu Fw. ww conusera Tg en Hish Fruvarn RM.
INSTRUMENTATION s
8ASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM .
IM5TRUMENTATION (Continued) (,
- o
- The Engineered safety Features Actuation system interlocks perfom the- jj following functions: ;E
, o
, P-4 Reactor tripped - Actuates Turtsine trip, closes main feedwater Ii valves orr T,,, below 5etpoint, prevents the opening of the main g 1 {E E\feedwatervalveswhichwereClosedbyasafetyInjectionorHigh f ;$ $
w i norre.o~ / 5 team Generator water LevelAsignal, allows safety Injection block so
- rc >
. that components can be reset or tripped, no Ar.rvam P-#5. ggl l
Reactor not tripped prevents manual block of Safety Injection. 2If
( E E *'
i P-11 On increasing pressurizer pressure, P-11 automatically reinstates .___L '
- . i Safety Injection actuation on low pressurizer pressur3.j'~lTn decreasing
- pressure, P-11 allows the manual block of Safety Injection actur. tion 1 -
on low pressurizer pressure /on su,ssivt cam.osaw saam,ano issus sfre.
- w.c %*rieg eu n,c us. i.t sraw uw assans nar,, ,
- P-12 On increasing reactor coolant loop temperature, P-12 autcastically - '- '"- '
l .{ - ' : tete- M et;- '='::- ::^ r': r ' : r l 1' 1 --i t; ei " _. ': ': T*vt'ee '__ _^ - _ ' _ _ _ _ _ , _ : provides an i
arming signal to the Steam Dump Systes. On decreasing reactor
~
l
, coolant loop temperature, P-12 C ___ 'J._ __.__! ;i-i ;^ ^;T;i-I--jecti _ ectreti: = if er__ '!--- _=;- _'__ ^ -;i= +;r'_ '_ ':
i...^...7....... .... automatically removes the aming signal free the Steas Dump Systas. .
l
- =t* 0= ' m = C -; ; t _ _ . ;t; --2;r ; _ . _1, 2- 1+ ;;-. r_;--!!- i:',_
_': : _ ._...- _ __1. . :. . . . .
N ;-
l 3/4.3.3 MONITORING INSTRUMENTATION i 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS l The OPERASILITY of the radiation monitoring instrumentation for plant l
operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof resches its i
Setpoint. (2) the specified coincidence logic is maintained, and (3) suffi-l cient redundancy is maintained to permit a channel to be out-of-service for testing or saintenance. The radiation monitors for plant operations sense [
radiation levels in selected plant systems and locations and detemine6whether or not predetermined limits are being exceeded. If they are, the signals are I combined into logic matrices sensitive to combinations indicative of various
' accidents and abnormal conditions. Once the required logic combination is
- completed, the system sends actuation signals to initiate alams or automatic l
isolation action and actuation of Emergency Exhaust or ventilation Systems.
07/15/86 h-STS 8 3/4 3-3 NUREG 0452/STPEGS COMPARIS0N i
.-._____._.---m_,_._..-
t
~ INSTRUMENTATION l BASES 3/4.3.3.2 MOVA8LE INCORE DETECTORS Tne OPERA 8ILITY of the movable incere detectors with the specified minimum .
complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The OPERA 81LITY of this system.is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve. -
For the purpose of measuring F q (Z) or F"g a full incore flux map is used.
Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux saps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable.
l 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capa-bility is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation is consistent with the recomendations of Regulatory Guide 1.12. " Instrumentation for Earth- ]
. quakes," April 1974.
3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that
, sufficient meteorological data are available for estimating potential radiation l doses to the public as a result of routine or accidental release of radioactive l saterials to the atmosphere. This capability is required to evaluate the need l for initiating protective measures to protect the health and safety of the public and is consistant with the recommendations of _,_i;L , -2;id; 1.22, y ,
_=;=-.__.--g.;; .._
_, __. _2 NUREG 0654 Mo *mance W E m mer Gw n..1.23a.s c.W4%d % M vsnR. Tee 3.ut-) . P;5 boer3/4.
B 3 4A ->"
3.5 REMOTE SHUTDOWN SYSTEM f@
l The OPERABILITY of the Remote Shutdown System ensures that sufficient l capability is available to pemit safe shutdown of the facilitydron locations I outside of the control room. This capability is required in the event control I
room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part $} Arptuon h.
"- ^ ^ ~"f2 ! LIT' -- ? U -
- i= ?---i -- h: - ---- ==^-=1 "-- m
= mm W-STS 8 3/4 3-4 07/15/86 NUREG 0452/STPEGS COMPARISON
REACTOR COOLANT SYSTEM BASES
, OPERATIONAL LEAKAGE (Continued)
- . .. _ _ . _ . , _ _ . . . - . .__. --- _-_ - m - . -
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pg Thes.gpaleakage1 any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure.
It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in tta pair can go undetected for a .
substantial length of time, verification of valv'e integrity is required.
Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containmente these valves should be tested periodically to ensure 1o1 probability of gross failure.
The Surveillance Requircments for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure
' isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
3/4.4.7 CHEMISTRY ,
The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion l protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion
, studies show tnat operation say be continued with centaminant concentration levels in excess of .he Steady-State Limits, up to the Transient Limits, for j the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The ties interval pemitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concen-trations to within the Steady-State Limits. l The Surveillance Requirements provide adequate assurance that concentrations l in excess of the if aits will be detected in sufficient time to take corrective action.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reacter coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY wi11 nct exceed an ,
W-ST5 8 3/4 4-5 07/15/86 NUREG 0452/STPEGS COMPARISON
REACTOR COOLANT SYSTEM 8ASES -
l l
SPECIFIC ACTIVITY (Continued) i appropriately small fraction of 10 CFR Part 100 dose guideline values following.
- a steam generator tube rupture accident in conjunction with an assumed stea G-state reactor-to-secondary steen generator leakage rate of 1 spa. The values i for the limits on specific activity represent Itaits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative ,
l in that specific site parameters of the STPEG5 site, such as SITE SOUNDARY location and meteorological conditions, were not considered in this evaluation.
)
The ACTION statement permitting POWER OPERATION to continue for limited time pertods with the reactor coolant's specific activity greater than 1 aicrocurie/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine sp.iking phenomenon which may occur following changes in THERMAL POWER.
levels exceeding 1 alcrocurie/graa DOSE EQUIVALENT I-131 but within theOperation .
limits shown on Figure 3.4-1 aust be restricted to no more than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year (approximately 10% of the unit's yearly operating time) since the activity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at .
the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture.
The sample analysis for determining the gross specific activity and I can '-
l, exclude the radiciodines because of the low reactor coolant limit of 1 aferocurfe/
graa DOSE EQUIVALENT I-131, and becauss, if the limit is exceeded, the radiciodine level is to be determined every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the gross specific activity level and radiofodine level in the reactor coolant were at their limits, the radiciodine contribution would be approximately 1%. In a release of reactor coolant with a typical atxture of radioactivity, the actual radio -
iodine contribution would probably be about 205. The exclusion of radio-nuclides with half-lives less than 10 minutes from these determinattuns has been made for several reasons. The first consideration is the difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its release to the environment and transport to the SITE SOUNCARY, which is relatable to at least 30 minutes decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radionuclides in tne typical reactor coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life of 10 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay l to very low levels before they could be transported from the reactor coolant l to the SITE 80UNDARY under any accident condition.
PSTS 8 3/4 4-6 07/15/86 NUREG 0452/STPEGS COMPARIS0N
REACTOR COOLANT SYSTEM O =>"
i l
SPECIFIC ACTIVITY (Continued)
Based upon the above considerations for excluding certain radionuclides ,
from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to perfore the sampling, transport the sample, and perfore the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample'and countar having reproducible beta or gamma self-shielding properties.. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides.' The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about 1 day, about I week, and about 1 month.
Reducing T to less than 500*F prevents the release of activity should asteamgeneratE9 tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves.
