B12604, Proposed Tech Spec Changes Supporting Cycle 15 Operation
| ML20236D597 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 07/31/1987 |
| From: | CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | |
| Shared Package | |
| ML20236D556 | List: |
| References | |
| B12604, NUDOCS 8707300542 | |
| Download: ML20236D597 (105) | |
Text
{{#Wiki_filter:_ _ _ - _ _ _ _ _ - - - - _ - _ - - . l l Docket No. 50-213 l B12604 I J 1 l l l l Attachment 2 l Haddam Neck Plant l Resubmittal Proposed Technical Specification l Changes to Support Cycle 15 l l l July 1987 hok k 'g3 p P
1 t i f -DEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated e-signal into the channel as close to the primary sensor as prac-ticable to verify OPERABILITY including alarm and/or trip functions. CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any component ,' within the reactor pressure vessel with the vessel head removed and l fuel in the vessel. SHUTDOWN %RGIN
- l'.13 SHUTDOWN MARGIN shall be the instantaneous. amount of reactivity by which the reactor is suberitical or would be subcritical from its !
present condition assuming all rod cluster assemblies are fully l inserted except for the single rod cluster of highest reactivity I worth which is assumed to be fully withdrawn. IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAXAGE shall be: l
- a. Leakage except CONTROLLED LEAKAGE into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank l or .
l
- b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of UNIDENTIFIED LEAKAGE monitoring systems !
or not to be PRESSURE BOUNDARY LEAKAGE. ! 13 IDENTIFIED LEAKAGE' 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which has not been identified. I 1 I 1 j 1 1-3 ww_m+___m_--_______________m_________. _ _
- - ~ - -
)
!' ) TABLE 1.1
)i OPERATIONAL MODES i
REACTIVITY % RATED AVERAGE COOLANT ) MODE CONDITION, Keff THERMAL POWER
- TEMPERATURE 1
, I
- 1. POWER OPERATION >0.99 >5% >350 F 3
- 2. STARTUP > .99 55% >350*F 4
3: HOT STANDBY < .99 0 >350*F 1
- 4. HOT SHUTDOWN < .99 0 350 F > T avg > 200 F a i
A
- 5. COLD SHUTDOWN < .99 0 $200 F
- 6. REFUELING ** -1 94 0 $140 F
- Excluding decay heat.
** Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
i i 1-6 i
)
SECTION 2 , SAFETY LIMITS AND MAXIMUM SAFETY SETTINGS
2.1 INTRODUCTION
Safety Limit's are defined in order to protect the fuel cladding and the reactor coolant system. The integrity of these barriers must be maintained to prevent an uncontrolled release of radioactivity.
, Faximum Safety Settings are also established for protective devices related to the process variables on which the safety limits are based. The maximum safety settings are chosert such that protective action will prevent the safety limits from being exceeded.
2.2 S_AFETY LIMITS REACTOR CORE 2.2.1. The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop inlet temperature (T eo shol.1 not exceed the limits shown in Figures 2.2-1 andd) 2.2-2 for four and three loop operation, respectively. APPLICABILITY: MODES 1 and 2 ACTION: , Whenever the point define'd by the combination of the highest operating loop inlet temperature (Teol ) and pressurizerpressurehasexceededtheappropr1Ntepercent of rated thermal power (RTP) line, be in HOT STANDBY within 1 bour, and comply with the requirements of Speci-fication 6.7.1. 2-1
1 610 - * '
- 6. 105 goo .
M1 90: 590 goo: 580 1101 i I
. g 570 -
i w 118%
'E 5- 560 -
.g 1251 W~
. g 550 -
l 1
.j 540 530 -
'l
)
$20 -
1 510 ' * * * ' ' 8 1600 1700 1800 1900 2000 l 2100 2200 2300 1 1 PRES $URE. PSIA I FIGURE 2.2-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION k
63tt-i.
, g ens 620 -
50s 'i 610 - .
.- 60 l
- i l
600 - 70'.
$90 ..
s0: :
.$80 -
90% 570 - p
, E - 5s: -
9 C i g- 550 - - i 1.
, MO 530 -
520 - 510 e i n e i e i
. 1600 1700 1800 1900 2000 2300 2200- 2300 PRES $URI, PSIA j
. j l
FIGURE 2.2 2 l 1 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION, I 1
C. . J, i BASES-
'2.2.1" ' REACTOR CORE-l 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 by restricting-fuel operation'.to within the nuclaste boiling' regime where the heat transfer: coefficient is large and the cladding surface temperature'is slightly above the coolant saturation temperature.
'I Operation above.the upper boundary of the nucleate boiling- !
regime.could result in excessive cladding-temperatures J because.'of the onset of departure.from nucleate boiling .j
.(DNB) and the esultant sharp reduction in heat transfer (
coefficient. DhB is not a directly measurable' parameter l during operation and, therefore, THERMAL POWER and Reactor ' Coolant Temperature and Pressure have been gelated to DNB- = th' rough the W-3 correlation. The W-3 DNB correlation has been developed to predict the.DNB flux and the location of .. DNB for.. axially uniform and' nonuniform heat flux distri- '
- butions. The local DNB heat flux ratio (DNBR) is defined '
as the ratio of'the heat flux that would cause DNB at a- *
,particular core location'to the local heat flux, and is I indicative of the margin to DNB. ..
The minimum value of the DNBR during steady state: operation, normal operational transients, and anticipated transients-is limited to 1.30. This value corresponds to a 95 percent ; probability at'a 95 percent confidence level.that DNB will j not ocent and is chosen as- an appropriate margin to DNB i for all ope.ating. conditions. l The. curves of Figures 2.1-1 and 2.1-2'shov the loci of ) points'of THERMAL POWER, pressurizer pressure and. core inlet temperature for which the minimum DNBR is no less i than 1.30, and the core outlet. void fraction is'no greater than 0.32. These curvgs are based on nuclear enthalpy hot channel factors, F" , of 1.60 and 1.64 for four loop and three loopoperaNon,respgetively. An allowance is included for an increase in F"g at reduced power. These. limiting hot channel factors are higher than those calculated at full power for the range from all control rods fully withdrawn to at ximum allowable control rod J insertion. This insertion limit is described in
' Specification 3.10.2.6 and 3.10.2.7. The required reduction in power. level as dictated by Figures 3.10-1 and 3.10-2 insures that the DNB ratio is always greater than 1.30.
i l 2-2 _=_ .
-l
-l Intentionally Lef t Blank i
i l 2-3
h ,
}
2.4 MAXIMUM SATETY SETTINGS - PROTECTIVE INSTRUMENTATION-
-Applicability: Applies'to trip settings for instruments monitoring.
reactor power and reactor coolant pressure, temperature, and. flow. - 0bject'ive: To provide.for protective action in the event that the principle process variables approach a safety limit. Specification: Protective instrumentation trip settings'shall
~
. be as follows:
Four Reactor Coolant 'Three Reactor Coolant Pumps Operating Pumps Operating (1) Pressurizer Pressure. 12300 psig $2300 psig (2) Pressurizer Level *- 186% of. range 186% of range (3) Variable Low Pressure *** >17.4 (T*V8 + 1.17AT) >17.4
-8850 (T "8+1.174T)
~.8850 (4) Nuclear Overpower ** $109% of rated power 174% of rated power,
~
(5) Low Coolant Flo @ M >90% of nominal four >84% of nominal j loop flow three loop flow i
'(6) Reactor. Coolant Loop $20*F $20*F Valve-Temperature ;
= Interlock. l (7) High Steam Flow 110% of full load 110% of full load ,
steam flow steam flow !
-(8) High Startup Rate **** 15 decades per minute 15 decades per minute
*May be bypassed when the reactor is at least 1.5 %k suberitical.
. l l
**The nuclear overpower trip is based upon a symmetrical core power distribution. When the reactor power is (10% the overpower ~ trip
~
setpoint is reduced to 25 percent of rated power
***May be bypassed below 10 percent of rated power. !
****Not required if the reactor trip breakers are open or the control !
rod drive' lift coils t.re de-energized; may be bypassed above 1 10 percent of rated' power. Basis: The reactor protective system is designed and constructed such that no single failure in any of the instrument systems will prevent the desired safety action if an applicable parameter exceeds a safety setpoint. . 2-5 l _______-______--____d
and shutdown. It is safe to block this trip below 10 percent power since the. protection afforded by_this_ trip is not required at this low level. Removal of unnecessary trip signals will reduce the number of spurious trips.
'(4) Nuclear Overpower
# As explained above, the nuclear overpower reactor trip, in conjunction'with the . ;
variable low pressure reactor trip, provides j overpower, overtemperatura protection. The nuclear overpower trip cha. nels will respond first to rapid reactivity insertion rates, detected by the increase in flux, before there are any significant changes in the system process variables. -A caximum error of 9 percent of full power due to setpoint, instrumentation, and calorimetric determination (see Section 4.3.6 of the FDSA) is considered in establishing the setpoint. In order to reduce the time to trip for certain accidents. occurring at low s power, the overpower setpoint is -lowered to 25 percent when reactor power is below 10 percent. This low overpower trip would ! terminate the postulated large steamline ! break accident from the hot zero power l condition. The lower setting for_three i loop operation provides protect' ion at the ; reduced power level equivalent to'that l provided by the setting for four loop i operation at full power. (5) Low Coolant Flow a The low coolant flow reactor trip protects j the core'against an' increase in coolant temperature resulting from a reduction in l coolantflowwhily3 e reactor is at ! substantial power This trip will prevent DNB in any loss-of-flow incident, which eliminates the possibility of clad damage. Flow detection in each reactor {' coolant loop is from a measurement of pressure drop from inlet to outlet of each steam generator. The 90 percent and ! 84 percent low flow signals.are high enoagh : to activate a trip in time to prevent DNB, and low enough to reflect that a loss-of-flow condition truly exists. A maximum instrument and setpoint error of 5 percent full flow is considered in determining the setpoint. Loss-of-flow protection is also provided by reactor coolant pump breaker and from undervoltage 2-7
(- - - - - on a reactor. coolant pump motor bus. This trip signal may be defeated below 10 percent power since at this level, natural circulation, if need be, could cool the core. Spurious trips can be eliminated by defeating the signal. 0 (6) Reactor Coolant Loop Valve -- Temperature ' Interlock The reactor coolant loop valve-temperature interlock prevents the return to service of an isolated loop, whose temperature is substantiallybelowthehighestco{g) leg temperature of the operating loops . The setting of 20 F (plus 10 F instrument and set point error) limits the positive reactivity insertion that could occur due - to admission of cooler water to the reactor coolant system to a value below that which ; would result in DNB. : (7) High Steam Flow Thiscircuitprovidesprgggetionagainsta large steam line rupture . This signal t closes the main steam line isolation valves : and trips the reactor, thereby limiting the ! cooldown. An error of 5 percent full steam flow is used in determining the trip settings. 1 (8) High Startup Rate The Intermediate Range High Startup Rate .I trip provides core protection during reactor startup. This trip function provides protection for large reactivity insertion events initiated from a suberitical mode of operation. This trip function is credited in the rod withdrawal from suberitical analysis.
