ML072000394
| ML072000394 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/13/2007 |
| From: | Dominion Nuclear Connecticut |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 07-0450 | |
| Download: ML072000394 (33) | |
Text
{{#Wiki_filter:Serial No. 07-0450 Docket No. 50-423 ATTACHMENT 3 LICENSE AMENDMENT REQUEST STRETCH POWER UPRATE MARK-UP OF THE OPERATING LICENSE AND TECHNICAL SPECIFICATIONS PAGES DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 3
August 11, 2005 e-INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
- G9~6 ~(
SECTION 3/4.9.6 DELETED 3/4.9.7 DELETED ,. 3/4 9-7 V 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level 3/49-8 Low Water Level 3/4 9-9 3/4.9.9 DELETED 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL 3/49-11 3/4.9.11 WATER LEVEL - STORAGE POOL 3/49-12 3/4.9.12 DELETED 3/49-13 3/4.9.13 SPENT FUEL POOL - REACTIVITY 3/49-16 3/4.9.14 SPENT FUEL POOL - STORAGE PATTERN 3/49-17 FIGURE 3.9-1 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 1 4-0UT-OF-4 STORAGE CONFIGURATION 3/4 9-18 FIGURE 3.9-2 REGION 1 3-0UT-OF-4 STORAGE FUEL ASSEMBLY LOADING SCHEMATIC A 3/49-19 FIGURE 3.9-3 MINIMUM FUEL ASSEMBLY BURNUP.rvE~~~E~~~~e.. INITIAL ENRICHMENT FOR REGION 2' y STORAGE CONFIGURATION 3/49-20 FIGURE 3.9-4 MINIMUM FUEL ASSEMBLY BURNUP AND DECAY TIME VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 3 0410 SPECIAL :;:T~:~E~:~~U~~~~~Up;;H~~~V\w:E-t~~~E (34 \\ V\\Nt~ LO'i-F 3/4.10.1 SHUTDOWN MARGIN 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS 3/410-2 3/4.10.3 PHYSICS TESTS 3/410-4 3/4.10.4 REACTOR COOLANT LOOPS 3/4 10-5 3/4.10.5 DELETED 3/4.11 DELETED "F\Cx\.)\2-~ ~ .'1-5'" \~\\~\"'\IV\ ~\JEll\~~M~L'\ 3/4.11.1 DELETED '¥:.\l"-\J\R ~~ \)E'C1\'( "\ \ \AS ,,\,: ~S> \Jo\--\ \ iJ ~L 3/4.11.2 DELETED \\)\\\"l.- E\.)<:(...\( ~V\t~ ~DR 'RG&\'\j\J '3 5\1)"-~6cE CD\Sf 'G:.lR.r'f"\ ~ r:D~ ~S5H-1~USS 3/4.11.3 DELETED 1-~b"" ~tt>\-\YVRt\-rf(~10 S1) t\\UtJ ~~\1..E~ . MILLSTONE - UNIT 3 Xll Amendment No. ;W, 89, +&-&, W,~, m,m,Jtl
~une 29, 2005..Q..,
INDEX LIMITING CONDITIONS FOR OPERATION.AND SURVEILLANCE .REQUIREMENTS SECTION Air Temperature 3/4 6-9 Containment Structural Integrity 3/4 6-10 Containment Ventilation System 3/46-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Contairunent Quench Spray System 3/4 6-12 Recirculation Spray System 3/4 6-13 3/4.6.3 CONTAINMENT ISOLATION VALVES 3/4 6-15 3/4.6.4 DELETED 3/4.6".5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector 3/4 6-18 3/4.6.6 SECONDARY CONTAINMENT Supplen1entary Leak Collection and Release System 3/4 6-19 Secondary Contailill1ent 3/4 6-22 Secondary Contairune11t Structural Integrity 3/4 6-23 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves 3/4 7-1 OPERA~ L.G. mss.\IS VE.Rsus \~AX\MU\~ ALLOl.OABt...6 TABLE 3.7-1 ~iA)(IfyfUhf ALLO'\VABLE PO\\TER Rz(\}~GE ~iEUTR. Ot~ FLUX IIIOIi* PoWE" 1 SETPOl}1T VlITH [~(OPERABLE STEAlvI LINE S*f1ET \TALTvES .... 3/4 7-2 3 { TABLE 3.7-2 DELETED 3/4 7-2 MILLSTONE - UNIT 3 IX Amendment No. -59, B, &1, -89, -l-9{},
~,m,m,jJ/t
B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses; (1) ONC, pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use and operate the facility at the designated location in New London County, Connecticut in accordance with the procedures and limitations set forth in this license; Central Vermont Public Service Corporation and Massachusetts Municipal Wholesale Electric Company, pursuant to the Act and 10 CFR Part 50, to possess the facility at the designated location in New London County, Connecticut in accordance with the procedures and limitations set forth in this renewed operating license; (2) ONC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended, (3) ONC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) ONC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) ONC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operations of the facility. C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: ( 1) Maximum Power Level
...,.."",.=""3650 ONC is auth rized to operate the facility at reactor core power levels not in excess 0 4 1 egawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
Renewed License No. NPF-49
DEFINITIONS PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests perfonned to measure the fundamental nuclear characteristics of the reactor core and related instrumentation: (l) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the r~ommlSSlOn. 1.22 DELETED PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3+t MWt. I I I 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC. REPORTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster jjj assembly of highest reactivity worth which is assumed to be fully withdrawn. MILLSTONE - UNIT 3 1-5 Amendment No. 6ll, H'I, m#,
~a[ch 14, ~M'f2-2.0 SAFETY UMITS AND LIMITING SAFETY SYSTEM SETTINGS 2,1 SAFETY UMITS REACTOR CORE 2.1,1 The combination of THERMAL POWER, Reactor Coolant System highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the CORE OPERATING UM/TS REPORT; and the following Safety Lill1ltS shall not be exceeded: I I I
2.1,1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater 1 than or equal tc,lTi'7tfor the t:YR~WRB-JDNB cOll*elations.