The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSURE / TEMPERATURE LIMITS l The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:
l l
- 1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
- a. Allowable combinations of pressure and temperature for specific l temperature change rates are below and to the right of the lief t lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and
- b. figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant l characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
l l -
O B 3/4 4-7 PSTS 07/15/86 l NUREG 0452/STPEGS l COMPARISON 1
1
BASES PRESSURE /TEMPERATlJRE LIMITS (Continued)
- 2. These limit lines shall be calculated periodically using methods provided below. -
- 3. The secondary side of the steam generator must not be pressurized above -
200 psig if the temperature of the steam generator is below 70*F,
- 4. The pressurizer heatup and'cooldown rates shall not exceed 100*F/h and l 200'F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than
, (all $3 06854*F, and l
- 5. System preservice hydrotests and inservice leak and hydrotests shall be perfomed at pressures in accordance with the requirements of ASME Boiler l and Pressure Vessel Code,Section XI. .
The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-73, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Susumer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the .
calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975.
Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the.end of N)ffective full power years (EFPY) of servi'ce life. The k hFPY service l life period is chesen suc.h that the limiting RT h0T at tne 1/4T location in -
the core region is greater than the RTHOT of the limiting unirradiated material.
l The selection of such a limiting RTNOT assures that all components in the l Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. .
The reactor vessel materials have been tested to determine their initial RTNOT; the results of these tests are shown in Table 8 3/4.4-1. Reactor opera-tion and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTNOT. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART NDT computed by either Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials,"
i or the Westinghouse Copper Trend curves shown in Figure B 3/4.4-2. The heatup l
and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjust N ments for this shift in RTNOT attheendofCtSlgFPYaswellasaojustments i i
for possible errors in the pressure and temperature sensing instruments. ~
W-STS
- B 3/4 4-8 07/15/86 NUREG 0452/STPEGS COMPARISON
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' 07/15/86 NUREG 0452/STPEGS COMPARISON
REACTOR COOLANT SYSTEM O BASES PRESSURE / TEMPERATURE LIMITS (Continued) , , Values of ART NDT determined in this manner may be used until the results . from the material surveillance program, evaluated according to ASTM E185, are
. available. Capsules will be removed in accoreance with the requirements of i ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen with-
. drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-tionship between the fast neutron flux density at the location of the. capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART NOT determined from the surveillance capsule exceeds the calculated ART NDT for the equivalent capsule radiation exposure.
Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived free Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix G . to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A. The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures t semielliptical surface defect ! with a depth of one quarter of the wall thickness T, and a length of 3/2T , l 1s assumed to exist at the inside of the vessel wall as well as at the l outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference , crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNOT' I8 "'*d and this includes the radiation-induced shift, ARTNOT' "" " "di"8
- l the end of the period for which heatup and cooldown curves are generated.
The ASME approach for calculating the allowable limit curves for various I heatup and cooldown rates specifies that the total stress intensity factor. l Kg , for the combined thermal ar.d pressure stresses at any time during heatup l or cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time. K IR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The KIR curve is given by the equation: O W-STS B 3/4 4-12 07/15/86-- NUREG 0452/STPEGS COMPARISON
P REACTOR COOLANT SYSTEM BASES
?RESSURE/ TEMPERATURE LIMITS (Continued) ,
Kgg = 26.78 + 1.223 exp [0.0145(T-RTET + 160)] (1) Where: K gg is the reference stress intensity factor as a functicn of the metal . temperature T and the metal nil-ductility reference temperature RTET. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: CKgg + kit i "IR (2) Where: Kgg = the stress intensity factor caused by membrane (pressure) stress, ! Kgg = the stress intensity factor caused by the themal gradients, l Kgg = constant provided by the Code as a function of temperature l relative to the RT NOT of the saterial, . C = 2.0 for level A and 8 service limits, and ! C = 1.5 for inservice hydrostatic and leak test operations. l At any time during the heatup or cooldown *c ansient, K gg is determined by the metal temperature at the tip of the 'postula; id flaw, the appropriate value for RTNOT, and the reference fracture toughness curve. The thermal stresses l resulting from tasoerature gradients through the vessel wall are calculated j and then the corresponding themal stress intensity factor, KIT, for the I reference flaw is computed. factors allotmed pow are obtained and, from these,
- nth % coins 4% k theareallowable ressures calculated. p& on%4 COOLDOWN " ar vess.) %,igr'al te m 4; g, p n d b.se..c, swvAbme. sg.. Wens p.,b,r4 8"WR3b, Appendw % et Lg acsaedect MSt. Shah,e. \ k ~k % L y,y. wl taY For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to axist at the inside of the vessel wall. During cocidown, the controlling location of the flaw is always at the inside of the wall because the themal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rata of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the V4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not trut for the steady-state situa-tion. It follows that at any given reactor coolant temperature, the AT l, developed during cooldown results in a higher value of Kgg at the U4T location e W-STS
- B 3/4 4-13 07/15/86 .
NUREG 0452/STPEGS COMPARISON
i law TemPeRATuftE oveMassautE WeM (TunoT (6'b/94-lig}
... . . . . . . , . . . . . ...... - sen The OPERABILITY of two PORV's, or an RCS vent opening capable of relieving at least 1310 gpm at an RCS pressure of 570 PSIG ensures that the RCS will be protected from pressure transients which could e.xceed the limits of 10 CFR Part 50 Appendix G, when one or more of the RCS cold legs are less than or equal to 350' F. Either PORV has adequate relieving capacity to protect the RCS from over pressurization when th'e transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50' F above the RCS cold leg temperatures, or (2) the maximum credible mass injection flow rate due to the startup of a single HHSI pump plus 100 gpm net charging flow, while the RCS is in a water solid condition and the RCS temperature is between 350' F and 200* F.
For RCS temperatures less than 200* F, the maximum overpressure event consists of operating a centrifugal charging pump with complete termination of letdown and a failure of the charging flow control valve to the full flow condition. O : G { Q (9t g 3/y y./S/) 07/15/86 NUREG 0452/STPEGS COMPARISON
, .s, REACTOR COOLANT SYSTEM .1 l BASES 3/4.4.10 STRUCTURAL INTEGRITY -
The inservice inspection and testing programs for ASME Code Class 1, 2, - and 3 components ensure that the structural integrity and operational readiness l of these componenta will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1). Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, J,q13 Edition and Addenda through 1975. . j 3/4.4.11 REACTOR COOLANT SYSTEM VENTS vcsst mAc Reactor 4 :..: ...'. :, . '_ - vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural c ,3 circulation core cooling. The OPERA 8ILITY of 4esspeaufReactor T-;E._c I ' 33 , staojsvent paths" ; ^>= D,- t;5 _:;;! _ j]; 4- Q _ _ '; ,C;;1 ==t Igr{ f -
,....,, ... a.......... . . . - . , . . . . , , . . . . - . . , , . . . . . . . . . . . . . . _ _ . .,
-;--:---i] ensures that the capability exists to perform this function. ,
vesset.nEAD The valve redundancy of the Reactor (- ' , vent paths serves to i minimize the probability of inadvertant or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the v' ent path, ye.u. stro The function, canabilities, and testing requirements cf the Reactor atestans
+fetosi vents are consistant with the requirements of Item II.B.1 of NUREG-0727,-
" Clarification of TMI Action Plant Requirements," November ISSO.
PREssunizin PORVb .senvr As vrurs to rue facssuszrx w Akt cosento By SmeiM ArtoM 3/4.4.M, l W-STS 8 3/4 4-16 07/15/86 NUREG 0452/STPEGS COMPARISON
3/4.5 EMERGENCY CORE COOLING SYSTEMS Nik LI g,
.% 8 O au BASES Tjg
'. EfE 3/4.5.1 ACCUMULATORS Es~
sti The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures Jt=
,w.e, that a sufficient volume of borated water will be immediately forcad into the *ae reactor core through -- ' er the Acold legs in the event the ACS pressure falls below the pressure of the accumulators. This initial surge of water into the 4lE g - E.
core provides the initial cooling mechanism during large RCS pipe ruptures. yu 7u .=. The limits on accumulator volume, boron concentration and pressure ensure sT 5 that the assumptions used for accumulator injection in the safety analysis are T E',j met. t'u The accumulator power operated isolation valves are considered to be IET g a g:
" operating bypasses" in the context of IEEE Std. 279-1971, which requires that ugw bypasses of a protective function be removed automatically whenever permissive *-w conditions are not met. In addition, as these accumulator isolation valves O2%
fail to meet single failure criteria, removal of power to the valves is required.j "
- e ..
The limits for operation with an accumulator inoperable for any reason *2f except an isolation valve closed minimizes the time exposure of the plant to o". .