References:
(1) FDSA - Section 7.2 (2) FDSA - Section 10.3.5 (3) FDSA - Section 10.3.2 (4) FDSA - Section 10.2.2 (5) FDSA - Section 10.3.3 l l 2-8
l' ". f, l 3.3 REACTOR COOLANT SYSTEM 3.3.1.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION
- 3.3.1. I ' The'following reactor coolant loops sha!! be in operation, with the associated loop stop valves OPERABLE:
- a. All reactor coolant loops in operation with the reactor above 65%
of RATED THERMAL POWER, or-
- b. At least three coolant loops in operation
- with the reactor less than or equal to 65% of RATED THERMAL POWER.
APPLICABILITY: MODES I and 2. ACTION: s
' With less than the above reauired reactor coolant loops in operation'or the associated loop stop valves not OPERABLE be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENT
- 1. At'least'once per 12 hours the above required reactor coolant loops shall
' be verified to be in operation and circulating reactor coolant, and that power is available to the loop'stop valves.
- 2. At least once per 18 months, cycle the loop stop valves through one 4 complete cycle of full travel.
I i
* . The loop out of service may be idled (cold jeg stop valve closed) or isolated (cold and hot leg stop valves closed). -
3-3 i
s , . p"' % 4 > 1 -.
- l. , p-[' t m
; e
,, o :i c .
' REACTOR' COOLANT SYSTEMU ,
HOT STANDBY - ,
~ LIMITING CONDITION FOR OPERATION i 3.3.1.2 The following number of Reactor Coolant Loops listed below shall be .
, OPERABLE and in operation.*-
- , a.. . At least three reactor coolant loops shall be OPERABLE and at least two
- reactor coolant loops shall be in operation if.the reactor trip breakers j
< are closed and the control rod drive lif t coils energized, or !
- b. At least two reactor coolant loops shall be OPERABLE and at least one. -
i reactor coolant loop shall be in operation if the reactor trip breakers are open or the control rod drive lif t coils are de-energized. The reactor coolant loops are;
- a. - Reactor Coolant Loop 1 and its associated steam generator and ,
reactor coolant pump, , q
- b. Reactor Coolant Loop 2 and its associated steam generator and : '!
reactor coolant pump, 1
- c. Reactor' Coolant Loop 3 arid its associat.ed s' team generator and i
, reactor coolant pump, and !
- d. ' Reactor Coolant Loop kand its associated steam generator and reactor coolant pump.
i APPLICABILITY: MODE 3. - ,; 1 ACTION: a. With less' than the above required reactor' coolant loops OPERABLE, .; restore the required loops to OPERABLE status within 72 hours or be in. ! HOT SHUTDOWN within the~ next 12 hours. : 3 j
- b. .With only one reactor coolant loop in operation and the reactor trip '
system breakers closed and the control rod drive lif t coils energized, within I hour either open the reactor trip system breakers or de-energize .l the control rod' drive lif t coils. !
- c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System,immediately open or verify open the reactor trip system ;
breakers, and initiate corrective action to return the required loop to operation. - ;
-* All reactor coolant pumps may be de-energized for up to I hour provided:(1) no operations are permitted that would cause dilution of the Reactor Coolant System
. boron concentration, and (2) core outlet temperature is maintained at least 100F
. below saturation temperature.
3-4
p, . 7
. REACTOR COOLANT SYSTEM-
' HOT STANDBY .'
L SURVEILLANCE REQUIREMENTS l l . a.. At least once per 7 days the above required reactor coolant pumps,if not in operation, shall be determined OPERABL".
- b. At least once per 12 hours the required steam generators shall be determined OPERABLE.
- c. At least once per 12 hours the required reactor coolant loops shall be l verified in operation and circulating reactor coolant.. '
- d. At least once per 12 hours, if required, verify that the reactor trip ;
system breakers are open or the control rod drive lif t coils are de-energized. c , l l A R I I I t e 3-4a
l REACTOR COOL ANT SY'Si'EM - HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.3.1.3 The following number of heat removal loops listed below shall be OPERABLE + and in operation:
- a. At least three reactor coolant loops shall be OPERABLE and at least two reactor coolant loops shall be in operation if the reactor trip breakers are closed and the control rod drive lif t coils energized, or
- b. At least two heat removal loops (RCS or RHR)+
- shall be OPERABLE and at least one heat removal loop (RCS or RHR) shall be in operation if the reactor trip breakers are open or the control rod drive lif t coils are de-energized.
The heat remo/al loops are;
- a. Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,"+
- b. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,"*
- c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,"*
- d. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump,"*
- e. RHR loop A, and
- f. RHR loop B.
, APPLICABILITY: MODE 4.
ACTION: a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status and open or verify open the reactor trip system breakers within I hour. If . the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN' within 24 hours. i
- b. With no loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System, immediately open or verify open the reactor trip system breakers and initiate corrective action to return the required loop to operation.
+
All reactor coolant pumps and RHR pumps may be deenergized for up to I heur provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant Systern boron concentration, and (2) core outlet temperature is maintained at least 100F below saturation temperature.
*
- An RHR loop consists of a Residual Heat Removal (RHR) pump, a dedicated RHR heat exchanger either from the same train or from the opposite train and all other necessary piping and components required to receive and cool reactor coolant.
* * *A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg. temperatures less than or equal to 3150F unless the secondary water temperature of each steam generator is less than 200F above each of the Reactor Coolant System cold leg temperatures.
3-4b
+
< s l
, 9
!I
' REACTOR COOLANT SYSTEM -
SUR'YEILL ANCE REQUIREMENTS
' a.
.. . At least once per 7 days the required reactor coolant pump (s), if not in ;
operation, shall be determined OPERABLE. j
- b. '
' At least once per 12 hours the required steam generator (s) shall be i
' determined OPERABLE by verifying secondary side narrow range water
. level to be greater than or equal to 25%
- c. At least once per 12 hours one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant.
- d. At least once per 12 hours,if required, verify that the reactor trip system breakers are open or the controi rod drive lift coils are de-energized.
i
}
}
.)
i
'l l
1 a I,1 3-4c j
m- - g, , c
+
'RE ACTOR COOLANT SYSTEM t r
~
COLD SHUTDOWN - LOOPS FILLED. , LIMITING CONDITION FOR OPERATION i 3.3.1.4.1 - At least one RHR loop shall be . OPERABLE and in operation,* the j reactor trip system breakers shall be open or the control rod drive lif t coils shall be de-energized, and either:
' a'. n ' One additional RHR loop shall be OPERABLE, or
- b. The secondary side narrow range water level of at least two unisolated ll steam generators shall be greater than 25%
j i APPLICABILITY: . MODE 5 with reactor coolant Joops filled.** A CTION: .
- a. ' With one of the R'HR loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return
, 4 the inoperable RHR. loop to OPERABLE status or restore the required steam generator water level ~as soon as possible.
- b. With no RHR loop in operation, suspend all operations involving a i reduction in boron concentration of the Reactor Coolant System and
. immediately initiate corrective' action to return the required RHR loop to operation. -
- c. . With the reactor trip system breakers closed and the control rod drive lif t coils energized, within I hour either open the reactor trip system breakers or de-energize the control rod drive lif t coils. .
' \
SURVEILLANCE REQUIREMENTS ~
- a. . At least once per 12 hours the secondary side water level of at least two f
. steam. generators when required shall be determined to be within limits. (
\
At least once per 12 hours,'an RHR loop shall be determined to be in j b. j operation and circulating reactor coolant. - J
- c. The RHR loop not in operation but required shall be determined OPERABLE at least once per 7 days by verifying breaker alignments and indicated power availability.
- d. At least once per 12 hours verify that the reactor trip system breakers are open or the control rod drive lift coils are de-energized.
- The RHR pump may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 100F below l l
Saturation temperature. l
*
- A reactor coolant pump in an unisolated loop shall not be started with one or mere of'the Reactor Coolant System cold leg temperatures less than or equal to 3150F unless the secondary water temperature of each steam generator is less than 200F l above each of the Reactor Coolant System cold leg temperatures. l 1
4,,P 3-4d
y ,
.c ,,.
y 4a mz x 3.
~ .
REACTOR COOLANT SYSTEM ~ r.
,
- COLD SHUTDOWN ' LOOPS'NOT FILLED LIMIk!NG CONDITION FOR OPERATION - ,
.3.3.1.4.2 Two'RHR loops shall be OPERABLE
- with at least one RHR loop in operation ** and either the reactor trip system breakers shall be open or. <
J! the control rod drive lif t coils shall be de-energized. APPLICABILITY: MODE 5 with reactor coolant' loops not filled.
- ACTION:
- a. .
With less than the above required RHR loops OPERABLE, immediately.
~ initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
b'. .With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately. initiate corrective action to return the required RHR loop. to operation.
- c. With the reactor trip system breakers closed and the control rod drive lift coils energized, within I hour either open the reactor trip system
- breakers or.de-energize the control rod drive lif t coils.
SURVEILLANCE REQUIREMENTS -
- a. At least once per 12 hours, a RHR loop s' hall be determined to be in operation and circulating reactor coolant.
- b. . At least once per 7 days the RHR !.oop not in operation shall be determined OPERABLE by verifying breaker alignments and indicated -
power availability. W '
- c. At least'once per 12 hours verify that either the reactor trip system
,, ' breakers are open or the control rod drive lift coils are de-energized.
.One RHR loop.may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
** The RHR pump may be de-energized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 100F below .
saturation' temperature. 3-4e
~- ~- ~ ~
REACTOR COOLANT SYSTEM ISOLATED LOOP LIMITING CONDITION FOR OPERATION l 3.3.1.3 The RCS loop stop valves of an isolated loop
- shall be shut and either:
- a. The power removed from the valve operators, or
- b. The boron concentration of the isolated loop shall be maintained greater than or equal to the boron concentration of the operating loops.
APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: With the requirements of the above specification nct satisfied, either:
- a. Retnove power from the valve operators within one hour, or *
- b. Increase the boron concentration of the isolated loop to within the limits within 4 hours, or
- c. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS
- a. At least once per 12 hours,if required, verify that power is removed from the valve operators.
- b. At least once per 12 hours, if required, verify that the boron concentration of an isolated loop is greater than or equal to the boron concentration of the operating loops.
- A loop is considered to be isolated when the hot and cold leg stop valves are both closed.
s
- 3-4f j ..
WAf },
~
Q , ll[ l 2 p
. REACTOR ' COOL ANT SYSTEM -
IISOLATED LOOP STARTUP' u- LIMITING CONDITION FOR OPERATION?
.).3.1.6 ? A reactor' coolant loop shall remain isolated until:
i- a. The temperature at the cold leg of the isolated loop is within 200F of the highest cold leg temperature of the operating loops l .