~lr WRB*_*2.M 2.1.1,2 The peak fuel centerline temperature shall be maintained les, than 5080°F,
/ decreasing by 58°F per 10,000 fv/WD/MTU of bUJl1up. APPLICABILITY. MODES I and 2. ACTION Wl1enever the Reactor Core Safely Limit is violated, restore compliance and be in HOT STANDBY within 1 houl'. REACTOR COOLANT SYSTElvl PRESSURE 2.\ ,2 The Reactor Coolant Syslem pressure shall not exceed 2750 pSla APPLICABILITY: lV/ODES 1.2. J, 4, and 5 ACTION. MODES 1 and 2: Whenever the Reactor Coolant System pressure has exceeded 2750 pSla be in HOT STANDBY with the Reactor Coolant System pressure within its llmll \.vithin 1 hour. MODES 3,4 and 5: Whenever the Reactor Coolant System pressure has exceeded 2750 pSla, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. MILLSTONE - UNIT 3 2-1 Amendment No, ++.3-, ~'#f
NOMINAL FUNCTIONAL UNIT TRlP SETPOINT ALLOWABLE VALUE
- 8. Overpower ~T See Note 3 See Note 4
- 9. Pressurizer I>ressu,re-Low 1900 psia ~ 1897.6 psia
- 10. Pressurizer Pressure-High 2385 psia ~ 2387.4 psia
- 11. Pressurizer Water Level-High 89% of instrument S 89.3% of instrument span span 12.
13. Reactor Coolant Flow-Low Steam Generator Water 18.1 % of narrow ~ 17.8% of narrow I Level Low-Low range instrument range instrument span span
- 14. General Warning Alarm N.A. N.A.
- 15. Low Shaft Speed - Reactor 92.4% of rated ~ 92.2% of rated Coolant Pumps speed speed J easured Flov,;
r .~
~
P TABLE 2.2-1 (Continued)
~....., =RE=A=C:..:T,-,=O<.:.R::...T~~P,"""S",-,Y<::...:S",-,T::.::E=2\=l..:I::'i:...:..*:::..ST=-:R=_U=ME""-=::.:..N,-"T~A=T.....IO=N"-,-"T~R=I,,,-P-,=,S=E~T-,,,-P.=..:=,-,-T~S ~ OOMm&
tTJ FUNCTIONAL UNIT TRIP SETPOmT ALLOWABLE VALUE
~ 16. Turbine Trip w a. Low Fluid Oil Pressure 500 psig 2:: 450 psig
- b. Turbine Stop Valve 1% open 2:: 1% open Closure
- 17. Safety Injection Input N.A. N.A.
tv I from ESF -.l
- 18. Reactor Trip System Interlocks
- a. Intermediate Range 1 X 10- 10 amp 2:: 9.0 X 10-)) amp Neutron Flux, P-6
- b. Low Power Reactor Trips Block, P-7
- 1) Power Range Neutron Flux, 11 % ofRTP** :$ 11.6% ofRTP**
§' P-I0 input (Note 5)
[ 2) Turbine Impulse Chamber Pressure, 10% RTP** Turbine :$ 10.6% RTP** Turbine S (ll P-13 input Impulse Pressure Impulse Pressure g Equivalent Equivalent Z o v$
- c. Power Range Neutron Flux, P-8 uc:% ofRTP**
5°.. 0>
- $ ESJr"~fRTP**
t.. !S"().(;' 1 v~ v~ ** RTP = RATED THERMAL POWER v t
.A LE 2,2-1 (Continued) 3'::
P l' UJ TABLE NOTATIONS
--3 o NOTE 3: OVER POWE R .6.T Z
tTl Where : .6.T is measu red Reacto r Coolan t System!::..T, OF;
.6.To is loop specifi c indicated.6.T at RATED THERM AL POWE R, OF; (1 +'t s) 1 (1 + 't s) is the functio n genera ted by the lead-lag compe nsator on 2 measur ed .6.T;
~
'1:1 and't2 are the time constants utilized in the lead-lag compe nsator for
.6.T, 'tJ ~ [*] sec, 't2:::; [*] sec; Ks ~[*]/°F for increas ing Tavg and K ~ [*] for decreasing T ;
s avg I ('t s) 7 (1 + 't s) is the functio n generated by the rate-lag compe nsator for Tavg ; 7
't7 is the time constan t utilized in the rate-lag compe nsator for T avg , 't7 ~ [*] se~
T is measu red averag e Reacto r Coolan t System temperature, OF; zo 1" is loop specifi c indicat ed Tavg at RATED THER MAL POWE R, ~ [*]OF;
~ ~ [*]/°F when T > T" and K 6 ~ [*]rF when T ~ T";
s is the Laplac e transfo rm operator, sec -1; (The values denote d with [*] are specified in the COLR.)