. a LOCA event occurring concurrent with failure of an additional accumulator * ; ,g $
which may result in unacceptable peak cladding temperatures. If a closed ig*g isolation valve cannot be immediately opened, the full capability of one g 3, accumulator is not available and prompt action is required to place the g gg reactor in a mode where this capability is not required.
- e. g , ,
.- . .c 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS mE*
m et IEIo The OPERABILITY cf twe4 1 ndependent ECCS subsystems ensures that sufficient . u emergency core cooling capability will be available in the event of a LOCA Et I ' assuming the loss of one subsystem through any single failure consideratien. E * ~E***use), subsystem, operating in conjunction with the accumulators,is capable of me"" 8**% supplying sufficient core cooling to limit the peak cladding temperatures '
"2125 within acceptable limits for all postulated break sizes rangina from the
- E .g double ended break of the largest RCS cold leg pipe downward.[In addition, c**;
- each ECCS subsystem provides long-tars core cooling capability in the recirculation mode during the accident recovery period.
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O y-STS B 3/4 5-1 07/15/86 NUREG 0452/STPEGS COMPARISON
/
s/v.s.2 and 3/V. S.3. ECCS, SMdw (.caur.) When the RCS temperature is bel $w 350*F, the ECCS requirements are balanced between the limitations imposed by the low temperature overpressure protection, and the requirements necessary to mitigate the consequences of a LOCA below 350*F. At these temperatures, single failure considerations are not required because of the stable reactivity condition of the reactor and the limited core cooling requirements. Only a single Low Head Safety Injection 8 pump is required to mitigate the effects of a large break LOCA in this mode. However two are provided to accommodate the possibility that the break occurs in a loop containing one of the Low Head pumps. For a small break LOCA in MODE 4, a single Low Head Safety Injection pump is sufficient if a High Head Safety Injection pump can be OPERABLE in 30 minutes. Thus a second High Head Safety Injection pump is required OPERABLE in 30 minutes to accomodate the possibility that the initial High Head Safety Injection pump is located in the same RCS loop as the break. This configuration provides assurance that a mass addition transient will involve only one High Head Safety Injection pump, and can be relieved by the operation of a single PORV. Low Head Safety Injection pumps are not required inoperable below 350'F because their shutoff head is too low to impact the low temperature overpressure protection limits. Below 200*F (MODE 5) no ECCS pumps are required, so the High Head Safety Injection pumps are locked out to prevent colo over pressure. W h E 07/15/86 8 3/9 S- M NUREG 0452/STPEGS COMPARISON
. j "N "
CONTAINMENT SYSTEMS BASES g Yg CONTAIMENT SPRAY SYSTEM (Continued)
$ {C . '. t ^ ' . ' : ' - d' n n : . ;!)
E The Containment Spray Systes and the containment Cooling System are redundant l [ to each other in providing post-accident cooling of the containment atmosphere.
- s. However, the Containment spray Systte also provides a mechanism for removing, - I
)o. iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERA 8LE status have been maintained l
=h consistant with that assigned other inoperable ESF equipment.
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.;. g s t 3/4.6.2.2 SPRAY ADDITIVE SYSTEM 10PM9 met MME3 (
{n The OPERA 81LITY of the Spray Additive System and:ttw ^ -ri:.; by = ^in 1Etik ensures that sufficient Ne0H is added to the containment spray luut l d{t -: ^ '. rc. __,, in the event of a LOCA, to guarantee a pH value or between 1
,f 6.5 and 10.0 for the solution recirculated within containment after a LOCA. i l
-Ayn the evolution of iodine and minimizes the effect of chloride and caustic l gD stress corrosion on sechanical systems and components. The contained solution w
volume limit includes an allowance for solution not usable because of tank 5 , discharge line location or other physical characteristics. These assumptions 1 are consistent with the icdine removal efficiency assumed in the safety analyses. l g ,,. 3/4.6.2.3 CONTAINMENT COOLING SYSTEM (9M99mt$ j E y$ The OPERABILITY of the Containment Cooling Systes ensures that: (1) the ! t g } containment air temperature will be maintained within limits during normal .
$- E operation, and (2) adequate heat removal capacity is available when operated [
t%g in conjunction with the Containment Spray Systems during post-LOCA conditions. E "it ini
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The Containment Cooling Systas and the Containment Spray System are redun-dant to each other in providing post-accident cooling of the containment atmos-I h phere. As a result of this redundancy in cooling capability, the allowable
$Th out-of-service time requirements for the Containment Cooling System have been l 2&L appropriately adjusted. However, the allowable out-of-service time requirements in C 8 for the Containment Spray System have been maintained consistent with that
$ assigned other inoperable ESF equipment since the Containment Spray System also
{Lh,providesamechanismforremovingiodinefromthecontainmentatmosphere. v) p ... s... . . _ .
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x-_- : - - - l 3 . 3/4.6.( CONTAIl#9ENT ISOLATION VALVES The OPERA 81LITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside ehvironment in the event of a release of radioactive material to the containment atmosphere or
- pressurization of the containment and is consistent with the requirements of General Design Criteria 54 through 57 of Appendix A to 10 CFR Part 50. Contain-ment isolation within the time Ifmits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent witn the assumptions used in the analyses '
for a LOCA. 4 3/4.6. COM80STI8LE GAS CONTROL The OPERA 8ILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit :.. 2 _ ^ _ . ,_ - l j tystem) is capable of controlling the expected hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, ^^ and (3) corrosion of metals within co,ntainment. v..,.
,..... .. ... . . . . . . . -._____=....__,.;-..-._---- . . . . - - _
- r ^ ^ C ' _. .:. These Hydrogen Control Systems are consistent with the recoe-sendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA," 4emreMeit Samasea 1976.
. . .. n
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e PATHOSPHERIC 8 3/4 6-5A 07/15/86 NUREG 0452/STPEGS COMPARISON
;-- ,1.- .
E - - - Q 1 O 3/4.9 REFUELING CPERATIONS BASES 3/4.9.1 BORON CONCENTRATION . The limitations on reactivity conditions during REFUELING ensure that: (1) the reactor will remain suberitical during CORE ALTERATIONS, and (2) a uniform baron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assmed for the boron dilution incident in the safety analyses. The value of 0.95 or less for K,ff includes a 1 Similarly, the boron hc% Ak/k conservative .a_llowance for uncertair; ties.oncentration value offleMS ppa or greater allowance of 50 ppm boron. The locking closed of the required valves during refueling operations precludes the possibility of uncantrolled boron dilution of the filled portion of the RCS. This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water. 3/4.9.2 INSTRUMENTATION . The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. 3/4.9.3 DECAY TIME The minie.un requirement for reactor suberiticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay cf the short-lived fission products. This decay time is consistent with the assumptions used in the safety analyses gror TNr amo Atructiam assics. l 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. TMc OPERA 81uTY OF THe co=merMr wariterieu 1.wsracv SYsrum erwes ror rne covrAMENT hige pewtMTrods m gr MrwetAncAuy asuntfD ecN M& RActaTICW Lt6M5 MfM/M cCNTAIMWNro T'Mito SPEC $lCAT*V t*WAS rNT CONrAtynyWr PORg 3/a.9.5 COMMUNICATIONS The requirement for comunications capability ensures that refueling O station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS. y-STS 3 3/4 9-1 R 52/STPEGS COMPARISON
REFUELING OPERATIONS BASES REFWLING MACHINE 3/4. 9. 6 A- . . . ArnetwdMAcnurko AuxuinRf msT Emsw.c The OPERABILITY requirements for theA; ." . ha- t- __ ensure that: - MAuwAwo AcudW (1)3==a4 7 aistucoseums will be used for movement of_ drive rods and fuel assem-[M Af**- est blies, (2) hihas assembly,.and (3) sufficient iona capacity the core internals and reactorto vessel 11rt aare arive roc orfrom protected ruel(M4ml#8 excessive lifting force in the event they are inadvertently engaged during lifting operations. FUSL HANDLlNG 15UILDING (FHD) l 3/4.9.7 CRANE TRAVEL 4980 p=M G M 90RA00=WIRERP i watas nwa at rye mm wsr or rnt FHS 15/2 m enut,
- s The restriction on movement of loadsgin excess of the nominal weight of a l
3[I'T) fuel and control _ rod assembly and associated handling tool over other fuel ! asseenlies in thefstomoys-pool ensures that in the event this load is dropped: t l (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses. TNi ewsmsr on rar FRS is/z chr u4s aars Dasecuro Tc strr sMTrury ReuvroRY Gmor I.104,'OmewAo Cow HAwauxo .5 n rents n n M u cA R Men Rant.s, FEBRUMY, I97(o, 3/4.9.8 RESIDUAL HEAT RU40 VAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is availaole to remove g decay heat and maintain the water in the reactor vessel t:elow 1309ffas required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained * ' N6 through the core to minimize the effect of a boron dilution incident and prevent boron stratification. The requirement to have two RHR loops OPERABLE when there is less than - 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RNR loop will not. result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is avail-l able for core ecoling. Thus, in the event of a failure of the operating l RHR loop, adequate time is provided to initiata emergency procedures to cool the core. l l u- .: =--: --
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0 RARIL f is sy ns es tha th c nm t en an l l [e e tr4(io s 11 e ut t al y cla ed o d ec io of i rd ti n ljhe w th t nta t. Th OP LI t is ys a s l e i d 36 r st the ct hviren as of.ra cac ve ma riIfro th co tai t ma,chefe nt. y-STS B 3/4 9-2 07/15/86 NUREG 0452/STPEGS COMPARISON .