- b. The boron concentration of the. isolated loop is greater than or equal to the boron concentration of the operating loops, c.. The reactor is subcritical by at least 1000 pcm.'
APPLICABILITY:' MODES 3, 4, 5 and 6.
~ ACTION .
With the requirements of the above specification not satisfied, do not open the
? isolated loop stop valves.
' SURVEILLANCE REQUIREMENTS
, .a.- The !solated loop cold leg temperature shall be determined to be within 200F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve.
- b. The reactor shall be determined to be subcritical by at least 1000 pcm within 30 minutes prior to opening the cold leg stop valve.
- c. Within 30 minutes prior to opening the loop stop valves, the isolated loop shall be determined to have a boron concentration greater than or equal-to the boron concentration of the operating loops.
'e e
3-4g _ ._.__m...___.___ - . _ _ _ ._ ,___.______ . _ _ _ _ _ _ _ _ _ __
y; REACTOR COOLANT SYSTEM IDLED LOOP - LIMITING CONDITION FOR OPERATION i
-3.3.1.7 The cold leg loop stop valve 'of an idled loop (s)* shall be shut and either:
a .' The power removed from the valve operator, or i b. The boron concentration of the idled loop (s) shall be maintained greater.than or equal to the boron concentration of the operating loops. APPLICABILITY: ,. MODES 1,' 2, 3, 4, and 5. ACTION:
' With the requirements of the above specification not satisfied, either:
- a. Remove power from the valve ape ..itor within one hour, s
' b. Increase the boron concentra; ion of the idled loop (s) to within the limits within 4 hours, or
- c. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN -
within the following 30 hours. . SURVEILLANCE REQUIREMENTS
- a. 'At least once per 12 hours, if required, verify that powt ' is removed from the valve operators,
- b. At least once per 12 hours, if required, verify that the boron -
concentration of an idled loop (s) is greater.than or equal to the boron
. concentration of the operating loop (s).
- A loop is considered to be idled when the hot leg stop valve is open and the cold leg stop valve is closed.
3-4h
REACTOR COOLANT SYSTEM IDLED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.3.1.8 A reactor coolant loop shall remain idled until:
- a. The temperature at the cold leg of the idled loop (s) is within 200F of the highest cold leg temperature of the operating loop (s),
- b. The boron concentration of the idled loop (s)is greater than or equal to the boron concentration of the operating loop (s), .
c.- The reactor is no greater than 60% RATED THERMAL POWER if
~
only one loop is idled or is suberitical by at least 1000 pcrn if more than one loop is idled. APPLICABILITY: All MODES. ACTION: With the requirements of the above specification not satisfied, do not open the idled loop stop valve. SURVEILLANCE REQUIREMENTS
- a. The idled loop cold leg temperature shall be determined to be within i 200F of the highest cold leg temperature of the operating loops withio 30 minutes prior to opening the idled loop stop valve,
- b. Within 30 minutes prior to opening the idled loop stop valve, the reactor shall be determined to be either:
- a. Less than 60% RATED THERMAL POWER if only one loop is idled, or
- b. Subcritical by at least 1000 pcm if more than one loop is idled.
- c. Within 30 minutes prior to opening the loop stop valve, the idled loop shall be determined to have a boron concentration greater than or equal to the boron concentration of the operating loops.
i 3-41
=' ..'
. REACTOR COOLANT SYSTEM j
. . l 3.3.2 SAFETY VALVES '
. i SHUTDOWN LIMITING' CONDITION FOR OPERATION- 1
'e
- 3.3.2.1 -- A minimum of one pressurizer Code safety valve shall be OPERABLE ' .. l with the lif t setting associated with the OPERABLE Code safety valve 1
, within 3,1% of its design Setpoint*. :
- APPLIC ABILITY: - MODE 4 except when Specification 3.3.4.2 Is applicable.
ACTION: ' With'no pressurizer Code safety valve OPERABLE,immediately suspend all operations ;
- involving positive reactivity changes and place an OPERABLE RHR loop into operation '!
- in the shutdown cooling mode.
SURVEILL ANCE' REQUIREMENTS - s
. No additional Surveillance Requirements other than those required by Specification 4.2.
~!
l
]
I l 1 i1 !
.i. j
-t
- The lif t setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. ,
3-4j , I
.o
v. 1 i jEACTOR COOL ANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION -
' 3.3.2.2' ,' All pressurizer Code safety valves sha!! be OPERABLE with respective lif t settings of 2485 psigi 1%, 2535 psig 1 1% and 2585 psig1 1 % in accordance with their respective nameplates.*
. APPLICABILITY: MODES 1, 2, and 3.
ACTION: With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within . the following 6 hours.: i SURVEILi.ANCE REQUIREMENTS No additional Sur elllance Requirements other than those required by Specification 4.2.
- The lif t setting pressure shall corrc spond to ambient conditions of the valve at nominal operating' temperature and pressure.
l 1 l 1 i 3-4k _________O
REACTOR COOLANT SYSTEM 3.3.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.3.3 The pressurizer shall be OPERABLE with:
- a. Water level within 5% of the programmed level of Figure 3.3-1 during periods when THERMAL POWER is maintained constant,*
uad
- b. At least two groups of pressurizer heaters capable of being powered from an emergency power source and each having a capacity of at least 150 kW.
APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reacto. Trip 5ystem breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS
- a. At least once per 12 hours the pressurizer water level shall be determined to be within its limit.
- b. At least once per 92 days the capacity of each of the above required groups of pressurizer heaters shall be verified.
During periods when THERMAL POWER is being changed the pressurizer water level may be outside the 5% band for periods not to exceed I hour.
, r l
\
l 3-4 L j
\ t . . l .1 i I l
- 50 - - - -
q 1 40 . . _ _ _ . . . . . _ . . . . . . . . . _ H . _ _ .. . . . . . . . _ .. . a . _ . _ . _ _ . . . . _ . .. . . . . . . W
- g. . . . _ .. . .. .. . . . _ . . . .. .. . . . _ .
. . _. 30 , , . _ _ _- ~ . . . . . . - - .. _..
a w . _ _ _ _ . . . . . _ _ _ . _ . . _ . _ _ _ . . . . . . . . . . . . . _ _ . . . . . , 9= - i e . ._. . _ 2 5 . . . _ ~ . - _ _ . . .
- s _
w
.a . . _.. go , . _..._ .. _ . _ _._. .. _ . _ _ . . _ _ . . . . _ .
g
.g.
ha _ . . ..__ ____. _ - 10 . . ._ _ . . _ . _ _ _ .
, . _ . . 0
. .500_ . 52.0 53.5. ... 550._ 562 575 .
Tave. ,F. 4 f
.1
. 1 i
i t l
. FIGURE 3.3-1 ( <
W PRESSURIZER PROGRAMMED WATER LEVEL I s g t. F 1 9
+
I I
~
/. ' ] J L ,
' REACTOR COOLANT SYSTEM 3.3.4 RELIEF VALVES -
LIMITING CONDITION FOR OPERATION 3.3.4.1 . All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. The Setpoint for the PORV's and their block valves shall be greater than or equal to 2325 psig and less than or equal to 2350 psig. The emergency control air supply shall have a minimum pressure of 118 psig. > APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or more PORV(s) inoperable, because of excessive seat leakage, within I hour either restore the PORV(s) to OPERABLE status or close/ verify close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within !
the following 30 hours.' ! i
- b. With one PORV inoperable due .to causes other than excessive seat l leakage or low emergency control air sopply pressure, within I hour either restore the PORV to OPERABLE status or close/ verify close the a'ssociated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or bd in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With both PORV(s) inoperable due to causes other than excessive seat leakage or low emergency control air supply pressure, within I hour either restore each of the PORV(s) to OPERABLE status or close/ verify close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours,
- d. With the emergency control air supply pressure for the PORVs less than 118 psig restore the emergency control air supply to OPERABLE status within 6 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS
- 1. In addition to the requirements of Specification 4.2, each PORV shall be demonstrated OPERABLE at least once per 18 months by: !
a.- Performance of a CHANNEL FUNCTIONAL TEST, and
- b. Operating the valve through one complete cycle of full travel.
1 3-4m j
y . ., REACTOR COOLANT SYSTEM
. SURVEILLANCE REdUIREMENTS-2.- Each PORY block valve shall be demonstrated OPERABLE at least once per_18 months by performance of a CHANNEL FUNCTIONAL TEST.'
3.' : Each PORY block valve shall be demonstrated OPERABLE at l' east once iper_92 days by operating the valve through one complete cycle of full travel unless.the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.3.4.1.
- 4. The control air supply for_ the PORVs'shall be demonstrated OPERABLE.
at least once per 18 months by verifying that the control air supply does
' not drop more than 0.3 ps! in one hour when isolated from Containment Air Supply System.
- 3. The air supply for the PORVs shall be verified by loca.i pressure Indication once every 92 days.
9 4 9 l- , 3-4n l
4;. -..
. :J ;
q 1 ,
.j REACTOR COOLANT SYSTEM E LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION lj 1
' j 13.3.4.2 At least one of the following Low Temperature Overpressure Protection j
, .(LTOP) Systems shall be OPERABLE: =
4 1
. a. Two LTOP ' spring-loaded relief valves (SLRV)'with a lift setting i of 380 psig (13%) with their respective motor-operated isolation valves in the open position, or
- b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 7 squar.e inches (3 inches nominal diameter).
APPLICABILITY: -
~
- MODE.A when the~ temperature of any RCS cold leg of an 'unisolated loop.is less than'or equai.to 3150F. MODE 5 and MODE 6.with the reactor vessel head on. 'The .
'above overpressure protection system shall be placed in service prior to placing the
. RHR system into service and shall rernain in service unless the requirements of-
' Specification 3.3.4.2.b are satisfied.
ACTION: . . a'. - With o_ne LTOP SLRV inoperable, restore the inoperable LTOP
' SLRV to OPERABLE status within 7 days, or depressuriz,e and vent the RCS through at least a 7 square inch vent within 8 hours.
~
- b. With bot.h LTOP SLRVs inoperable, depressurize and vent the RCS through at least a 7 square inch vent within 8 hours.
- c. In the event either the LTOP SLRVs or the RCS vent (s) of
~
Specification 3.3.4.2.b are used to mitigate an RCS pressure - <; transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.' The report shall describe the circumstances initiating the transient, the effect of the LTOP SLRVs or the above RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence. 3-4o h - - .- a --,,---.w_--,u--.sn.u----wa--_--,,--,--,,----
c, .. .. [s; ,
,4 REACTOR COOLANT SYSTEM '
SURVEILLANCE REQUIREMENTS' i i
- a. ; Each LTOP SLRV shall be demonstrated OPERABLE by verifying the -
i SLRV isolation valve is open at least once per 31 days when the LTOP SLRV is required to be OPERABLE.