fiL Mal eI I 14, 205'7' POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION N 3.2,3.1 The indicated Reactor Coolant System (RCS) total flow rate and F6H shall be maintained as follows: ~l3,2.oG a, RCS total flow rate ?lf1,m iJ gpl11 and greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR), and N RTP b, F6H SF 6H [1.0 + PF 6H( 10 - P) J Where: P = THERMAL POWER I) RATED THERMAL POWER'
- 2) F:H = Measured values of F: obtained by using the movable H
incore detectors to obtain a power distribution map. The measured value N of F6H should be used since Specification 3,23, I b. takes into consideration a measurement uncertainty of 4% for incore measurement, RTP. N
- 3) F 6H = The F 6 H limIt at RATED THERMAL POWER in the COLR, N
- 4) PF 6H = The power factol' multiplier for FL'lH provided in the COl~R, and
- 5) The measured value of RCS total flow rate shall be used since uncertainties of2.4% for flow measurement have been included in Specification 3.2.31 a.
APPLICABILITY: MODE 1, ACTION: N With the RCS total flow rate or F 6H outside the region of acceptable operation:
- a. Within 2 hours either:
- 1. Restore the RCS total flow rate to within the limIts specified above and in N
the COLR and F6H to within the above limit, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
3/4 2-19 Amendment No. H, §{f, eG, +t4, ct+, MILLSTONE - UN IT 3
!fe1
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION: (Continued)
- b. Within 24 hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate that the RCS total flow rate is restored to N
within the limits specified above and in the COLR and FLH is restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours.
- c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b., above; subsequent POWER OPERATION may proceed N
provided that FLH and indicated RCS total now rate are demonstrated, through incore nux mapping and RCS total flow rate comparison, to be within the region of acceptable operation prior to exceedll1g the following THERMAL POWER levels. I. A nominal 50% of RATED THERMAL POWER,
- 2. A nominal 75% of RATED THERMAL POWER, and
- 3. Within 24 hours of attall1lJ1g greater than or equal to 95% of RATED THERMAL POWER.
SURVE1LLANCE REQUIREMENTS 4 IJ 4.2.3. 1.1 The provisions of Specification 4.0.4 are not applicable. LrP - 4-"..\~e . .~*1--
-e.e\ ~\AJJI 4.2.3.1.2 RCS total flow rate and FflH shall be detennined to be within the acceptable range: ~I
- a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and
- b. At least once per 31 Effective Full Power Days.
I 4.2.3.1 J The indicated RCS total fiow rate shall be verified to be within the acceptable range at N least Once per 12 hours when the most recently obtained value of FLH' obtained per Speci fication . 4.2.31.2, is assumed to exist. 4.2.3.1.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be
,calibrated within 7 days prior to the perfonnance of the calorimetric flow measurement.
MILLSTONE - UNIT 3 3/4 2-20 Amendment No. GG,:P), +{fG,
-1J. ~
J
INSERT 1 to Page 3/4 2-20 4.2.3.1.2 F NAH shall be determined to be within the acceptable range:
- a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and
- b. At least once per 31 Effective Full Power Days.
4.2.3.1.3 The RCS total flow rate shall be determined to be within the acceptable range by:
- a. Verifying by precision heat balance that the RCS total flow rate is ~
363,200 gpm and greater than or equal to the limit specified in the COLR within 24 hours after reaching 90% of RATED THERMAL POWER after each fuel loading, and
- b. Verifying that the RCS total flow rate is ~63,200 gpm and greater than or equal to the limit specified in the COLR at least once per 12 hours.
4.2.3.1.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.
--t.(a~h I! I I ~
POWER DISTRIBUTION U S SURVEILLANCE REQUIREMENTS (Continued) 4.2.3.1.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. Within 7 days prior to performing the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi M' in the calorimetric calculations shall be calibrated. l 4.2.3.1.6 If the feedwater venturis are not inspected at least once per 18 months, an additional 0.1 % will be added to the total RCS flow measurement uncertainty. LSTONE - UNIT 3 3/42-21 Amendment No. E'lf
T E 3.3-3 (Continued) NGINEERED SA ~" FEATURES ACTUATION SYSTEM _ STRUMENTA: ION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE ~~ FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION L0 9. Engineering Safety Features Actuation System Interlocks
- a. Pressurizer Pressure, 3 2 2 1,2,3 21 P-ll L0
- 4 b. Low-Low Tavg , P-12 4 2 3 1,2,3 21 L0
~
t3 c. Reactor Trip, P-4 2 2 2 1,2,3 23
- 10. Emergency Generator 2 2 1,2,3,4 15 Load Sequencer 1.1.r Cdd 1-6!~ "I~ ~c.+ioV) tr 3 ~ 'Z./~
~b
\
r~mi.$S-IVC7 (?-.l~ zo
MdI ~j, 2", 260~ TABLE 3.3-3 (Continued) TABLE NOTATIONS
# The Steamline Isolation Logic and Safety Injection LOgIC for this trip function may be blocked in this MODE below the P- I I (Pressurizer Pressure Interlock) Setpoll1t.
~Y\t.I L-
- MODES I, 2, 3,~4lfalli! 6 ~-
DUling fuel movement within containment or the spent fuel pool.