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SECTION 5.0 DESIGN FEATURES d 1 i O 4 01/15/86 NUREG 0452/STPEGS COMPARISON
Fage No. 415 07/15/86 COMPARISON OF NUREG 0452. REV.5 AND STPEGS TECH. SPECS. NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE
.....--........ --- -- ...............--.........--..--..... ....-.-..- - ---Operating Mooe-.- ** 5.0 PAGE: 5.0 5- 1.0 DESIGN FEATURES DATA 1) 5.1 Site. 5.2 Containment. STPEGS spectfic values 10CFR100.3(a).
o.ovided (brackets removed). 10CFR50.36a FSAR 1.1 1.2 OP MODE:t 234 56 ED 2) 5.2.1 and 5.2.2 ' containment butiding* qualifted as ' reactor containment butiding'. OP MODE: 1 23456 ED 3) 5.2.2 Changed to read "The reactor contatnment building is designed for a minimum internal OP MODE:1 23456 pressure of 56.5 psig and an accident temperature of 286 degrees F.* for clartftcationn and STPEGS design. DATA 4) 5.2.1.g Changed net free volume July 1986 TS Amend. e d!!!!
~
DRAFT ; 5.0 DESIGN FEATURES 5.1 SITE ,
~
EXCLUSION AREA 5.1.1 The Exclusion Area shall be as shown in Figure,.(5.1-I I LOW POPULATION ZONE 5.1.2 TheLowPopulationZoneshallbeasshowninFigureJ5.1-2[ l MAP DEFINING UNRESTRICTED AREAS AND SITE SOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENT 5 . 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as defini-tion of UNRESTRICTED AREAS within the SITE 80UNDARY that are accessibit to MEMBERSOFTHEPUBLIC,shallbeasshowninFigures)5.1-3and5.1-4F. The definition of UNRESTRICTED AREA used in implementing these Technical Speci-fications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE 80UNDARY, is utilized in the Limiting Conditions for Operation to keep levels O of radioactive materials in liquid and gaseous effluents as low as is reason-ably achievable, pursuant to 10 CFR 50.36a. 5.2 CONTAINNENT l CONFIGURATION 5.2.1 The"[$$$ainmentbuildingisasteel-lined,reinforcedconcretebuilding of cylincrical shape, wit,hgvdog
, roof and having the following design features:
- a. Nominal inside diameter = 150 feet.
- b. Nominal inside height stditsst:
- c. Minimum thickness of concrete walls = 4 feet.
- d. Minimum thickness of concrete roof = 1 feet.
- e. Minimum thickness of concrete floor pad = 1 feet.
- f. Nominal thickness of steel liner = 3/6 inches.
- g. Net free volume =3.58h[ cubic feet.
DESIGN PRESSURE AND TEMPERATURE
- 5. 2. 2 The785Eainmentbuildingisdesigned...,.. . _- .- for a maximum internal pressure of 5s.5 psig and eVtemperature of gg*F.
AN ACCIDulT W-STS 5-1 07/15/86 NUREG 0452/STPEGS COMPARISON
Page No. 416 07/15/86 COMPARISON OF NUREG 04'52. REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE
- NOTES FSAR CROSS REFERENCE
............... ...... .................................................. ...Opaeattng Moce--- ,
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a ^ s .. O g I Page No. 418 i l 07/15/86 COMPARISON OF NUREG 0452. REV.5 AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE , ............... ---... ............................--............--...... ..-Operatteg Moce--- * ** 5.0 PAGE: 5.0 5- 4.0 ! DESIGN FEATURES l l \ I l DATA 1) Figure provided. July 1986 TS Amend. 9 l l l l 1 l i i I m O 5 \\ 'h s UM' (,b R e.tLROAD SPUR , - c ^ i blk7 5 O' N .O - N - " " " . . \ * [. ,4 g ' C' pb 400' WIDF TR AN g(Ss10( ( ~- " R . i M(Ns. N .- s LINE R/W-* TO W A.PARISM, OUNTRt HOLNA '% N.s * \s s ' SKYLINE,'W4 LON HitL g s Q 'N' RESTRICTED. AREA & cN i j1 t. T "SMISSION LINE i, J , ^ # 'ngM'$ttMig9 EXCLUSION AREA / ),,g ~**** ' ' / ~ >) -RELOdATED FM SN .y , ^ D I$li488Illid e / , god WIDE TRANSMISSION LINEm . / \ R\W TO VELASCO I ESSENTIA( COOLING POND f gg EOROLOdiCAL OWER [/. ' _k ' SWITCH f ": . a :-_ == , ~- g- a ; YARD - I \" 4 PLANT .s , Y AAbioACTIVE l_. hED ~ ' ' "=?q [.C e ? h-h ~ 0M & !s a.km a hr es 6 n p- ~ ' + e = Sc 7W%gigOgf : ~ -J p - .. 7 p p Q y %'-x- K V? swTeo m.-'y . ~ . I ROBBINS p ,, - -. y ?K s, . jStOuss y _ = F'.. x ,, M * *: \ a- / ql , \' -. - =' % / - N = ,,. g HEAVT..IAUL ,RO, nD y c 0- 0rE -~~- L P "L %N CIRCULATION '6 x,c.u.. , : an1 SOUTH TEXAS PROJECT w / 2 i l . , . 4 i d\ ,) i ( CHANNEL 2 . - AKE UP UNITS 1 & 2 ' \ PIPELINE Figure 5.1-3 E R [ S E N' V O / R Restricted Area and o e -- - / ' , . . . ~ . Site Boundary for R*' j [. E { e, . , , g l, . , Radioactive Gascous Effluents ui .' n \\ g, m t , Page No. 419 07/15/86 COMPARISON OF NUREG 0452. REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE --------------- ------ -------------------------------------------------- ---Operating Moce--- ** 5.0 PAGE: 5.0 5- 5.0 DESIGN FEATURES DATA 1) Figure provided. July 1986 TS Amend. t l l I l l l b = O l ' (,/ RAILROAD d 7 I / f> C '% ' 400' WIDE TR AN 3510, /+ L . . ' * \ s'N s LINE R/"W' TO W A. PARISH,@ttCOUNTRY, HOLNA D * ,i, O s' s SKYLINEAN LON HILL RESTRICTED AREA & N \ .' 'ngM'o ett I"^"SMISSI N LINE iNG EXCLUSION AREA . 'd , " ~~~^b% k, ' ' * * * *
- _ ____ _ _ - = - - - -
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g- UNITS 1 & 2 L.M AKE" UP ( i . PIPELINE Figure 5.1-4 , s R , E S E RU 15, O / R -Restricted Area and S V'* - lI / 's f '.,, , . - 't i **,. ' 'I ,' Site Boundary for Radioactive Liquid N l f-) ~ g Effluents g ,; . .', \\ ll . e . Page No. 420 12/17/85 COMPARISON OF NUREG 0452 REV.5 AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES FSAR CRCSS REFERENCE ............... ...... .................................................. ... Operating Mode--- ** 5.0 PAGE: 5.0 5- 6.0 DESIGN FEATURES DATA 1) 5.3. 5.4 and 5.5.1 STPEGS specific values FSAR 4.2. CH 5 provided (Drackets removed 5.4.1 5.5.1). 2.3.3 CP MCOE:t 23456 DESIGN 2) 5.4.2 Reactor Coolant System nominal T(avg) to ce FSAR CH 15 . provided LATER. OP M00E: 123456 DESIGN 3) 5.3.2 The absorber material is Hafntum STEGS FSAR 4.2.1.6 design. OP MODE:t 234 56 _=. m r-. - DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with: ,, ,
- a. A k,ff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of f.63% Ak/k for uncertainties as described in Section/4.3[
3 of the FSAR, gd
- b. A nominal l$bgrinch center-to-center distance between fuel l assemblies placed in the storage racks.