- b. The RCS vent (s) of Specification 3.3.4.2.b shall be verified to be open at
.least once per 31 ' days when the vent (s) are being used for overpressure
.' protection. .
9
?
i l 3-4p I
3.3 REACTOR COOLANT SYSTEM BASES , 3.3.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES I and 2.with one reactor coolant loop not in operation this specification requires that the plant remain below 65% power. With less than the required reactor coolant loops in operation, the plant shall be in at least HOT STANDBY within 6 hours. The loop isolation valves are required to be OPERABLE in the operating loops in order to terminate the primary to secondary leak path in the event of a steam generator tube rupture. In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident. Single . failure considerations require that three loops be OPERABLE. A single reactor coolant . loop provides sufficient heat removal capability for decay heat if a bank withdrawal accident can be prevented (i.e., by opening the reactor trip system breakers or de-energizing the control rod drive lif t coils).. Single failure considerations require that two loops be OPERABLE. j in MODE 4, two reactor coolant loops provide sufficient heat removal capability for j removing decay heat even in the event of a bank withdrawal accident. Single failure l considerations require that three loops be OPERABLE. A single reactor coolant or l RHR loop provides sufficient heat removal capability for decay heat if a bank j withdrawal accident can be prevented, i.e., by opening the reactor trip system breakers '{ or de-energizing the control rod drive lift coils. Single failure considerations require ! that two loops be OPERABLE. In MODE 5 with reactor coolant loops filled, a single RHR loop provides sufficient heat ) removal capability for removing decay heat. A bank withdrawal accident is prevented by opening the reactor trip system breakers or de-energizing the control rod drive lif t coils. Single failure considerations require that at least two RHR loops be OPERABLE. In MODE 5 with reactdr coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat. A bank withdrawal accident is prevented by opening the reactor trip system breakers or de-energizing the control rod drive lif t coils. Single failure considerations and the unavailability of the steam generators as a heat removing component require that at least two RHR loops be OPERABLE. The operation of one Reactor Coolant Pump (RCP) or one RHR pump provides adequate i flow to ensure mixing, prevent stratification and produce gradual reactivity changes I during boron concentration reductions in the Reactor Coolant System. The reactivity l change rate associated with boron reduction will, therefore, be within the capability of ; operator recognition and control. ] 3-4q I l
g u <
,s. ! . 3.3.1 REACTOR COOLANT SYSTEM h i
BASES ,,_ r 3.3.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION (continued) . The restrictions on starting an RCP with one or more RCS cold legs.less than or equal
.-to 3150F_are provided to prevent RCS pressure transients, caused by energy additions [
from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 j CFR Part,50. The RCS will be protected against overpressure transients and will not ! exceed the limits of ' Appendix G by restricting starting of th_e RCPs to when the l secondary water temperature of each steam generator is less than 200F above each of
~
the RCS cold. leg temperatures. (The requirement to ' maintain the boron concentration of an isolated / idled loop greater j! than or equal to the boron concentration of the operating loops ensures that no . . reactivity addition to the core could occur during startup of an isolated / idled loop. ll Verification of the boron concentration in an isolated / idled loop prior to opening the {
.stop valves provides a reassurance of the adequacy of the boron concentration in the
~ isolated / idled loop. - ;
Start'up of an isolated / idled loop could inject cool water frem the loop into the core. The reactivity transient resulting from this cool water injection is minimized by
- prohibiting isolated / idled loop startup until its temperature is within 200F of the operating loops.' Making the reactor subcritical prior to isolated loop startup prevents any power spike.which could otherwise result from this cool water-induced reactivity transient.
3.3.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized
.above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 293,300 lbs. per hour of saturated steam at 2485 psig. The relief capacity of a single safety )
valve is adequate to relieve any overpressure condition which could occur during . shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures. During operation, all pressurizer Code safety valves must be OPERABLE to prevent the ! L RCS from being pressurized above its Safety Limit of 2735 psig. The combined relief
. capacity of all of these valves is greater than.the maximum surge rate resulting from a D- complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip
[ Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump .
' I valves.
Demonstration of the safety valves' lif t settings will occur only during shutdown and
.will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
3-4r
REACTOR COOLANT SYSTEM - r BASES 3.3.3 PRESSURIZER The limit on the water level in the pressurizer assures that the parameter is mitintained . j within that assumed in the safety analyses. The 12-hour periodic surveillance is 3
- sufficient to ensure that the parameter is restored to within its limit following -
expected transient operation. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation. 4 e A l e 3-4s
- s. >
t REACTOR COOLANT SYSTEM-
~
BASES 1
'3.3.4 RELIEF VALVES Operation of the power-operated relief valves (PORVs) minimizes the undesirable opening of the spring-loaded pressurizer Code safety. valves and provide an alternate
.means of core cooling.' Each PORV has a remotely operated block valve to provide a .
positive shutoff. capability should a PORY become inoperable.' One of two redundant PORY relief trains must be OPERABLE to adequately cool the core in the' event that-the steam generators are not available.to remove core decay heat. When in automstic
; mode, all PORVs and block valves open automatically on high pressurizer pressure. The PORVs and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump..
The.OPER' ABILITY of two spring-loaded relief valves.(SLRVs) or an RCS vent opening
'of greater than 7 square inches ensures that the RCS will be protected from' pressure
.. transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or
. : more of the RCS cold legs _ are less than or equal to 3150F. : Either SLRV has adequate
~
relieving capability to protect the RCS from overpressurization when the transient is limited to either:(1) the start of an idle RCP with the secondary water temperature of .
- the stea.m generator less than or equal to 200F above the RCS cold leg temperatures,'or (2) the start of a char'ging pump (centrifugal) and its injection into a water-solid RCS.
The' Maximum Allowed SLRV Setpoint for the Low Temperature Overpressure Protec .
' tion System (OPS) is derived by analysis which models the performance of the OPS assuming various mass input and heat input tran'sients. O.peration with a SLRV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be -
---.vilo ated with consideration for a maximum pressure overshoot beyond the SLRV
- Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one charging pump (centrifugal or metering) while in MODES 4,5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 200F above RCS cold leg temperature.
4 f ll L 3-4t
1 jj w >
~
-3.5 CHEMICALcAND VOLUMELC0NTROL SYSTEM g
, = Applicability: Applies to the operational status of the chemical t
and volume control system. . Obj ctive: .To specify thoseilimiting conditions for operation of the chemical and: volume control system which must be met in order to ensure safe reactor operation.
-Specification: ~A. The reactor shall not be critical unless the following chemical'and volume control system conditions:are met:
(1) Either two charging pumps or the metering pump and one charging pump operable.
'(2) :Two'borie: acid pumps'or one-boric' acid pump and gravity fillEline to' metering pump operable.
(3) The borie acid tank shall'contain at least 12,000 gallons'of solution whose concentration shall be at least. 8 percent boriccacid, but not greater-than.13 percent boric acid. The temperature shall be 140"F'or higher. (4) : Maintenance, which requires draining of the boric acid mix tank,-shall be allowed only.when the plant is shut down and the-reactor coolant system borated'to the refueling boron concen-tiation. (5) System piping and valves operable to the extent required to establish.two flow paths for' boric acid injection to the reactor coolant system. (6) Valve BA-V-399 shall.be locked'open. and shall not be closed except when the plant is shutdown and the reactor coolant system borated to the refueling boron concentration. B. A maximum of one centrifugal charging pump shall be operable whenever the temperature
'of one or more nonisolated RCS cold legs is less than or equal to 315*F and the RCS is {
not vented by a minimum opening of a 3-inch diameter. 3-8
., i Basis: .The chemical and_ volume control system provides Lcontrol of the reactor coolant systed boron'inventpry.
~
Either a 360 spm charging pump or a 30.gpm metering pump is-capable of injecting concentrated boric acid
'directly into the coolant. system. Approximately-
, '10,000 gallons ~of the 8 percent solution boric acid 4 are required to meet cold shutdown requirements.
Thus, a minimum of 12,000 gallons'in the mix tank is l specified. An upper concentration limit of 13 percent i
.. boric acid'in the mix tank'is specified to maintain i solution solubility at the specified low temperature.
-1 limit:of^140*F. Limits on draining'the tank are j specified to afford opportunity for scheduled. main- "
tenance. The. refueling boron concentration'is
-specified before boric acid mix tank maintenance is-undertaken to preclude a. return to criticality under any circumstances,:even fuel movement. .j When boric acid blender and concentrated boric acid -
o (20,000 ppm boron) are used to change the. react.or coolant' system boron' concentration,.a flow.of 25 spm ,
- can be realized which results'in a change of 600 pps/hr '
in the system. If a blender is bypassed, a flow of k 80 gpm can be realized which will change.the reactor 3 coolant system' concentration at the rate of 2000.. ppm per hour. ' Specification B ensures that'the assumptions'of the.
. Low Temperature RCS Overpressurization Analysis are met by allowing'a' maximum of one charging. pump to be :
operable when. low temperature overpressurization protection is required.-
References:
(1) FDSA Section 4.2 4 Y e 3-9
3.7 MINIMUM WATER VOLUME AND BORON CONCENTRATION IN THE REFUELING WATER STORAGE TANK Applies to the inventory of. borated refueling Applicability: water for core cooling systems, or containment spray. , Objective: To protect against fuel damage and reduce containment pressure by ensuring immediate availability of core cooling and containment spray water. Specification: Whenever the cdre cooling or containment spray systems are specified to be operable, the refueling water tank shall contain not less than 230,000 gallons and.shall have.a boron concen-tratien of not_less.than 2200 ppm. Basis: The volume and boron concentration' requirements of 230,000 GPM and 2200 ppm, respectively, are consistent with the transient and LOCA, analyses. The volume requirements bound the maximum expected boration capability requirement based on end of cycle conditions after xenon decay and cooldown to 140'F. s 3-11
f t ia ov e sl s . . .s tl ce s
. P I e nl il eL s s M ou on P. P. P. n o n nn y n iF diR p2 o oo r oe iFFFi ne r o r i B t
t i% t i%%%i t ovwero/ Crot oL I i t ii tt t e e d0 d04 0d ul at h i i i m b ct n1 n1 71 n nSFW a gt d dd o m A o o o - o n nn e e n C w w - rni o oo gn o C wwwC orrreiw C CC G t p na nl ooo d oeond e we nl l l n S e i C t - oB weeew oBBB o t t t t enn uaa aG ow po n w nn ww r o L ac d . d - d h rwr o oo C ri td t dddt S ed emsd pena et d dd eg ua uaa a u t t t n po h o h oooh tOeeeru o u uu i OL SL SLLLS Hs FGt rh h hh ' l S oS S SS s I dI nn nnnnn ud m C e e - i i : iiiii nonad n n nn g
'n rn a a) at aaaa ianua eeei i ii n m im t t .t t t t t t t a a a a a u uu nn nnn t imS cat in i n P. i i i i i t t t h l ql nt a el n n nn o eo i nedf oi i C
C RC a aF a a a a a aot nf s a a ii MM( MMMMM MCS aAI M M a a MM N o S n r o N oi O Ft I a T d r I D I ee rp N 33 22432 1 4 iO // ///// 43 3 21 1 O n u 21 1 1 1 1 1
/ / // / / / /
C m qr 1 1 22 2 1 1 1
~
u ee G l R w N o o I C cP T i
~
A gl 1 R ol
- $ L u 9 P F O
3 N E O nn L I oo R T ii A A t t T T aa N rr Ef nn ee pp 9 1 p oo ii OO R i t t T r p p pp S a a i ooy N T i r ry r oot r eet T I r Tpppe LL e o .f M iOOf t r 43 a U t c rr a no S oT ppS M I N a e t croor oo et d c p
.- r R a oLL o ia ce o I
et F i F M e nR r ( r R c43( i- T a rl u ee- - l C e e l ps rR a a u p v v n p i s re n h e l g sl i g cL a i Tr s er i r m P evT i t V S T r r re S ar ow PL r e mo n e r l a t o o r st o r o A cL h rt u ia i u t a gec s M r t s c ee e a s a l e Rl it a Ha e s e wn l e ep v b WR r oe o r R o e
'ra d l G s P t L s ei P) p F I ) RS wr eh w m h p U t
n xgo wi ra w t n oa iil or ee o gi Sd n o e pV FHF LT t t i r t o e r m r l HT n eR m t e er rrt rn pw aS F e g e u r ve een eo id w m t n no m e nR r e i u Ozi zza zi reo a r aU i N iil i t TeL e ei RS u q rr rro rcd F mt i q h e a u uuo ueel - h t nc u ee e g R es s sC sjt a mt S i e q t g R i l s s s s neuai aj e an H r ce eew h t n R i a g e ur r r o eI leat r new g nI dR n n w NP PPL P i o p e o DMS H C u me i l w o P t r c e d
. .. r a
er u t t 1 2 345 6 7 89 0 t u f u A I t no e h 1 I S I S R S
o,
,7-f
- g '
~
- 3. l 0 - REACTivlTY CONTROL' SYSTEMS 3.10.1L BORATION CONTROL'
' SHUTDOWN'M ARGIN ?