- Trip function automatically blocked above P-II and may be blocked below P-II when Safety Injection on low steam line pressure is not blocked.
ACTION STATEMENTS ACTION 14- With the number of OP ERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; however, one charmel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.2. I, provided the other channel is OPERABLE. ACTION 15- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable charmel to OPERABLE status within 12 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDO\Al'N within the following 30 hours; however, one channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE. ACTION 16- With the number of OPERABLE chalUlels one less than the Total Number of Channels, operation may proceed until perfonnance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within I hour. ACTION 17- With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.2.1. ACTION 18- With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 7 days. After 7 days, or ifno channels are OPERABLE, immediately suspend fuel movement, if applicable, and be in HOT STANDBY within the next 6 hours and in COLD SHUTDOVlN within the following 30 hours. ACTION 19- With the number of OPERABLE chalUlels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. MILLSTONE - UNIT 3 3/43-24 Amendment No. ~, 1Q, 84, m. ~, wJ}11
TABLE 3.3-4 (Continued) 03: g: .... ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS WI I VI
-i o
z rn c
- z:
FUNCT IONAl UN IT
- 8. Loss of Power NOMINAL TRIP SETPOINT ALLOWABLE VALUE (r
- a. 4 kV Bus Undervoltage 2800 > 2720 volts (Loss of Voltage) volts with with a < 2 a < 2 second second time time delay. delay.
- b. 4 k~ Bus Undervoltage 3730 volts > 3706 volts (Grid Degraded Voltage) with a < 8 with a < 8 second time second time delay with ESF delay with ESF actuation or actuation or
< 300 second < 300 second w time delay time delay ~
without ESF without ESF W actuation. actuation. I W o 9. Engineered Safety Features Actuation System Interlocks
- a. Pressurizer Pressure, P-ll 1999.7 psia 5. 2002.1 psia
- b. Low-Low Tavg ' P-12 553' F L 552.6*F
- c. Reactor Trip, P-4 N.A. N.A.
~
- 10. Emergency Generator load N.A. N.A.
Sequencer
- z:
1.1
- Co\J ~e5 :r~~~ho~ 19 00 PS;C\.. ' l f~7.b pSt&.
r f-e'r tr-ii 5s~\( 41 1'- it. 't-1
o 3 TABLE 4.3-2 (Continued) Febiudry EO, gOO~ --J ..... ~ I I
~
<.J'> ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 0
'Z SURVEILLANCE REQUIREMENTS r
TRIP c::
- z ANALOG ACTUATING MODES
..... CHANNEL DEVICE MASTER SLAVE FOR WHICH
-l
'oJ.)
CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 7. Control Building Isolation (Continued)
- e. Control Building Inlet Ventilation Radiation S R Q N.A. N.A. N.A. N.A. * ((
- 8. Loss of Power
- a. 4 kV Bus N.A. R N.A M(3) N.A. N.A. N.A. 1, 2, 3, 4 Undervoltage (Loss of Voltage) w
......... b. 4 k Bus N.A. R N.A. M(3) N.A. N.A. N.A. 1, 2, 3, 4 w Undervoltage (Grid I
.;::. Degraded Voltage) c
- 9. Engineered Safety Features Actuation
):>
System Interlocks 3 ttl
~
0..
- a. Pressurizer N.A. R Q N.A. N.A. N.A. N.A. 1, 2, 3 3
ro Pressure, P-ll
~
.,-T
- b. Low-Low Tavg , P-12 N.A. R Q N.A. N.A . N.A. N.A. 1, 2, 3
- c. Reactor Trip, P-4 N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3 0
~ -~~ 10. Emergency Generator N.A. N.A. N.A. N.A. Q(1, 2) N.A. N.A. 1, 2, 3, 4 Load Sequencer N.A*, tJ-./'r* ~-ft* NJ*/'t '. tl '2./~
'-..I '11.. cold l-f.5 ~~ ec.nm,-" S fl Y-Iml~i\l'e I f- ~,
--..i
~
'I'ABLE 4.3-2 (Continued)
TAHI.11: NOTATION
- 1. Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
- 2. This surveillance may be perfonned continuously by the emergency generator load sequencer auto test system as long as the EGLS auto test system is demonstrated OPERABLE by the performance of an ACTUATION LOGIC TEST at least once per 92 days.
- 3. On a monthly basis, a loss of voltage condition wi II be initiated at each undervoltage monitoring relay to verify individual relay operation. Setpoint verification and actuation of the associated logic and alann relays will be performed as part of the CHANNEL CAUBRATION required once per] 8 months.
- 4. For Engineered Safety Features Actuation System functional units with only Potter &
Brumfield MDR series relays used in a clean, envtronmentally controlled cabinet, as discussed in Westinghouse Owners Group Report WCAP- 13900, the surveillance interval for slave relay testing is R.
~~ ..-
- MODES I, 2, 3,~4.Slqw! 6 \I \l During fuel movement within containmenl or the spenl fuel pool.