5.6.1.2 The k,ff for new fuel for the first core loading stored dry in the spentfuelstorageracksshallnotexceed/0.98[whenaqueousfoasmoderation is assumed. , I DRAINAGE 5.6.2 Thespentfuelstoragepoolisdesignedandshallbemajntainedto prevent inadvertent draining of the pool below elevation 62T CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 12J fuel assemblies.- 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. l l l l l W-STS 5-7 01/15/86 NUREG 0452/STPEGS COMPARISON pcge No. 422 07/15/86 COMPARISON OF NUREG 0452. REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE ............... ...... .................................................. ... Operating Mode--- . ** 5.0 PAGE: 5.0 5- 8.0 DESIGN FEATURES DATA 1) Table 5.7-1 Component Cyclic or Transtant Limits. STPEGS specific values provided. OP MODE: 1 23456 ED 2) Table 5.7-1 Component Cyclic or Transtant Limits. ' Secondary Coolant System
- changed to 'Seconcary CP M00E: 1 23456 System
- for clartty.
DATA 3) Table 5.7-1 Changed cyclic and design cycle July 1986 TS Amend. for auxiliary pressurizer spray. t I l t i e e I l I TABLE 5.7-1 tat h w COMPONENT CYCLIC OR TRANSIENT LIMITS . j CYCLIC OR DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT 2D0 Reactor Coolant System [f80Lheatup cycles at < 100*F/h Heatup cycle - T"'8 from 1 230*F l 2M ahd;tfS0] cooldown cycles at to > 550*F. l i < 100'F/h. Cooldown cycle - T*'8 from I j > 550*F to < 200*F
- 71M
[$90hpressurizer cooldown cycles Pressurizer cooldown cycle l l at i 200*F/h. temperatures from > 650*F to ~ j $ 200*F. ,
- 80 .
$100]Al oss of load cycles, without > 15% of RATED THERMAL POWER to l immediate Turbine or Reactor trip., 'dX of RATED THERMAL POWER. lo # lS&] cycles of loss-of-offsite Loss-of-offsite A.C. electrical l A.C.4 electrical power. ESF Electrical System. ycles of loss of flow in one Loss of only one reactor l reactor coolant loop. coolant pump. I N j [S0033 R eactor trip cycles. 100E to OE of RATED THERMAL POWER. l
- \O gg:InnovERTent PREssum2ER
} Et4]3 aux 111aryaspray praywatertemperaturedifferential actuation cycles. p " ^
- 7. 4. 4 2.l
- F.
[$e] ak tests. Pressurizedto>D2485[psig. l [5] ydrostatic pressure tests. Pressurizedto>*T3100[psig. l 0 { 9E3 Secondary tuotant System fldteamlinebreak. Break in a > 6-inch steam line. l . wa m I i ) @$ fij3 ydrostatic pressure tests. Pressurized to > [1358]3ps g. I D 4 . g . w . RET O . l t l t l ' SECTICF 6.0 - ADMINISTRATIVE CONTROLS l l l 01/15/86 NUREG 0452/STPEGS COMPARIS0N l Pcge No. 423 07/15/86 COMPARISON OF NUREG 0452 REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE ............... ...... .................................................. ... Operating Mode--- 4. . ** 6.0 PAGE: 6.0 6- 1.0 ADMINISTRATIVE CONTROLS DATA 1) 6.1.1 Responsiblitty. Correct title for STPEGS is FSAR 13.1 Plant Manager. OP MODE: 1 234 56 DATA 2) 6.1.2 Responsiblitty. Reference to "Vice FSAR 13.1 President. Nuclear Plant Operations" for the OP MODE:1 23456 substitute " Highest level of corporate management STPEGS Organization. DATA 3) 6.2.2 Organization. STPEGS specific titles FSAR 13.1 provided. OP MODE: 1 23456 ED 4) 6.1.2 Changed Vice President-Nuclear Group July 1986 TS Amend. 1 l l q<=gr .. m l - J1 a ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 ThehPlan ['^:^-t]shallberesponsibleforoverallunitopera- l tion and shall delegate in writing the succession to this responsibility 1 during his absence. 6.1.2 The Shift Supervisor (or during his absence from the control roos, a i designated individual) shall be responsible for the control room command function. A management directive.to this effect, signed by the D 'f- !
- 11: ;: : tT!shall be reissued to all station personnel on an annual basis. (muP Wct-Pass,asur.Nman Sisuun.
6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown in Figure 6.2-1. UNIT STAFF 6.2.2 The unit organization shall be as shown in Figure 6.2-2 and:
- a. Each on-duty shift shall be composed of at least the minimum shift .
crew composition shown in Table 6.2-1; REMTOR O b. At least one licensed,0perator shall be in the control room when ft.el is in the reacter. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed SentprV0perator shall be ,in the l .untrol room; KE^ crop,
- c. A Health Physics Technician" shall be on site wnen fuel is in the reactor;
- d. All CORE ALTERATIONS shall be observed and directly supervised by ,Rpx ag RNoFeither a licensed Seniorloperator or licensed Seniorf0perator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;
- e. A site Fire Brigade of at least five members' shall be maintained on site at all times. The Fire Brigade shall not include the Shift '
Supervisor and the'*ttwo75other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency; and l "The Hr 71th Physics Technician and Fire Brigade composition may be less than the r 'tiimum requirevnents for a period of time not to exceed 2 hours, in order to at:ssunodate unexpected absence, provided immediate acticn is taken to fill the required positions. W-STS - 6-1 07/15/M NUREG 0452/STPEGS COMPARISON l L _ -_-_ ! P;ge No. 424 07/15/86 COMPARISON OF NUREG 0452 REV.5 AND STPEGS TECH. SPECS. NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE '; --------------- ------ -------------------------------------------------- ---Operating Moco--- e ** 6.0 PAGE: 6.0 6- 2.0 ' ADMINISTRATIVE CONTROLS OATA 1) 6.2.2.f Organization. Plant Manager and Senior FSAR 13.1 Reactor Operators are the appropriate titles for OP M00E: 1 23456 STPEGS. DATA', 2) 6.2.2.f Deleted shift technical advisor. July 1986 TS Arnend. e e e h e 19' l ADMINISTRATIVE CONTROLS UNIT STAFF (Continued) .
- f. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related
# ~Tunctions (e.g. ,11 censed Senior 0perators, licensed [ Operators, #N" health physicists, auxiliary operators, and key maintenance personnel). -- -- g _. ,.--.. - -.. _ . , . . . ..-., _ -.7-. . - . - - . . . , ,_, ; z a.; : - -_- ; m _. z.. , :.. .. . . . . , . . _ _ - . . . . . . - . . - . . . =...' . .. .. . -. '.z -. = ; ' T,; r ~' "- --.a M Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shut-down for refueling, major maintenance, or major plant modification, on a temporary -9or s%4t te hm.is bas the following guidelines shall be followed(e.**.h c4 .Aviser p et,ennaO:
- 1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time. i i 2. An individual should not be permitted to work more than 16 hours w
l in any 24-hour period, nor more than 24 hours in any 48-hour period, nor more than 72 hours in any 7-day period, all excluding shift turnover time. :
- 3. A breali of at least 8 hours should be allowed between work periods, including shift turnover time.
- 4. Except during extended shutdown periods, the use of cvertime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the QPlantlrg-- :__..__...] or his deputy, or higher levels of manage-ment, in accordance with established procedures and with documenta-tion of the basis for granting the deviation. Controls shall be included in the procedures abch that individual overtise shall be NwerR" reviewed monthly by thej{Plantbr,-- '- _ =---4] or his designee to assure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.D-O PSTS 6-2 07/15/86 NUREG 0452/STPEGS COMPARISON ~x . Page No. 425 12/17/85 COMPARISON OF NUREG 0452 REV.S. AND STPEGS TECH. SPECS. NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE ............... ...... .................................................. ... Operating Mode---
- 6.0 PAGE: 6.0 6- 3.0 ADMINISTRATIVE CONTROLS
- 1) Figure 6.2-1 Of f site Organization. Figure to be provided LATER. FSAR Chapter 13 is updated CP MODE:1 2 3 4 5 6 periodically to reflect organization changes as the STPEGS organization evolves.