FOUR LOOPS OPERATING. ..
. LIMITING bONDITION FOR' OPERATION 3.10.1.1- iThe SHUTDOWN MARGIN shall be greater' than or equal to 1800 pcm. 1 l
* ' APPLICABILITY: MODES I and 2* (Four Loop Operation). .
A CTION: .
." With.the SHUTDOWN MARGIN less than 1800 pcm,immediately initiate and continue:
boration at greater than or equal to 30 gpm'of a solution containing greater than or equal to 14,000 ppm boron or its equivalent until the required SHUTDOWN MARGIN is - restored..
. SURVEILLANCE REQUIREMENTS
' l.: The SHUTDOWN MARGIN shall bed' etermined to be greater than or
-. equal to 1800 pcm:
- a. Within I hour af ter detection of a6 inoperable control rod (s) and at least
~
once per 12 hours thereaf ter while the rod (s)1s inoperable. If the
. inoperable control rod is linmovable 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);
b.. When in MODE 1 or MODE 2 with Keff greater than or equal to I at least once per 12 hours by verifying that control bank withdrawals within the limits of Specification 3.10.2.6;
- c. - When in MOD'E 2 with K eff essl than 1, within 4 hours prior to achieving reactor . criticality by verifying that the predicted critical control rod.
position is within the limits of Specification 3.10.2.6;.
- d. Prior to initial operation above 5% RATED THERMAL POWER af ter )
each fuelloading, by consideration of the factors in 2 below, with the' control banks at the maximum insertion limit of Specification 3.10.2.6 '
.and
*See Special Test Exceptions Specification 3.24.1.
3-16
i
- REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) '
i
- 2. During MODE 1 operation, the overail core reactivity balance shall be ;
compared to predicted values to demonstrate agreement within 2 1000 i pcm at least once per 31 Effective Full Power Days (EFPD). This ' comparison shall consider at least those factors stated below. The predicted reactivity values shall be adjusted (normalized), if required, to , L I correspond to the actual core conditions prior to exceeding a fuel burnup ! of 60 EFPD af ter each fuelloading. a) Reactor Coolant System boron concentration, b) Control rod' position, i c) Reactor Coolant System average temperature, d) Fuel burnup based on gro's s thermal energy generation, i e) Xenon concentration, and f) Samarium concentration.
- s e
e 2 2 l I i e
+
. )
l 3-17
D REACTIVITY CONTROL SYSTEMS . b
.. SHUTDOWN MARCIN, 1 J
' LIMITING CONDITION FOR OPERATION !
l
. . -I 3.10.1.23 The SHUTDOWN MARGIN shall be greater than or equal to: {
1
- a. 1800 pcm with no reactor coolant loop idled or isolated,* or j
- \
- b. 2600 pcm with one or more reactor coolant loops idled or isolated, {
except during the startup of a reactor coolant pump. APPLICABILITY: MODE 3. J ACTION: , With the SHUTDOWN MARGIN less that, the required value, immediately initiate and + continue boration at greater than or equal to 30 gpm of a solution containing greater . than or equal to 14,000 ppm boron or its equivalent until the required SHUTDOWN i MARGIN is restored. , SURVEILLANCE REQUIREMENTS The SHUTDOWN MARG 11 shall be determined to be greater than or equal to the required vah e:
- a. Within I hour af ter detec tion of an inoperable control rod (s) and at least once per 12 hours therea ter while the rod (s) is inoperable. If the inoperable control rod is isnmovable or untrippable, the SHUTDOWN MARGIN shall be verifie3 acceptable with an increased allowance for the withdrawn worth of !.he immovable or untrippable control rod (s); and l
- b. At least once per 24 hours by consideration of the following factors:
- 1) Reactor Coolant System boron concentration,
- 2) Control rod position,
- 3) Reactor Coolant System average temperature,
- 4) Fuel burnup based on gross thermal energy generation,
- 5) Xenon concentration, and j
. l '
- 6) Samarium concentration.
- Idled or isolated refers to the position of the loop stop valves, not the reactor coolant 1 pump operating status.
l
)
3-17a 1 4
REACTIVITY CONTROL SYSTEMS i SHUTDOWN MARGIN - l L . LIMITING CONDITION FOR OPERATION 3.10.1.3 The SHUTDOWN MARGIN shall be greater than or equal to 2600 pcm. APPLICABILITY: MODE 4 and 5. ACTION: e With the SHUTDOWN MARGIN less than 2600 pcm,immediately initiate 'and continue boration at greater than or equal to 30 gpm of a solution containing greater than or
; equal to 14,000 ppm boron or its equivalent until the required SHUTDOWN MARGIN is restored.
SURVEILLANCE REQUIREMENTS The SHUTDOWN MARGIN shall be determined to be greater than or s . equal to 2600 pcm:
- a. WIthin I hour af ter detection of an inoperable control rod (s) and at least once per 12 hours thereaf ter while the rod (s)is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN '
MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and
- b. At least once per 24 hours by consideration of the following factors:
- 1) Reactor Coolant System boron concentration, l, 2) Control rod position, -
- 3) Reactor Coolant System average temperature,
- 4) Fuel burnup based on gross thermal energy generation,
- 5) Xenon, concentration, and
- 6) Samarium concentration.
l
.' 3-17b
s l
} 1 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN t l
THREE LOOPS OPERATING l E LIMITING CONDITION FOR OPERATION . 3.10.1.4 The SHUTDOWN MARGIN shall be greater than or equal to 2600 pcm. APPLICABILITY: MODES I and 2* (Three Loop Operation). 1 ACTIONS . With the SHUTDOWN MARGIN less than 2600 pcm,immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 14,000 ppm boron or its equivslent until the re, quired SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS
- 1. The SHUTDOWN MARGIN shall be determined to be greater than or equal to 2600'pcm:
- a. Within I hour af ter detection of an inoperable control rod (s) and at least once per 12 hours thereaf ter 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 incre'ased allowance for the withdrawn worth of the immovable or untrippable control rod (s);
- b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1 at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.10.2.7;
- c. When in MODE 2 with Kefg less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.10.2.7;
- d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the f actors in 2 below, with the control banks at the maximum insertion limit of Specification 3.10.2.7; and i
'See Special Test Exceptions Specification 3.24.1.
3-17c l
7 . j i E REACTIVITY CONTROL S'YSTEMS ll SURVEILLANCE REQUIREMENTS (Continued)
.2. During MODE 1 operation, the overall core reactivity balance shall be
- compared to predicted values to demonstrate agreement within +1000 pcm at least once per 31 Effective Full Power Days (EFPD). ThTs comparison shall consider at least those factors stated below. The
~,
predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60
~
-EFPD af ter each fuel loading.
a) Reactor Coolant System boron concentration, b) Control rod position, .j c) Reactor Coolant System' average temperature,
' d) Fuel burnup based on. gross thermal energy generation, e) Xenon concentration, and f) ~ Samarium concentration.
9 0 3-17d
g - - - - - - - s.
. i REACTIVITY CONTROL SYSTEMS-
' MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION F 3.10.1.5, The moderator temperature coefficient (MTC) shall be:
a .- Less positive than .5 pcm/0F for the all rods withdrawn, beginning of -
, . . cycle life (BOL), hot.zero THERMAL POWER condition;'and b.' Less positive than 0 pcm/0F for the allrods withdrawn, BOL, RATED THERM AL POWER condition; and
- c. - Less negative than -29 pcm/0F for the all rods withdrawn, end of cycle-g-
life (EOL), RATED THERMAL POWER condition.. , APPLICABILITY:- ' MODES I and 2* * *. ACTION:
- a. .With the MTC more positive than the !!mits of Specifications 3.10.1.5a '
. or 3.10.1.5b above, operation in MODES 1 and 2 may proceed provided:;
1.- Control rod withdrawal limits are established and maintained - t sufficient to restore the MTC to less positive than 5 pcm/0F at hoti zero THERMAL POWER or 0.pcm/0F at RATED THERMAL POWER within 24 hours or be in HOT STANDBY within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits of Specification 3.10.2.6 or 3.10.2.7;-
- 2. The control rods are maintained within the withdrawallimits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and
- 3. A Special Report is prepared and sub'mitted to the Commission pursuant to Specification 6.9.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the MTC to within its limit for the all rods withdrawn condition.
- b. With the MTC more negative than the limit of Specification 3.10.1.5c above, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
E *With Kegg greater than or equal to 1.
* *See Special Test Exceptions Specification 3.24.2.