MILLSTONE - UNIT 3 3/43-4 ]
May 27, 1998 PRESSURIZER LEVEL CONTROL 70 ...**. *., ***.******** T****--**** TO""" .-._. T" ****.*****.**** --- .. - ..* y *********** ' " ************ , .************ ':
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o \. I 551 557 562 567 572 577 582 587.1 T (AVG) FIGUR 3.4-5 I ( I ~lLlSTONE \
- UNIT 3 3/4 ,-lla Amendment
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--------'--~-~
I 550 555 560 505 570 575 580 585 590 ________J
QS8SltiBer lQ, 20~ 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION r-CMfSVS) 3.7.1.1 All main steam line Code safety valve~shan be OPERABLE with lift settings as specified in Table 3~7-3. -", APPLICABILITY: MODES 1, 2, and 3. ACTION: [~oTE. ~ s:-e1?Q)vq-l-~ c...o~*,\;o"" e..\'\.t,,)' '" al\owe.J -ftvy ~dA ..-ssvJ
- a. With one or more main steam line Code safety valves inoperable, operation in MODES 1,2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7 -I; otherwise, be in at least HOT STANDBY within the next 6 hourS and in HOT SHUTDOWN within the following 6 hours. ....&.-
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5, ~\~
~~. ?yC)v',GCOlY'S of ~fe.~i?cCl\~~H.~ 4-0-4 ~)J~ \1~1r ~@pl.i*ce,hlc:.
40y "t..""n-:,.. \~h Mo~ C; ~ " LSTONE - UNIT 3 3/47-1 Amendment No.-s:t, 1/1/
INSERT 'A' to Page 3/4 7-1
- a. With one or more steam generators (SOs) with one MSSV inoperable, and the Moderator Temperature Coefficient (MTC) zero or negative at all power levels, within 4 hours reduce THERMAL POWER to less than or equal to 60.1 %
RATED THERMAL POWER (RTP); otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With one or more SOs with two or more MSSVs inoperable, within 4 hours reduce THERMAL POWER to less than or equal to the maximum allowable %
RTP specified in Table 3.7-1 for the number of OPERABLE MSSVs, and reduce the Power Range Neutron Flux High setpoint to less than or equal to the maximum allowable % RTP specified in Table 3.7-1 for number of OPERABLE MSSVs within the next 32 hours*; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- c. With one or more SOs with one MSSV inoperable and the MTC positive at any power level, within 4 hours reduce THERMAL POWER to less than or equal to the maximum allowable % RTP specified in Table 3.7-1 for the number of OPERABLE MSSVs and reduce the Power Range Neutron Flux High setpoint to less than or equal to the maximum allowable % RTP specified in Table 3.7-1 for number of OPERABLE MSSVs within the next 32 hours*; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- d. With one or more SOs with four or more MSSVs inoperable, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- Applicable only in MODE 1.
t)eeem&er 10,20'03 Q-.
- -7 ,- ~ TABLE 3.7-1 , o' ,
O,Pl:i12ABl..tt. MS&Ys Ve.-.,su..s lV\~*,m,-,W\ A\\,oU)q.b'e. rOlOE..y "MAXJ)fuM ALM;mULE fO'HER R&NYE tfEUTBON FLux mWSEDQ~
--WI1'B INQPEBa~BLE STEl\M LINE 54fET¥ '1\L1TES ~
MAxIMUM NUMBER OF mOPERABLea- *~~~:fBJ::POWER
. , SAFETY 'VALVES'ON AN\Y. 0., NEUTRON FLU* HIGH,'8ETPOI~ .OfflRAw<J s'mAivt 6EmftATORC2. (PERCENT"OF':RATED rimRMAt,.POWER)
[D-lt l[SJ- 60-1 ill-3 Wl 4-2-B lJl- 1. GID. Z 5
- 5 NUB~.,~ OF C>,P.ERABL£
{Yl s*sV.-;* p ~ t<. S-rs..AN\ G E-I\J Ete lrro1<--. ~ABLE3.'7~2 DELETED Amendment NoC{OO;, 111
PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 OPERABLE.# Two independent Control Room Emergency Air Filtration Systems shall be APPLICABILITY: r~WCl{tt MODES 1, 2, 3~,IS ana 6-!- rI During fuel movement within containment or the spent fuel pool. ACTION: MODES 1,2,3 and 4:
- a. With one Control Room Emergency Air Filtration System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With both Control Room Emergency Air Filtration Systems inoperable, except as specified in ACTION c., immediately suspend the movement offuel within the spent fuel pool. Restore at least one inoperable system to OPERABLE status
( within I hour or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
- c. With both Control Room Emergency Air Filtration Systems inoperable due to an /
inoperable Control Room boundary, immediately suspend the movement of fuel within the spent fuel pool and restore the Control Room boundary to OPERABLE status within 24 hours or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. f ~~~~~..,.,mlfr~";'V\~ unm fuel movement within containment or the spent fuel pool:
- d. With one Control Room Emergency Air Filtration System inoperable, restore the inoperable system to OPERABLE status within 7 days. After 7 days, either initiate and maintain operation of the remaining OPERABLE Control Room Emergency e.
Air Filtration System in the recirculation mode of operation, or immediately suspend the movement of fuel. With both Control Room Emergency Air Filtration Systems inoperable, or with the ( OPERABLE Control Room Emergency Air Filtration System required to be in the recirculation mode by ACTION d. not capable of being powered by an r OPERABLE emergency power source, immediately suspend the movement of fuel.
- The Control Room boundary may be opened intermitte tly under administrative control.