9 0 ADMINISTRATIVE CONTROLS O 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) - FUNCTION 6.2.3.1 Thd ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources l of unit design and operating experience infomation, including units of similar - I design, which may indicate areas for improving unit safety. The ISEG shall ) maka detailed recommendations for revised procedures, equipment modifications, , maintenance activities, operations activities, or other means of improving 1 1_ _ _ ; _;11 - -- ' - '_ M l uni t. .s,a fety , _ __to>_{ _ _ '. ' ?, _ ,' _ .11 __ r ,-- _-t= . _ =,__^ ^ "- ? :1 " . _ a w gg . 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engi ne, ers loca, ted on si_,te. , __:::t _ , ..._l , ' _ ' _ ! _. , _ __..J .. ' ; __,. ._ 1. - _... : ._ _,t_
- - :;r <- - __cm;_,, ~-
. . .,- , _ z_ .,_ 7:"-- T : = ,: ::- RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for saintaining surveillance of unit activities to provide independent verification
- that these activities are performed correctly and that human errors are reduced as such as practical. '
RECORDS 6.2.3.4 Records of activities perfomed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to {: '",h I_ _1_;: u;i; ^^" , I;I i -- ; L-2 : . 11.7 ;; W. ;;;i tie r.; % ..;^_ ' 'J.; __:. .e 1 ^_ M M ' r j ,___. 7. _ L-;th..]. T4 Mwur.4 tWC6 EAR A55pMar. 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shift Supervisor in the areas of themal hydraulics, reactor engineering, - and plant analysis with regard to the safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific , or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control roos. 6.3 UNIT STAFF QUALIFICATIONS . r:.:___ _._: :._::----_ e... - . _ .. e : :- - m _ : _ r _ - _ :__ m _: :-- Im _i -- ' U _ ' N ,3I _' h T _ g g ' ' ' _ '[ ~ ' '" ' I ___' _ '_ SU ' e ' ' ' 'C _' _"_ ; '_f_ =_ _ - - = - -' 'sFi e _ F' '.. _i TiJE i._ _~ - ' i EiO 1" Ti ~ _i _' ~
- : - -- '-a_-
.;;g". ;. .m :' _--. . ;, ,:. , _ _. . . _ ~.- : ------ - r - l TZE1' E F- i - i _- - - q . . . .,. .z. ,; _. .- g, . T.. ,..-;., _ , . . ; . . - . . 7 -_ 7: -- c - ; :: _r _---_: 27 _ y _ g ; _ , - 7 , __._. ,__. .__ _.... _ _ -- _ - - - O *Not responsible for sign-off function. W-STS 6-6 01/15/86 NUR'EG 0452/STPEGS COMPARISON L.___ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ Prge No. 431 07/15/86 COMPARISON OF NUREG 0452. REV.5. AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE - --------------- ------ -------------------------------------------------- ---Operating Mode--- . ** 6.0 PAGE: 6.0 6- 7.0 ADMINISTRATIVE CONTROLS DATA 1) 6.3.1 Unit Staff Qualifications. STPEGS specific FSAR 13.1.3.1 dotat1s provided in accordance with FSAR. OP MODE:t 234 56 OATA 2) 6.4.1 Tratning. STPEGS specific title and FSAR 13.2 appropriate standard provided. OP MODE: 1 23456 ED 3) 6.5 Revtew and audit. Delete - this is an information item and the STPEGS spectftc OP M00E:1 23456 ceganizations are described in later sections. DATA 4) 6.5.1 Plant Operations Review Committee (PORC). FSAR 13.4 TS. STPEGS specific organization appited. The PORC is Catawba TS Callaway designed based on the organization in place at VC TS Summer. Catawba and Callaway. The STPEGS Tech OP MODE:t 234 56 Spec to stellar. DATA 5) 6.5.1.3 Statement added that the Plant Manager FSAR 13.4 will serve as PORC Chairman and will not be'an OP MODE: 1 23456 alternate. EO 6) 6.5.1.2 Changed Superintendent to Manager. July 1986 TS Amend. . .? . ADMINISTRATIVE CONTROLS U O rsaaT. .1 NIT i.ch STAFF. emeerQUALIFI, of the unit CATIONS staff shail(Continued) .eet or e ceed the .ini.u. uaiifica-El "' b.I5_!$_ bN 1-iN.! -_ __ N " N - .n_ I "[..N RrAcce l'iii' H :' i .5-~ U h-e' i ~i f f f 85_ ~ The Vicensedj0perators and% REAC seniorA0perators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees. 6.4 TRAINING DAMW6. 4.1 A retraining and replacement tra<.ning pro MA N /:be maintained under the direction ol' thSL__ :^.: gram for the unit staff shall exceed the requirements and recommendat'ons of .. L W and shall meet or . ANSI N /8. / -/Q) 3 L- --d :1;=pr ^ te n= =rc :t:rf: and Appendix A of ID CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience. . 6.5 as a sisctuc ctuct:- . "'t- '-i- h ;= -i== L ; -- ! - - tr i b :_:_:t:"_. _ ;- - - - -- == - - ; : :;-d =;; E : : ]:: _ :: ,; f : ;, :I ^ _ - ; . x_-- :::-g
- t
=7 ::_77- --:: Un _ ; -' .; n ;-g;n ;;p;;n;; .. ; ;- ; ;;; :n_ ---y-----_ : =--- g- ur ; y )= g :: , un:.,up- :_: : -- - : ,- _: ; ;; ; 1 --- r-- 3 --p , _- ge:: e: :: r_ ::: : ;-- r : m - - :x :-:::.: .__ _ : : --;:- u r : er l [^ I- ::q=11: : u,f Uj: -Ex;u ::-d f::{!;;x:: :h:! " g:; : ', urir '- ! 22. _._;: x:: ::_ _.:; gz;;:.;; :;- .n; . . . . . . . . . . . ...:-- z_ zz_:: p: _ --- ; ' I u: b Ld ihr iul30- ng .-._- yl: 27-_$'$_ri _ --- ) , 6.5 T.3' R'"5 CE^ {d^5)3 PLAT 7 OPERATIONS REVIEW COMY~EE (PORC) FUNCTION gg 6.5.1.1 The @A6}3shall function to advise theJPlantgxv= ' n=ru= :C on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The PORC shall be compose'd of the: Chairman: Plant Superintendent or the Plant Manager Member: Technical Support S perint nd:n: ta=*yr Member: Reactor Operations Sup:rintend:ntr h y r Member: Maintenance Superin: cadent Neape-Member: Chemical Operations and Analysis Sup;rin::-d;n: Member: Health and Safety Services Manager N"f# Member: Operations QA Manager ALTERNATES b 6 3 6.5.1.3 All alternatepded shall be appointed in writing by the PORC , Chairman to serve on a temporary basis; however, no more than two altern ges 0 shall participate as vpting members in PORC activities at any one time. ge Th Plant SuperintendenGMlant Manager wfli serve as r0RC Chrritmanc neitheraf-6 shall be considered as an alternate. 07/15/86 NUREG 0452/STPEG5 COMP,miSON w.STS M l Page No. 432 i 12/17/85 COMPARISON OF NUREG 0452. REV.5, AND STPEGS TECH. SPEC 5. NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE ......................-----........--........--.-- ... Operating Moce--- * ' o* 6.0 PAGE: 6.0 6- 8.0 ADMINISTRATIVE CONTROLS DATA 1) 6.5.1.4 PORC - poeting Frequency. STPEGS specific VC Summer TS. titles provided. The STPEGS PORC ts designeo Catawea TS. Callaway based on the organization in place at VC Summer. TS, FSAR 13.4 Catawea and Callaway. OP M00E: 1 23456 OATA 2) 6.5.1.5 PORC - Quorum. STPEGS specific titles FSAR 13.4 provided. OP MODE: 1 234 56 OATA 3) 6.5.1.6 PORC - Responsteiltties. Section coleted FSAR 13.4 and STPEGS specific insert proviceo. OP MODE:t 2 3 4 5 6 9 ADMINISTRATIVE CCNTROLS O - - - - '- ." - j-- i % ~ ; s=_ 21 = ;; ! :: __ ; . _ j ; g- _ = == --J :-2 - ' it =1 --ra, -
- n. m- .n_ _a ___us__ _, . _.e s n >_ =
; - - -- _ , _7 ,__7.__.._-., _;-_ ,- :p :; = _ n=,z 2 n. :;z- _ u :' 2 - - ; _L :=_. ;z :: _ 'L - - - _ =. . . __ _ __ Ib. -.___.,r-.,- 7 _;----;---------,_.., _ _ _ . _ , . . . . . _ _ _ _ -. eeE , - ,_ - < - _ _ _. .. - - - _ _ . . - - - - . - - - - - . . . .. ....--. ..... IW - .. : .. r . 21_ r - - . . : . '; '= . . -- ; = = = r d- -----;- == = = = 5C . .._ _ . . , . _ - . . : 1 ** cy P3R , ,g$Y - C' w 6.5.1.7 The ititt]3shall: MAuurg j 2 a. Recommend in writing to theMPlant, -;::' '_: i '] approval or dis-j E]g & approval of items considered under Specification 6.5.1.6a. through st.e e3 @ - >. ,. . i; E;;-- :_,' .:;i?-- , ANc L n w w I ses A q I oevi Q i E DE si i b. Render deteminations in writing with regard to whether or not each &I item considered under Specification 6.5.1.6a,V: - - - -- constitutes ; EEk, i an unreviewed safety question; and C 4"E d;4 8*' hws U 2% 1 I b (f trga.{S ' / d.