'3-17e
=- _ _ - _ _ _ _ _ _
' REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS The MTC shall be determined to be within its limits during each fuel cycle as follows: l
- a. The MTC shall be measured and compared to the BOL limit of 'i Specification 3.10.1.5a. above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuelloading;
- b. The measured MTC shall be adjusted to the BOL, RATED THERMAL POWER conditions and compared to the BOL limit of . Specification 3.10.1.5b above, prior to initial operation above 5% of RATED THERMAL POWER, af ter each fuel loading; and
. c. The' measured MTC shall be adjusted to EOL, RATED THERMAL POWER conditions and compared to the EOL limit of Specification 3.10.1.5c ;
above, prior to initial power operation above 5% of RATED THERMAL - POWER, af ter each fuel loading, l i I 4 { y a
't i
e t > 1 3-17E 1
-J
,V , ..
m
, .#, 0:
4
- n. .
t REACTIVITY CONTROL SYSTEMS , 4 MINIMUM TEMPERATURE FOR CRITICALITY
- LIMITING CONDITION FOR OPERATION '
3.10.1.6 The Reactor Coolant System operating temperature (Tavg) shall be' greater than or equal to $250F. ; f" 9: - APPLICABILITY: l MODES 1 and 2* **. ACTION: . Withto a Tavg Re$ctor within Coolant its limit within System 15 minutes operating or be,in HOT STA BY temperature within the next 15 '(Tavh)less minutes.- , s
. SURVEILLANCE REQUIREMENTS
.The Reactor Coolant System temperature (Tavg) shall be determined to . '
be greater than or equal to 5250F: t a.L .Within 15 mint.t.es prior to achieving reactor criticality, and
, b. At least once per 30 minutes when the reactor is critical and the - -
Reactor Coolant System Tavg is less than 5300F with the
, - Tavg -Tref Deviation Alarm not reset.
- With Keff greater than or equal _ to 1., ,
;l
.;... **See Special Test Exceptions Specification 3.24.2.-
.; ' l- ;if > l u
?
; g+ ,
, /'I ;
- j
/
/ lY l
3-17g m
y REACTI'VITY CONTROL SYSTEMS 3'.10'.2' MOVABLE CONTROL ASSEMBLIES BANK HEIGHT-LIMITING CONDITION FOR OPERATION t: 3.10.2.1. All shutdown and control rods shall be OPERABLE and positioned within i 24 steps indicated position (RPI) of their bank position, as indicated by the average of the RPI for the respective bank. APPLICABILITY:' MODES 1* and 2*. ACTION:
- a. With one or more rods inoperable due to being immovable as a result of ~
excessive friction or mechanicalinterference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.10.1.1 or Specification 3.10.1.4 is satisfied within I hour and be in HOT STANDBY within 6' hours.
- b. With one rod trippable but inoperable due to causes other than addressed by !
ACTION a., above, or misaligned from its bank position by more than124 steps, !
.;,. . POWER OPERATION may continue provided that within I hour:
- 1. The rod is restored to OPERABLE status within the above alignment requirements, or:
- 2. The rod is declared inoperable and the remainder of the rods in the bank with the Inoperable rod are aligned to within124 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figure ,
3.10-l'or Figure 3.10-2. The THERMAL POWER Level shall be restricted pursuant to Specification 3.10.2.6 during subsequent four loop j operation or Specification 3.10.2.7 during subsequent three loop operation, or
- 3. The rod is declared inoperable and the SHUTDOWN MARGIN l requirement of Specification 3.10.1.1 for four loop operation or Specification 3.10.1.4 for three loop operation is satisfied. POWER 3 OPERATION may then continue provided that: l a) The THERMAL POWER levelis reduced to less than or equal to 55% of RATED THERMAL POWER within the next hour and within the following 4 hours the Nuclear Overpower Trip Setpoint is reduced to less than'or equal to 74% of RATED THERMAL POWER.
b) The SHUTDOWN MARGIN requirement of Specification 3.10.1.1 for four loop operation or Specification 3.10.1.4 for three loop operation is determined at least once per 12 hours;
*See.Special Test Exceptions Specifications 3.24.2.
3-17h !
<l
_ _ _ _ _ _ ______a-
, l i
[ REACTIVITY CONTROL SYSTEMS, LIMITING' CONDITIONS FOR OPERATION ACTION (Continued) c) A power distriburbn map is obtained from the movable incore detectors and LHGR within 72 hours; and and FhH are verified to be within their limits
.d) A reevaluation of each accident analysis of Table 3.10-1 is performed within 10 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these' conditions; 4
- c. With more than one rod misaligned from its bank position by more than g24.
steps (indicated position), be in HOT STANDBY within 6 hours. . SURVEILLANCE REQUIREMENTS a.. The position of each rod shall be determined to be within the position limit by verifying the individual rod positions at least once per 12 hours : excep.t during time intervals when the rod position deviation monitor (Datalogger) is inoperable, then verify the individual rod positions at least once per 8 hours. '
- b. - Each rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at .i least once' per 31 days. '
i' ( i I V i; P h 3-171 I
hbl7 .
- TABLE 3.10-1.
ACCIDENT ANALYSIS REQUIRING REEVALUATION
- IN THE EVENT OF AN INOPERABLE CONTROL ROD;
' Rod Cluster Control Assembly insertion Characteristics '
, Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant frem Small Ruptured Pipes or from Cracks in.
Large Pipes Which Actuates the Emergency Core Cooling System Major. Reactor Coolant System Pipe Ruptures (Loss of Coolant' Accident)- i
' Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Closter Control Assembly Ejection) '
l O I; 3-17j !
l' j REACTIVITY CONTROL SYSTEMS l t POSITION INDICATION SYSTEMS-OPERATING LIMITING CONDITION FOR OPERATION ' 3.10.2.2 The Digital Rod Position Indication System (utep counters) and the 1 Analog Rod Position Indication System (RPI) shall be OPERABLE and capable of determining the control rod positions within 116 steps. APPLICABILITY: MODES I and 2. ACTION:
- a. With a maximum of one analog rod position indicator per bank inoperable, restrict the movement of the rod bank which includes the nonindicating rod to i 8 steps from the position last deterrnined prior to loss of the nonindicating rod.
If the position of the nonindicating rod is not determined within 8 hours, declare the rod inoperable and refer to Specification 3.10.2.1.
- b. With a maximum of one digital rod position indicator per bank inoperable either:
- 1. Verify that all analog rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 32 steps of each other at least once per 8 hours, or
- 2. With four loops operating, reduce THERMAL POWER to less than 65% of RATED THERMAL POWER within 8 hours.
- 3. With three loops operating, reduce THERMAL POWER to less than 16%
of RATED THERMAL POWER within 8 hours.
- c. With more than one analog rod position indicator or digital rod position indicator per bank inoperable, be in at least HOT STANDBY within 6 hours with the reactor trip breakers open or the control rod drive lif t coils deenergized.
SURVEILLANCE REQUIREMENTS Each digital and analog rod position indicator shall be determined to be OPERABLE by verifying that the bank average analog rod position and the digital rod position agree within 16 steps at least once per 12 hours except during time intervals when the rod position deviation monitor (Datalogger) is inoperable, then compare the analog rod position and the digital rod position at least once per 8 hours. 3-17k
.,:' yf 4 (I> \ / %
REACTIVITY CONTROL' SYSTEMS
' PbSIT!ON INDICATIO'N SYSTEM-SHUTDOWN-LIMITING CONDITION FOR OPERATION 3.10.2.3 The analog rod position indicator shall be OPERABLE and capable of determining the control rod position within 216 steps for each shutdown -
or. control rod not fully inserted. APPLICABILITY: MODES 3 * * *, 4 * * *, and 5 * * *.
' ACTION:
. With less than the above required rod position indicator (s) OPERABLE,
.immediately open the Reactor Trip System breakers.
~ ~
SURVEILLANCE REQUIREMENTS Each of the above required analog rod position indicator (s) shall be determined to be OPERABLE at least once per refueling by verifying that the digital rod position indicators agree with the analog rod position , indicators within 16 steps when exercised over the full-range of rod i travel. ' 4 i
. *With the Reactor Trip System breakers in the closed position.
L **See Special Test Exemptions Specification 3.24.3. f, i . y I I 1 l i 3-17L
- -- )
7, . -
' REACTIVITY CONTROL SYSTEMS
' ROD DROP TIME 4
' LIMITING CONDITION FOR OPERATION 3.10.2.4. The measured individual rod drop time from fully withdrawn position-shall be less than 2.5 seconds from the fully withdrawn position to the bottom of the dashpot with Tavg greater than 5250F and four reactor coolant pumps operating. A drop time of 2.45 sec. shall be used if only 3 reactor coolant pumps are operating while the drop tests are made.
< APPLICABILITY: MODES I and 2.,
ACTION: With the drop time of any rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2. SURVEILLANCE REQUIREMENTS The rod drop time shall be demonstrated through measurement prior to s reactor criticality:
- a. For all rods following each removal of the reacter vessel head,
- b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific ro'b, and
- c. At least once per 18 months.
I o 3-17m
, _________________________________________j
~
',.C -
'hy.It h
> REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT '
LIMITING CONDITION NOR' OPERATION - 3.10.2.5' All shutdown rods shall be withdraw'n to equal to or greater than 320 steps.-
- APPLICABILITY: MODES l' and ,2 + * *. . -
J ACTION:- , With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing,within I hour either: ~
- a. - Fully withdraw the' shutdown rod, or b.- Declare the shutdown rod to be Inoperable and apply Specification 3.10.2.1.
- SURVEILLANCE REQUIREMENTS Each shutdown rod shall be determined to be fully withdrawn:-
- a. Within 15 minutes prior to withdrawal of any rods in Control Bank
' A or B during an approa0 to reactor criticality, and
- b. - At least once per 12 hours thereafter.
3
*See Special-Test Exceptions Specification 3.24.2.
**With Keff greater than or equal to 1.
9 4 O 3-17n
7 . .. , , , . 7 ml. w R , p _
, . REACTIVITY CONTROL SYSTEMS 1 CONTROL GROUP INSERTION LIMITS l
.' FOUR LOOPS OPERATING L >
1 ,, i LIMITING' CONDITION FOR OPERATION- !
. 3.10.21 The control b'anks shall be ilmited in physical insertion as shown in 1
. Figure 3.10-1. -{
> APPLICABILITY: MODES 1*'and 2+ ** (Four Loop Operation).:
ACTION:
'.. -With the control banks inserted beyond the above inser'tlon limits, except for surveillance testing:L
._ a. . Immediately initiate action to restore the control banks to within the limits within 2 hours, or' . .
' b. Elmmed!stely initiate action to reduce THERMAL POWER.within 2
' hours to less than or equal to that fraction of RATED THERMAL.
POWER which is allowed by the bank position using the above
' figure, or c.' , Be in at least HOT STANDBY within 6 hours.
S' SURVEILLANCE REQUIREMENTS
' The po'sition of each control bank shall be determined to be within.the :.
Insertion limits at least once per 12 hours by verifying the individual rod - positions.
. *See.Special Test Exceptions Specification 3.24.2.