MILL TONE - UNIT 3 3/47-15 Amendment No.2, i-&-l-, M, W
~e~teffiBer 12, 28~
TABLE 3.7-6 (Continued) AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT (oF)
- 7. FUEL BUILDING FB-02, Fuel Pool Pump Cubicles, £1 24'6" ~ 119 FB-03, General Area, £1 52'4" ~ 108
- 8. FUEL OIL VAULT FV-OI, Diesel Fuel Oil Vault ~ 95
- 9. HYDROGEN RECOMBINER BUILDING HR-OI, Recombiner Skid Area, El 24'6" ~ 125 HR-02, Controls Area, £1 24'6" ~ 110 HR-03, Sampling Area, El 24'6" ~ 110 HR-04, HVAC Area, E1 37'6" ~ 110
- 10. MAIN STEAM VALVE BUILDING MS-OI, Areas above El. 58'0" ~ 140 MS-02, Areas below El. 58'0" ~ 140
- 11. ,iORBINE B ILDING
~: Entire Building
- 12. TUNNEL T -02, Pipe Tunnel-AuXiliary, Fuel and ESF Building ~ 112
- 13. YARD YD-Ol, Yard < 115 IllSTONE - UNIT 3 3/4 7-35 Amendment No. ~1, lP~: lt2/.'
REFUELING OPERATIONS REFUELING OPERATIONS 3/4.9.13 SPENT FUEL POOL - REACTIVITY LIMITING CONDITION FOR OPERATION 3.9.13 The Reactivity Condition of the Spent Fuel Pool shall be such that keff is less than or equal to 0.95 at all times. APPLICABILITY: Whenever fuel assemblies are in the spent fuel pool. ACTION: With kerr greater than 0.95:
- a. Borate the Spent Fuel Pool until keffis less than or equal to 0.95, and
- b. Initiate immediate action to move any..wel assembly which does not meet the requirements of Figures 3.9-1, 3.9-}~9-4, to a location for which that fuel assembly is allowed. l C5Y 3- <<:t. 5 SURVEILLANCE REQUIREMENTS 4.9.13.1.1. Ensure that all fuel assemblies to be placed in Region 1 "4-0UT-OF-4" fuel storage are within the enrichment and bumup limits of Figure 3.9-1 by checking the fuel assembly's desi~ and bum-up documentation.
' c:l~~ ,~(" I 4.9.13 .1.2.
~
"e..(~")
~
Ensure tha ' all fuel assemblies to be placed in Region 2 fuel storage are within the hN1~ enrichmen and burnup limits of Figure 3.9-3 by checking the fuel assembly's
- desi~and bum-up documentation. . \ ' p~ - ve~~ t.J \~flQL.)'cll l&sttJ ~cJ~~,,,e. ."V) - .... £ST J tt;7\-
r . t 1'\,1',., CI, 4.9.13.1.3. Ensure that all fuel assemblie~o be placed in Region 3 fuel storage are within the enrichment, decay time, and burnup Iimi ts of Figure 3.9-4 by checking the fuel assembly's design, decay time, and burn-up documentation. E'I(l>..s~ L ~ ~~ q\l
~~~. "S'~~~ \::>\\,>> u scJ '~ (fla.s\t:--1I;l ~1(~~.(.", (3. '-£0 """'1<.1 COh<l.~",
whtc,"" I , £ a ~'YJ' e ~i~_ ~,,,(,~ ',~.' ~e'IO'~ .3 ~e.t. sf,~ ~ y~ I W i,~,~ ,",-<- ~~')!"l (It\~e,,~ cA e '('Cl\~ +iV'f.\~ ~ ~ ~J. \>"",y,~,-~
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MILLSTONE - UNIT 3 3/49-16 Amendment No. 3-9, .g.g.,/f11
~EP.1L.~C6 ~r(f1~ tJt::)(\f rf~Gfi=- November 28, 2000 FIGURE 3.9-3 Minimum Fuel Assembly Burnup Versus Nominal Initial Enrichment for Region 2 Storage Configuration 35 -r----~i-----i-----+------f------f--/---#~ 30 1-----t----~-t-----+_-----f--~~--f--------1 ~ -r-------t---+~~~---+------~
--o
- 2:.