- Provi~de written notification within 24 hours to the,Qice President, Nuclear ^y -t - ] and the :: - n- ., ':_2 _ _ . 7._ . : _ _ _. ._ ^ _:' ^. -
6RevF =;Q N RSB POE of _disacreement between thilititD} and the/Planty,- ' - 7 )i\. Mui op ~ - t V - MAumVh owever, thQPlant/ _;_. ' .'__.._-..^.] shall have. responsibility 'for resolution of such disagreements pursuant to Specification 6.1.1. Asovt, , RECORDS 6.5.1.8 The14fA63 kil maintain written minutes of each meeting that._pogc at a minimum, document the results of all (tite $iactivities perfomed uncer tne ' responsibility provisions of these Technical S Copies shall be provided to the* g ce President,NuclearV"- _^pecifications. _..d .. . _ 'J. _ ! ! , _ . ., ._J.__. . . . :__ _._ +--2 Gaw. NUCl. EAR SAr _ _. _;?(tV REVIEW BOARD (NSRB) 6.5.2 ~ ~ =: =~ =^ ^ n : :=-= ==-'-===r-" n FUNCTION NSRB 6.5.2.1 The (48HtAet3sha11 function to provide independent review and audit of designated activities in the areas of:
- a. Nuclear power plant operations,
- b. Nuclear engineering, O
V-STS 6-9 01/15/86 NUREG 0452/STPEGS COMPARIS0N Page No. 435 ! /Ob COMPARISON OF NUREG 0452. REV.S. AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE , ............... ...... ---------..............----..------.....-- .-- ... ... Operating Mooe--- .. 6.0 PAGE: 6.0 6-10.0 ADMINISTRATIVE CONTROLS DATA 1) 6.5.2. NSRB. Thts section has been deleted - the FSAR 13.4 STPEGS spectftc information is presented on pages OP MODE: 1 234 56 9A througn 9F. EO 2) 6.5.2.1 Changed Vice President-Nuclear Group. July 1986 TS Amend. 1 98 ~ DRAFT ADMINISTRATIVE CONTROLS O FUNCTION (Continued)
- c. Chemistry and radiochemistry,
- d. Metallurgy,
- e. Instrumentation and control, *
- f. Radiological safety,
- g. Mechanical and electrical engineering,
- h. Quality assurance practices, and
. Dit r ;;. +-u te m Me e :::':t:f d ti tt: ::' :: :h::::t: ':ti:: --? the -acc*= ;;-.c v t. . ^. . ] The :(48dtA2) sYaf1 report to and advise the ice-Presiden ucleaN;$reth..C ; on those areas of responsibility specified Specificati 6.5.2.7 and 6.5.2.8. COMPOSITION NS4B 6.5.2.2 The (SHRA6)3sha11 be composed ofitino: A CMas4** Ano A7 LEast Fous MEMases. THE APPoWTW3 .$rN SE map 1 SY mi gVM* EM3MT, MMAll GEEP. -E-i . : - [r::=:r ::; 6*** ('D "'~!-[ f5I!!!!!S!!!!!i _r 7 -. .
- 1.. r.:.:::::r...
u ..... ::.:m::1 l l ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the ' Director to serve on a temporary basis; however, no more than two alternates shall participate as voting members in :tOWl409 ivities at any one time. CONSULTANTS 6.5.2.4 Consultants shall be utilized as detemined by the b Director ! to provide expert advice to the f9WIAG91 V NSRB. MEETING FREQUENCY NSRB 6.5.2.5 The-(tutte%3sha11 seat at least once per calendar quarter during the initial year of unit operation following f sel loading and at least once per 6 months thereafter. , O i-i 5 6 10 07/15/86 NUREG 0452/STPEGS . COMPARIS0N 1 Pcge No. 436 12/17/85 COMPARISON OF NUREG 0452 REV.5 AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE ............... ...... .................................................. ... Operating Mode--- 2 ** 6.0 PAGE: 6.0 6-11.0 ADMINISTRATIVE CONTROLS DATA 1) 6.5.2. NSRB. STPEGS specific information is FSAR 13.4 presentoa. OP M00E: 1 234 56 W 5 t. O gtNISTRATIVECONTROLS QUORUM NSRB NSRB 6.5.2.6 The quorum of the 14NAA8tanecessary for the performance of the @satmC3 review and audit functions of these Technical Specifications shall consist of the Director including or hisNo alternates. designated more than aalternate minority ofand the at least/ quorum shall kave lin responsibility for operation of the unit. REVIEW NSRB 6.5.2.7 The 50NR486 3 sha11 be responsible for the review of:
- a. The safety evaluations for: (1) changes to procedures, equipment,
; , or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59, to verify that such actions did not constitute an unreviewed safety question;
- b. Proposed changes to procedures, equipment,'or systems which involve
~ an unreviewed safety question as defined in 10 CFR 50.59; i
- c. Proposed tests or experiments which trivolve an unreviewed safety ,
question as defined in 10 CFR 50.59; l d. Proposed changes to Technical Specifications or this Operating License; ! e. Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance; . , f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety;
- g. All REPORTABLE EVENTS; .
- h. All recognized indications of an unanticipated deficiency in some aspect of design,or operation of structures, systems, or components that could affect nuclear safety; and
- i. Reports and meeting minutes of th AUDITS 6.5.2.8 V
Audits of unit activities shall be performed under the cognizance of the idNRhek These audits shall encompass: NSRB
- a. The conformance of unit operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months;
'O W-STS - 6-11 01/15/86 NUREG 0452/STPEGS COMPARISON Page No. 437 07/15/86 COMPARISON OF NUREG 0452 REV.S. AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE ............... ...... .................................................. ...Op.e.,,ng ucce... ** 6.0 PAGE: 6.0 6-12.0 AOMINISTRATIVE CONTROLS DATA 1) 6.5.2. NSRB. STPEGS specific information is FSAR 13.4 presented. OP M00E: 1 234 56 OESIGN 2) 6.5.3 Technical Review and Control. Section FSAR 13.4 added page 6-9F. The STPEGS administrative OP M00E: 1 234 56 procedure is based on that used by VC Summer. Callaway, and Catawca. . I EO 3) 6.5.2.8.k, 6.5.2.9.a, 6.5.2.9.b Changed Vice July 1986 TS Amend. President-Nuclear Group. e l 1 l - 1 l l DRAFT ADMINISTRATIVE CONTROLS AUDITS (Continued) -.-
- b. The performance, training, and qualifications of the entire unit staff at least once per 12 months;
- c. The results of actions taken to correct deficiencies occurring in .
unit eq:.:ipwat. structures, systems, or method of operation that afety, at least once per 6 months; ~ t (f4: - , d. The performance of activities required by the Operational Quality Assurance Progrce to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months; I ,e. The fire protection programmatic controls including the implementing procedures at least once per 24 ~eonths by qualified licensee QA ; . personnel;
- f. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection i consultant. An outside independent fire protectio'n consultant shall be used at least every third year;
! g. The Radiological Environmental Monitoring Program and the results l thereof at least once per 12 months;
- h. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months;
- 1. The PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months;
- j. The perfomance of activities required by the Quality Assurance Program for effluent and environmental sonitoring at least once per 12 months; and
- k. Any other area of unit operation considered appropriate by the NSRB l4htACQ or thef{Vice PresidentfNuclearYL . J.: _ . .C .