* *With'Keff greater than or equal to 1. -
A 1 9 1 3-170 1 1'_u__1_1_._____ ______.________.2_._.__.______
,; 1Ii}i
=,
m f l , _ 0
- . 2 3
_ 1I 0 i ! 9 2 _ ' i
- . n 0 w Il ,
5 a 2 r
- . d
_ - h R
/
t E
) i f
5 2 0 W l 0 - H 1 s P _ I 2 p
- . e L 0 t 4 ' ,
S A I ( s
- . B B I
E -
# r .
0 p 7 u l i j - . 1 o T r C S
)
[- i U S RN 4 3 7 / Ii 0 1 3 1 EO VI 1
, 0 T e 1 SAl Tl 5
1 (
/r '
0 0 3 I E MP
, 2 9 E
/ 3 I. O
- . R 1 .
U P f ' G I NO OO F IL ._ 0 0 T- - AI
, 8 n5 2 w RR a El l r SO d NI h I 0i t 0
/ -
, 4W 2
1 P l l
- . sO 0 p R t ,
e G t
/ - .
, 0A 0
S L 2 O - p R
^
u T
/ .
o r G N O C 0 ._
, 6 .
1 A i I-
-N 1
kII
- e- '_ " i_ i .
- - Z 2 -
0 0 0 0 0 0 0 0 0 0 O , 0 0 0 0 0 0 0 0 0 0 . 0 8 6 4 2 0 8 6 4 2 2 j* 1 1 1 1 m fa% s g .
,n ,
j a' , q ..
.x
~ REACTIVITY CONTROL SYSTEMS'- .
+
i CO'NTRO!. GROUP INSERTION LIMITS - sTHREE LOOPS OPERATING i. LIMITING CONDITION FOR OPERATION 3.10.2.7 ' The control banks shall be limited in physical insertion as shown in Figure 3.10-2.-
- APPLICABILITY: MODES 1* and 2+ ** (Three Loop Operation).
- e. ACTION: i I
With the control banks inserted beyond the above insertion limits, except for surveillance testing: .
-l 1
- a. Immediately initiate action to restore the control banks to'within the limits within 2 hours, or
- b. Immediately initiate action to reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above fi,gure, or
- c. Be in at least HOT STANDBY within 5 hours.
SURVEILLANCE REQUIREMENTS ' The position of each control bank shall be determined to be within the Insertion !!mits at least once per 12 hours by verifying the individual rod positions.
*See Special Test Exceptions Specification 3.24.2.
*
- With Keff greater than or equal to 1.
i i 3-17p i
I)Ili! - I1(f1 !jiIlI ; l n g .' 0
~ 2
! I gI I 3 6
lI 1, 0 9
' 2 l gI I g
I 0 n _ 5 w 2 a
.- r iI d R h E t
.- i W i 0W O gI 1 P 2 s
. - p I e A t
)
8 1 6 / . -
,0 S
B R E l M
/
1 i 1 7 p 1 u T 0 1 7> . - o r S 1 C U _ ( / . - SN
/ 0 3 RO EI 1
VT
'f .
n 2
- SR A
0 TE
. - 1 I P
- 0 0 .
MO
/ -
- 2 3
9 3 I I .P
/ E R NO O
J U OL 1l f !tl i 0 8 n5 2 w 0 G I F I - TE RE
/, . -
a r Eitl Si d h NT I t
/ 0i0 P n
4W1
/ -
2 s0 p U O R u e G - t
.- S
< 0 L itl ,
2 0A O
.- p R -
u T l , o N r O 0 6 G C - 1
- I I yTl- -
"_ T,
_ 0 4 1 0 0 0 2 1 0 1 0 0 0 0 8-O o s 0 0 4 0 0 2 0
- mh <3 gg 1. g w -
l \l L
b, I k= , . o . ' 1 3.10 REACTIVITY CONTROL SYSTEMS-
;c
- B ASES " .l 3.10.1 BORATION CONTROL
! 3.10.1.'I . "3.10.1.2. and '3.10.1.3 SH UTDOWN MA RGIN
.r . A sufficient SHUTDOWN MARGIN ensures that:(1) the reactor can be made subcritical
. from all operating conditions, (2) the reactivity transients associated with postulated .
. accident conditions are controllable within acceptable limits, and (3) the reactor will be I l
-maintained sufficiently subcritical to' preclude inadvertent criticality in the shutdown:
condition. SHUTDOWN MARGIN. requirements vary throughout core life as a function'of fuel depletion, RCS boron concentration, and RCS Ta
'during MODES 1,'2 and 3 occurs at end-of-cycle [i$e. The most restrictive condition (E postulated steam line break accident and resulting RCS cooldown.. In the accident fl -
' analysis,' a minimum SHUTDOWN MARGIN of 1800 pcm for four loop operation and 2600 pcm for three loo'p operation is assumed.. Operation in MODE 3 with two operatipg '
reactor coolant pumps is bounded by the four loop steam line break analysis. Operation , in MODE'3 with one operating reactor. coolant pump and two OPERABLE reactor -i coolant loops (both loop stop valves open'in each loop) is bounded by the three loop steam line break analysis.' Because of the short time involved, the 2600 pcm SHUTDOWN MARGIN limit need not be applied to the closure of the cold leg loop stop ;
, valve.in order to restart the reactor coolant pumps from an initial four loop operation
' condition. The most restrictive condition in MODES 4 and 5 is associated with the
' boron dilution accident. :In the analysis of this accident, a minimum SHUTDOWN MARGIN of 2600 pcm in . MODES 4 and 5 is required to control the reactivity transient.
Accordingly, the SHUTDOWN MARGIN requirements are based upon this limiting ' condition and are consistent with current design basis assumptions.
' 3.10.1.4 MODERATOR TEMPERATURE COEFFICIENT 3
The !!mits on the moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within'the limiting condition assumed in the
; accident and transient analysis.
L The MTC values of this specification are associated with a specific set of plant conditions; measurement of MTC values at conditions other than those explicitly stated
. with extrapolation to the specified conditions is acceptable. Correction factors shall account for fuel and moderator temperature and boron concentration. -
3-17q I I f
. 4 REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)-
The Surveillance Requirement for measurement of the MTC at the beginning of the fuel cycle is adequate to, confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3.10.1.5 MINIMUM TEMPER ATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 5250F. This limitation is required to ensurei(1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the ' reactor vesselis above its minimum RTNDT temperature. e e 3-17r l
9 .. ['1 n
~
n " REACTIVITY CONTROL SYSTEMS-BASES u > t
,\
' E 3.10.2 'MO'VABLE' CONTROL ASSEMBLIES '
; The' specifications of this section ensure thab (1) acceptable power distribution limits j are maintained,(2) the minimum SHUTDOWN MARCIN is maintained,(3) the potential '
effects of rod misalignment.on associated accident'ana yses are limited. -
- OPERABILITY of the control rod position indicators is required to determine control .
' rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees .with the bank average Analog Rod Position' indicator within 3 16 steps provides assurance that.
the' Digital. Rod Position Indicator and the Analog Rod Position Indicator is operating - correctly over the full range of indication. The' ACTION' statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Continued operation with an inoperable rod requires measurement of peaking' . factors and a' restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. - In addition, those safety analyses affected by an inoperable rod are reevaluated to confirm that the results remain valid - Jduring future operation.- The maximum rod drop time restriction-is consistent with the assumed rod drop time used in the' safety analyses. Measurement with Tavaequal to or greater than 5250F and
. with four or three reactor coolant pumps operating Ensures that the measured drop d times will be representative of insertion times experienced during a Reactor trip'at i
. operating conditions.
l Control rod positions and OPERABILITY of the rod position indicators are required to j be verified on a basis of once per 12 hours with more frequent verifications required if '{
.an automatic monitoring channelis inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.
, 1 i
i I l 3-17s , l
3.11 CONTAINMENT Applicability: ' Applies to the operating status of reactor containment. Objective: To insure containment integrity. Specification: A. Leakage 1 The reactor shall not be critical if the containment leakage exceeds 0.25 weight percent of the contained air per.24 hours
.when extrapolated to 40'psig in accordance
, with Surveillance Standard 4.4. f i
B. Containment Integrity (1) Containment integrity shall be maintained whenever the reactor coolant system is above 300 psig and 200*F. The shutdown margin shall be greater than 2600pcm when the containment is open. (1) Containment integrity shall not be l violated when the reactor vessel head is removed unless the reactor coolant system is borated to the refueling ; boron concentration. (3) Positive reactivity changes shall not be made by rod drive motion or boron ' dilution whenever the containment integrity is not intact. C. Internal Pressure The reactor shall not be critical if the containment internal pressure exceeds 3 psig, or the internal vacuum exceeds 2.0 psig. D. Air Recirculation System Three of the four air recirculation units shall be operable whenever the reactor is critical. E. Containment Spray System The containment spray system shall be operable whenever the reactor is critical. F. Containment Venting (1) Either the containment air. particulate monitoring system or the containment purge exhaust system shall be available at all times when the reactor is critical for post accident hydrogen venting. 3-l@ _
.%, 4 (2) Containment purge capability may be rendered inoperable when the reactor is critical by placing a blank _ flange on the 42-inch purge' air exhaust penetration inside the reactor contain-ment for: a period of seven days. If the blank flange' cannot be remoyed within seven days, then the reactor
, shall be shut down within 24 hours.
G. The containment isolation valves specified in Table 3.11-1-shall be operable while in Modes 1, 2, 3, and 4, or: With one or more of the isolation valve (s) specified in Table 3.11-1. inoperable, maintain at least one isolation valve OPERABLI in each affected penetration that is open and either; (1) Restore the inoperable valve (s),to OPERABLE status within four hours, or (2) Isolate each affected penetration within four hours by use of at least one deactivated automatic valve secured in the isolation position,.or (3) Isolate each affected penetration within four hours by use of at least one closed manual valve or blind flange, or (4) Be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN i within the following 30 hours. H. Containment Isolation The Containment Isolation actuation system shall be operable with the following trip
, setpoints:
Containment Pressure - HI < 5 psig Pressurizer Pressure - LOW > 1700 psig Basis: A containment leakage rate of 0.3 weight percent of the contained . air per 24 hours at an internal pressure of 40 psig under hypothetical accident conditions with 3 of 4 air recirculation units operatius will maintain public exposure well below 10 CFR 100 values (see Section 10.4 of the i FDSA). The reactor coolant system conditions of 300 psig and 200*F assure that no steam will be formed ~ and hence there'would be no pressure buildup in the containment if a reactor coolant system rupture were to occur. 3-19
-3.13 REFUELING Applicability: Applies to operating. limitations during refueling operations.
Objective:- To insure that no incident could occur during refueling operations that would affect public health and safety.
~
Specification: A. Radiation levels in the Containment ahd Fuel Storage Building shall be monitored continuously. B. Core suberitical neutron flux shall be continuously monitored by st least two neutron monitors, each with continuous visual and audible indication available, whenever core geometry is:being changed. When core geometry is not being changed, at least one neutron flux monitor shall;be in service. C. (1) Whenever the water level in the refueling cavity is less than 23 feet above the flange of the. reactor pressure vessel, two residual heat removal loops shall-be OPERABLE. With Jess than ! the required RHR loops OPERABLE, inrnediately initiate corrective action to return.the required RHR loops to OPERABLE status as soon as possible.