~ ~ -;;:: 20 t---"'~---+------+-~=---/--+----=*--I-----+---:r---l
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15 t-----+-~~-_r_-t-----+_----+-A----",L+---------4 5t--r-----t------:1rL--+------+--~ 0+---f--+----+-----+----+---~~----_1 2.0 2.5 3.0 3.5 4.0 4.5 50 Initial Fuel Enrichment ( w/o U-235 ) ILlSTONE UNIT 3 3/ -20 Amendment No.189
IYclli V\; m,1o\: ~ ~~ ~ ss!!"b '7 Bu. yY\ uti' 01 ~iJ De:" +J""'l~ Ve'f5!is Nort\IV\~ \\ -.1:~\4iQ.l ~. \'\ ~I e1h. """ ( \')'~ -$a Y (2~~CJL;1 ~ .s+a~~£. CG""~'eCif yC(.lri ~ ". Figure 3.9-3 Minimum Fuel Assembly Burnupkersus Noml1121 Initial Enrichment.' If.or Region 2 Storaqe Confi ~atiorr - 45 IrTTTlrrTlrrTT~rrTl-rTl,-rr--rT"'--T1-;-r-r--:-""---r--I
- ? t'---~ -fF~,~ +H-
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++:V-I 0 years 1H--rH-r--:-t-+++-'-t'-t-+-f-+++-T-++-H-..j'ri--t--h,-+-+-wl.J,L/W 40 I ! ,I I .,r. 1_ -; 1_1- -f-!- -I-l'--- -~l}-V
=r- -,- ---- __
1--1-->-- , --1-1-1-- --f ---_. -~- __ ~ 1/-1-/ -- V I-5 years 10 years 35 -t-+-----~ - --_. - I ~FC12 ;IL~ I-l.- -1-- -- - --I-i-+-+---Hf--+-I--l-I*-!--+--+ I1 _1--1 I-- _ .of-.f-I-- . _. - - - I - - - I - _J- .--hlI-A'_-1-.j....-"-_,
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') c -rH-rt-1H-i-H+f'-~~-++H+-H-f--WH-I/;I/iL+-t--+-t++-W..+-W 1- 1//1
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'- -- - - - _ _ -- ,/ -I- - - -
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- I~ I
__'__ ']' * -'I--t-f--I-f--I--I-+-t-t--l-I-l--+--l-I 1- - --1-1--1--1-1 Ii-I-' "If'r t-H-+~-t-I-+-H-++-I-++-H--1--l-I-J--+-I-+.....j +-+--1--1
-: i V +-1--1-11 1-+--+--1-1--+--1-+ - - I - -
-, ttA - 1 - -
O~...J-lL-I--f-L-l-L..J...+-L....L-L.L.-f-L--LL.L+-L--LL.L.f-.l-.LLL-W-.LLLJ ISO 2.00 2..50 300 350 4.00 4.50 5.00 Initial Fuel Enrichment (w/o U-235) M\L-L.~rONr; UNIT3
FIGURE 3.9-4 Minimum Fuel Assembly Burnup and Decay Time Versus,Nominal , (\ Jnitial Enrichment for Region 3 Storage Configuration "ffvy*,Ar.ss~t"\'\'bl1¢.s. I
..fYO\'\"\ e- \j ... '" ':3 41\ ttlU> 1.{0 )~o.:re,~ --_-_-_,
60 , - - - - - , - - - - - , - - - - - , - - - - - , - - - - - , - - - - - , 50 +-----+-----+-----+-----+-----+-;'----0 ACCEPTABLE DOMAIN
- J I-
~
o --+- 0 year decay time l'J -{}- 5 year decay time
~ 30 +----4-----4---,L-hL.--r-----r+-----+------1
- a. - . - 10 year decay time
- J
....c ~ 20 year decay time
- J m
~
- J LL 20 +-----+-+-7'~"-+/-;~---1r_---t----+---___j 10 ,J..Hc"L----+-----f------jr_---+-----+------1 O+-----f------+----\------+------t------j ...J'"
2.00 2.50 3.00 3.50 4.00 4.50 5.00 I Initial Fuel Enrichment ( w/o U-235 )
\J MILLSTONE - UNIT 3 3/4 9-21 Amendment No~1
FIGURE 3.9- Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 3 Storage Configuration-
~y ",~,<<mb.'e_~ ,f'yot'\'!) tlJJct--lJ~D'ef4r~ (3' $o)Yil~*{;) Cure..
1 60
-+ I I
I I
+ -~.
o years f
=1-
-7 I - -~
5 years 50 I , I /1 10 years I
.~- V /1 15 years
-+- ---i-
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_i
- ~
/
7-V~ 20 years I I Ih ~ V; ~~ ~ I I I ~ ~V
'F~'-
~
J , ..... IUNACCEPTABLE I I .- f-- c - - - -I- :- ~- 10 o "I . 1.00 1.50 2.00 2.50 3.00 3.50 4.00 4.50 5.00 Initial 235 U Enrichment (w/o) t:S~~;;'~ MILLSTONE* UNIT 3 3/49-2..ltt" Amendment No.~~_
DESIGN FEATUR S 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuels torage racks are made up of 3 Regions which are designed and shall be maintained to ensure a Kef! less than or equal to 0.95 when fl ooded wi th unborated water. The storage rack Regions are:
- a. Region 1, a nominal 10.0 inch (North/South) and a nominal 10.455 f inch (East/West) center to center distance, credits a fixed neutron absorber (BORAL) within the rack, and can store fuel in 2 storage configurations:
(l) With credit for fuel burnup as shown in Figure 3.9-1, fuel may be stored in a "4-0UT-OF-4" storage configuration. (2) With credit for every 4th location blocked and empty of fuel, fuel up to 5 weight percent nominal enrichment, regardl ess of fuel burnup, may be stored ina "3-0UT-OF -4" storage configuration. Fuel storage in this configuration is subject to the interface restrictions specified in Figure 3.9-2. cq-vJ ~d cJ(~~'J ~~
- b. Region 2, a nomi al 9.017 inch center to center distance, credits a fixed neutron bsorber (BORAL) within the rack, and with credit for fuel burnup as shown in Figure 3.9-3, fuel may be stored in c.
all available Region 2 storage locations. Region 3, a nominal 10.35 inch center to center distance, with credit for fuel burnup and fuel decay time as shown in Figure 3.9- ' \
~eel may be stored in al available Region 3 storage locations.
l i
~ Boraflex conta~ned inside these storage ;acks is not credited.