. (sacar, "Q RECORDS 6.5.2.9 Records of 14Huti act vities shall be prepared, approved, and dis-tributed as indicated below: NSRB
- a. Minutes cf each s meeting shall be prepared, approved, and forwardet to th Vice PresidentyNuclearY^ 6digu' 2:_.?' within 14 days following each meeting;
#9
- b. Reports c' reviews encompassed by Specif ation 6.5.2.7 shall be O preoared, approved, and forwarded to the Vice President-Nuclear 11Emm
^ - 0:: : _ _ ) within 14 days following completion of the review; arc
- w. 5 5 6-12 07/15/86
~ NUREG 0452/STPEGS l . COMPARISON l Insa RT 6.5.3 * .6.5.3 TECMMICAL REVIEW AND CONTROL ACTIVITIES I 6.5.3.1 Activities which affect nuclear safety shall be conducted as follows:
- a. Procedures required by Technical Specification 6.8, and other procedures which affect nuclear safety, and changes thereto, shall be prepared, reviewed, and approved. Each such procedure, or change '
thereto, shall be reviewed by an individual / group other than the individual / group which prepared the procedure, or change thereto, but who say be from the same organisation as the individual / group which prepared the procedure, or change thereto. Procedures other tt n j ' station administrative procedures shall be approved as delineated in writing by the Plant knager. The Plant knager shall approve station administrative procedures, security plan implementing procedures, and energency plan taplementing procedures. Temporary changes to procedures which clearly do not change the intent of the , approved procedures, shall be approved prior to implementation by two - members of the plant staff, at least one of whom holds a Senior Reactor Operator's License. Changes to procedures which may involve i a change to the intent of the original procedure shall be approved by the individual authorised to approve the procedure prior to , implementation of the change. .b. Proposed changes or modifications to safety-related structures, systems, and components shall be reviewed as designated by the Flant Manager. Each such modification shall be reviewed by an individual / group other than the individual / group which designed the
- modification, but who say be from the same organisation as the individual / group which designed the mod!fication. Proposed modifications to safety-related structures, systems, and components
,shall be approved by the Plant knager prior to implementation. l Mmen
- c. Proposed tests and experiments which affect nuclear safety andgare not addressed in the Final Safety Analysis Report shall de prepared, ,
reviewed, and approved prior to implementation. Each such test or - experiment shall be reviewed by an individual / group other than the individual / group which prepared the test or experiment but who say be from the same organisation as the individual / group which prepared the test or experiment. Proposed tests and experiments shall be approved by the Plant knager. SpruncArsed
- d. Individuals responsible for reviews performed in accordance with 6.5.3.1 (a) through W) shall be members of the plant management3 staff previously designated by the Plant knager. Each review shall include a determination of whether or not additional, cross-disciplinary review is necessary. If deemed necessary, such review shall be performed by empato personnel of the appropriate i
discipline. \ -
- e. Each review will include a determination of whether or not an l
unreviewed safety question is involved. Pursuant to section 50.59, 10CFR, NRC approval of items involving an unreviewed safety question will be obtained prior to Plant knager approval for implementation. 6.5.3.2 Records of the above activities shall be provided to the Plant i Manager, FORC, and/or the NSRB as necessary for required reviews. 01/15/86 NUREG 0452/STPEGS I 6 - 12. B Co m DISON 1 -----.-w---- . , . . , . _ . - . , , _ - , , _ - - - . - _ _ . - - - Ptge No. 438 07/15/86 CouPARISON OF NUREG 0452 REv.5. ANO STPEGS TECH. SPECS. NOTE TYPE NOTE a NOTES FSAR CROSS REFERENCE --------------- ------ -------------------------------------------------- ---Operating Mooe--- ** 6.0 PAGE: 6.0 6-13.0 ACMINISTRATIVE CONTROLS DATA 1) 6.6 Reportable Event Action. STPEGS spectftc VC su.4mer TS. titles provided. Callaway TS. Catawba TS. FSAR 13.1 OP M00E: 1 23456 CATA 2) 6.7.1 Safety Limit Violation. STPEGS specific FSAR 13.1 titles proviced. OP M00E:1 23456 CESIGN 3) 6.7.1.d Safety Limit violation. Change to read " Critical operation of the plant. .". Normal OP M00E: 1 23456 non-critical operation of the plant will be necessary untti permission is granted by the . Commission for critical operation. EC 4) 6.8.1.0 Procedures and Programs. Add "Section 7.1" of Generic Letter 82-33. This provides exact OP M00E: 1 234 56 reference. I ED 5) 6.5.2.9.c, 6.6.1.b, 6.7.1.a, 6.7.1.c Changed July 1986 TS Amend. Vice President-Nuclear Group p l l - l e e pp$q 8 ' ADMINISTRATIVE CONTROLS RECORDS (Continued) 0* c. 6teu? Audit ri tports encompassed by Specification 6.5.2.8 shall be forwarded to the [Vice President; Nuclear droactione Mand to the management post- e tions responsible for the areas audited within tionoftheauditbytheauditingorganization.yI days after comple-l i 6.6 REPORTA8LE EVENT ACTION l 6.6.1 The following actions shall be taken for REPORTA8LE EVENTS:
- a. The Comission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10,CFR Part 50, and
- b. Each REPORTABLE EVENT shall be reviewed by the k , and the
%results of this review W shall Vice PresidentyNuclearV!; . J. :be . L submitted NSRBto theVM and the 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a safety Limit is violated:
- a. TheNRCOperationsCentershallbenotifiedb]ytelephoneassoonas
. possible and in all cases within 1 hour. TheMVice PresidentyNuclear desma 4.._tt::.;) and theV" -- .0; shall be notified within 24 hours; N5YS u,
- b. A Safety _ Limit Violation Report shall be prepared. The report shall PORM be revrewed by theit#RG3 This report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence;
- c. The_ Safety Limit Viol tion Report shall be submitted to the Comission, N5RV theitcN#463, and the Vice.PresidentyNuclea#!;_ iti: :] within 14 days of the viola ion; and d$napa C= men
- d. Apperationoftheunitshallnotberesumeduntilauthorizedbythe Commission.
6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
- a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978;
- b. The emergency operating procedures required to implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33[Sta,ov 7.1.
O y-STS 6-13 07/15/86 NUREG 0452/STPEGS COMPARISON Page No. 439 12/17/85 COMPARISON OF NUREG 0452. REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES FSAR CROSS REFERENCE ............... ...... .................................................. ... Operating Mode---
- 6.0 PAGE: 6.0 6-14.0 ADMINISTRATIVE CONTROLS DATA 1) 6.8.2 Procedures and Programs. STPEGS specific FSAR 13.1 titles provided. OP MODE:t 234 56 ED 2) 6.8.2 Procedures and Programs. Change "..as set fceth in Aantnistrative procedures. ' to ' . .as set CP MODE:t 2 3 4 56 forth in Specification 6.5 above.' This is a more complete reference.
OTHER 3) 6.8.3 Procedures and Programs. Deleted. The VC Summer TS. making of temporary changes is covered in Callaway TS Spectf1 cation 6.5. Consistent with VC Summer and OP M00E: 123456 Callaway Tech Specs. ED 4) 6.8.4 Procedures and Programs. Renumbered 6.8.3 due to deletion of 6.8.3. OP M00E: 1234 56 DESIGN 5) 6.8.3.a Change 'rectreulation spray' to F S AR 7A. III .0.1.1 ' containment spray'. Delete ' gas stripper and CP M00E: 1 234 56 hydrogen recombiner'. Add 'RCS sampling' to correctly identify STPEGS systems implementing the program 10 reduce leakaQG outside Contatnment. 9 _ . _ _ _ , , _ . . - - _ _ . - _ .}}