, ~iJ"Y G:\S~~""'b\l~ tAsc.A Q~ C-~4~\\V(.\t \ - * 'i>~-Y-f?Yqt~ (34-umWb)CO~
DRAINAGE (.>"'( ~'i ~4Y~ 3*<1 -'$ ..relY ~s~~n.,.\'(;~1 ~e.J II",", ~s;~-U:fy~t~ (<i~S"(;oIl\1 w*>>cG'>eS 5.6.2 The spent fuel storage pool is design and shall be maintained to -e >J prevent inadvertent draining of the pool below elevation 45 feet. 5-6 Amendment No. ~~. J~, ~
JUL-03-2007 13:06 P.08 ADMINISTRATIVE CONTROLS PROCEOURES AND PROQRAMS (CQntinueg)
- 2) Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and
- 3) Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.
- f. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option 8, as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,!> dated September 1995, as modified by the following ex.ception to NEI 94-01, Rev. 0, "Industry Perfonnance Based Option of 10 CFR Part 50-Appendix 1": The first Type A test performed after the January 6, 1998 Type A test shall be performed no later than January 6, 2013.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa' islJrnpsig.
-c.:... 4 /-4 The maximum allowable containment leakage rate La' at Pa' shall be OJ percent by weight of the containment air per 24 hours.
Leakage rate acceptance criteria are:
\
- 1) Containment overall leakage rate acceptance criterion is S 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type Band
\ Type C tests, and ~ 0.06 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests;
- 2) Air lock testing acceptance criteria are:
- a. Overall air lock leakage rate is ~ 0.05 La when tested at ~ Pa' b, For each door, seal leakage rate is < 0,01 La when pressurized to c:: Pe.
The prqvisions of Specification 4.0.2 do not apply to lhe It:si. frel.! ucnl,;ies specifi~d in the Containment Leakage Rate Testing Program. The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program; .- -
- An exemption to Appendix J, Option A, paragraph IIl.D.2(b)(ii), of) 0 CFR Part 50, as approved by the NRC on December 6, 1985.
MILLSTONE - UNIT 3 6-17 Amendment No. 69, +S6, ~, #1
4villl'6lr"T4, 2007 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.) 6.9.1.6.b The analytical methods llsed to detelmine the core operating limits shall be those previollsly reviewed and approved by the NRC in: I. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," (W Proprietary). (Methodology for Specifications 2.1.I.l--Departure from Nucleate Boiling Ratio, 2.1.1.2--Peak Fuel Centerline Temp~rature, 3.1.1 J--Moderator Temperature Coefficient, 3.1.3.5--Shutdown Bank Insertion Limit, 3.1.3.6--Control Bank Insertion Limits, 3.2.l--AXIAL FLUX DIFFERENCE, 3.2.2--Heat Flux Hot Channel Factor, 3.2.3--Nuc1ear Enthalpy Rise Hot Channel Factor, 3.1.1.1.1,3.1.1.1.2,3.1.1.2 -- SHUTDOWN MARGIN, 3.9.1.1-- Boron Concentration.)
- 2. T. M. Anderson to K. Kniel (Chief of Core Perfonnance Branch, NRC), January 31, I980--
Attachment:
Operation and Safety-Analysis Aspects of an Improved Load Follow Package.
- 3. NUREG-800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981 Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Revision 2, July 1981.
- 4. WCAP-I 0216-P-A-R 1A, "RELAXATION OF CONSTANT AXIA L OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION,"
(W Propnetary). (Methodology for Specifications 3.2.1--AXIAL FLUX DIFFERENCE [Relaxed Axial Offset Control] and 3.2.2--Heat Flux Hot Channel Factor rW(z) survedlance requiremen,ts for FO Methodology]'l Q, \--E::-~' .h.\.Qc..A *
,l).)(:A~-~2-~4J 5-~-~ IrC-04e. ~\.I4l\\+;c"".\.;c>"" 1)o~~'fnof...".~..f.,,,,* .D'e~ ~... Q.tA ,A~\'1~".1
- 5. (WC -9561-P-A, ADD. 3, "BART A-I: A COMPUTER CODE FOR THE BES'[ff\p{?l'iIW~~l?1)
. ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS--SPECIAL REPORT ~" -
THIMBLE MO EUN " Pro rielaJ 'I ( etlo ology for Specification 3.2.2--Heat Flux HotChannel Factor) 6.
~gt-0W~SG~~ ft1s'iv~ii ;1T~6~9~~bi~'0~U$ANRASH CODIJ (W Proprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel \. \-:
j I Factor.) WLAP t' a~e,*:: (>-~:l' II t!~~t~kiG 1=Ay~e -~"'f~k t.,Q(A E~ ':~io1n
'\'n~~ o,/i¥lI ~'1 \.) s. \~" ~ \A;+.D~t."(..J c;.~.r,\\-1.(S.qt<.:l\' 1S' t~~.~ ...,j..- CI1' I~
- 7. WCAP-II ~46, "Satety Evaluatl6n Supportll1g a More Negabve EOL ~V'\(:c.y~~ \"\\{ 'T (~
Moderator Temperature Coefficient Technical Specification for the Millstone ~AlS:ji~*-,um)i,' Nuclear Power Station Unit 3," (W Proprietary).
- 8. WCAP-I0054-P-A, "WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL 17 USING THE NOTRUMP CODE," (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 9. WCAP-I 0079-P-A, "NOTRUMP - A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 10. WCAP-12610, "VANTAGE+ Fuel Assembly Report," (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.) MILLSTONE - UNIT 3 6-20 Amendment No. ;M, H, eG, 69,..~.: HG, -H-B, M, H:9,j)o(}}