. Also, with less than the required depth of water described above, suspend all operations involving movement of fuel assemblies or. control rods within the reactor pressur,e vessel.
(2) At least 'neo RHR pump and heat exchanger shall be-in operation except that the residual heat removal system may be removed from operation for up to I hour per 8 hour period during performance of core alternations in the vicinity of the. reactor pressure-vessel hot legs. With less than one residual heat removal pump and heat exchanger in operation except as described above, suspend all operation involving any increase in reactor decay i heat load or a reduction in boron concentration
~
of the reactor coolant system. Close all contain-ment. penetrations providing direct access from the containment atmosphere to the outside atmosphere within four hours. D. During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concentration of all filled, unisolated portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure a keff less than or equal to 0.94. i 3-23 . _ _ _ _ _ _ _ _ _ _ _ _ _ . - - - - - -I
w ~ m *
)
f . . . . . E. One charging pump capable of-injecting borated. water 6 < to the reactor ecolant shall be,available at all 1 n . time.when changes in core geomet.ry are taking place.
* . . .l F.~
~Whenever new fuel.is added to the. reactor core, a j 1/M plot shall be maintained to verify the s
suberiticality of the core.
~
G. Direct communication between the Control Room and the
. refueling-cavity manipulator crane shall be available whenever changes in core geometry are taking place.
1 H. Spent fuel casks shall not be handled above the spent . f fuel pool or its edge except as provided-in Section 3.13.I, until such time as NRC has received and approved the
' spent fuel cask drop evaluation. .j
< q I. After April 23, 1980,'.a spect fuel cask may be brought u into the Spent Fuel Building and may be moved into or over the spent fuel pool a total of ten times in order to remove fuel-from the pool for study at on off-site laboratory or to return the fuel from the laboratory to the pool. Movement of.the spent fuel cask under the j
-provisions of this paragraph is conditioned on compliance !
E (by the licensee) with all commitments made by the . ' Llicensee in its letters to the NRC' dated April 18, 1980, and April 23, 1980.~ In. addition, all fuel within the spent fuel pool shall have decayed for.at least 90 days before a spent' fuel cask is handled above.the pool.. Basis: The equipment and general procedures to be utilized during refueling are discussed in the facility Description and 1 Safety Analysis. Detailed instructions will be available for use by refueling personnel'. . These instructions, the I above-specified precautions, and the design of the fuel { handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could { occur during the refueling operationr that would result.in a hazard to public health and, safety. Whenever no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous monitoring ~of radiation levels (A above) and
' neutron flux provides immediate. indication of an unsafe c>ondition. The residual heat removal pump is used to maintain a uniform boron concentration. The refueling boron concentration indicated in Part D will keep the core j' suberitical even if all control rods were withdrawn from the core. For a core configuration of all rods in, an additional 0.05 kg f penalty is required to account for a heavy load
, crushing the core into a more reactive configuration.
Weekly checks of refueling i w 3-24
l _ , (' Intentionally Lef t Blank i 3-27
,' , l,'-
{:- l 1 '4
. Intentionally Lef t Blank I
3-28
7-
+;
x _y
- , , -e y *
~
? 3.17 POWER DISTRIBUTION LIMI'TS 3.17.l?AXIAllOFFSET p FOUR LOOPS OPERATING- 'll
. .- LIMITING CONDITION FOR OPERATION' 3.17.1.1 The AXIAL OFFSET shall be maintained within the limits of Figures; 3.17-la or b. o MODE 1, above 4096 of RATED THERMAL POWER'. APPLICABILITY:. ,
' ACTION:-
, , With the AX1AL OFFSET outside the Acceptable Operation Limits specified in the
-t above figures, within 15 minutes initiate corrective action and continue the corrective action so that the Axial OFFSET is within limits within 2 hours or reduce THERMAL
' POWER to less than 40% of RATED THERMAL POWER within the next 4 hours.
- SURVEILLANCE REQUIREMENTS
- a. The AX1AL OFFSET shall be determined to be within the Acceptable ,
Operation Limits of Figures 3.F1 a or b by monitoring the AXIAL; OFFSET.using at least two OPERABLE excore Power Range. channels _ , and applying,the excore/incore correlation on a continuous basis.- b.. The excore/incore correlation shall be verified at least once per 31
' EFPD and adjusted at least once per 92 EFPD using the results of the measurements obtained in accordance with Specification'3.17.2.~ '
1
- c. The excore/incore correlation shall.be determined after each fuel loading or ma[or change in excore Po_wer Range instrumentation prior to exceeding 80% of RATED THERMAL POWER.'
d.: The excore Power Range detectors shall be calibrated / correlated relative to the Movable Incore Detector System measurements within 7 days after completion of incore measurements. 4 d 3-30
- a ,, {
. )
? I
~(-15.100) --
(10,100) 100 90 . 80 (-30,70) - - (20,70) 7aU . t 60 n"
" 50 E
(-30,40) ,, (20,40) 40 )
. . 1 1
30'
);
20 10 a . . . . . , , . .
-$0 -40 -30 -20 -10 0 i0 2'0 3'8 4'0 5'0
% OFFSET l
FIGURE 3.17-1A l POWER vs OFFSET, 0-250 EFPD, FOUR LOOP l l l - i l
- . - - - . -------_____________________________________________________J
=- q l
. t
(-20,100) (10,100) aos , Avu . l A h, i' 30 43 l
~
(-30,70) (20,70)
- v. -
.0
<n M DU z
o . . c CD
" v4
(-30,40) . . (20,40)
.a o ad T.~'h
.~ra L L' en I 10
~
0~h~00-0 0 10 20 30 (0 5'0
% DFFSET 1
FIGURE 3.17-1B POWER vs OFFSET, 250 EFPD-EOC, FOUR LOOP i
s
- t. ,
o l THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.17.1.2 The AX1AL OFFSET shall be maintained within the limits of Figures. 3.17-2a or b. APPLICABILITY:' MODE 1, above 40% of RATED THERMAL POWER. ACTION: With' the AX1AL OFFSET outside the Acceptable Operation Limits specified in the . above figures, within 15 minutes initiate corrective action and continue the corrective action so that the Axial OFFSET is within limits within 2 hours or reduce THERMAL POWER to less than 4096 of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS
.a. The AXIAL OFFSET shall be determined to be within the' Acceptable
-Operation Limits of Figures 3.17-2a or b by monitoring the AXIAL OFFSET using at least two OPERABLE excore Power Range channels and applying the excore/incore correlation on a continuous basis.
- b. .The excore/incore , correlation shall be. verified at least once per 31 EFPD and adjusted at least once per 92 EFPD using the results of the measurements obtained in accordance with Specification 3.17.2.
- c. The excore/incore correlation shall be determined after each fuel loading or major change in excore Power Range instrumentation prior to exceeding 5096 of RATED THERMAL POWER.-
- d. The excore Power Range detectors shall be calibrated / correlated relative to the Movable Incore Detector System measurements within 7 days after completion of incore measurements.
3-31
~
r: 100 - t 90 . on
- U C" 7@..
(-12,65) (10,65) 60- - , 50- - (-30,45.5)
- (20,45.5) g 40" (20,40)
C-30,40) . O
. 30 l
gg. 10- -
- : l l l 0 ' ,
-50 -40 -30 -20 -10 0 i0 2'0 3'O 4'O 5'O i
% OFFSET l
i FIGURE 3.17-2A POWER vs OFFSET, 0-250 EFPD, THREE LOOP E
l * ~ 1 . 1 : t 100-90- - 80- -
]@. .
(-20,65) (10,65) 60- - 50- - (-30,45.5) (20,45.5) C-30.40I
, 40- -
(20,40)
$ 30- -
e a 2g. 10- -
. . . . . n . . . . .
~
-50 -40 -30 -20 -10 0 10 20 30 40 50
% OFFSET FIGURE 3.17-2B POWER vs OFFSET, 250-EOC, THREE LOOP 4
.n . , .
V_ . . r,j;;*v
;,(
s". ' Y;g 4 POWER DISTRIBUTION LIMITS 3.17.2 LINEA _R HEAT GENERATION RATE , F,gUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION - 3.17.2.1 All LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed the ,
,following kilowatt per foot limits for cycle residency time: f/
- a. Less Than 125 EFPD 14.3 kW/f t
.4 , , , 3
- b. 125 To 250 EFPD 14.5 kW/f t
,ff
- c. Greater Than 250 EFPD -
But Less Than END-OF-CORE LIFE ' 15.5 kW/ft APPLICABILITY: f MODE 1, above 40% of RATED THERMAL POWER. ACTION: ^ With the LHGR of any fuel rod exceeding the limits specified above, ' initiate corrective action within 15 minutes and continue corrective action so that the LHGR is within the limits within 2 hours or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER within the next 4 b: ors., SURVEILLANCE REQUIREMENTS [. .
~1. LHGR's shall be determined to be within the above limits by a core power distribution measureinent using the Moveable incore Detector System and in consideration of the factors listed in 2 below:
- a. At least once per 31 EFPD, i
- b. Prior to THERMAL POWER exceeding 80% of RATED THERMAL POWER af ter each fuel loading, and i
- c. After reaching 100% of RATED THERMAL POWER and achieving equilibrium xenon conditions after each refueling. j 3-31a
-,61 -
- w. ..
/ .
,'Y' Mf f .
W :,, -
' , , r:- i.
g _. f' POWER DISTRIBUTION LIMITS , j _ ., _ s , " l;.& SURVEILLANCE REQUIREMENTS (Cettinued) w , s. h 1 l(g { . j.
- 2." Measured values of core power peaking factors used in determining-g LHGRs shall include the following allowances:
+ ,
a.- . Normal power peaking * **, . "fg b. Flux peaking augmentation factors (Power Spike)*, Figure 3.17-3: pp p, 7-s,l c. Measurement uncertainty of 1.05,, ip:, , d. Statistical density factor of 1.012, e.- Engineering factor of 1.02, f.- Stack shortening / thermal expansion factor of 1.007, and ' p -
- g. Power level uncertainty of 1.02.
*ltems a. and b. are chosen at'a core height to' maximize the product.
** Determined using ti>e thimble location which yields the higher total core peaking
, factor. '
.)-
e
,Y , .m t
3-31b ,
.c u, !
1 y i POWER D1STRIBUTION LIMITS- ,
'l
'/L THREE LOOPS OPERATING '
c; n . LIMITING CONDITION FOR OPERATION t 13.17.2.2 All LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed the 'l following kilowatt per foot limits for cycle reriidency times
- a. ' Less Than 125 EFPD . 9.295 kW/f t .
' ;)
I b'. '125 To 250 EFPD 9.425 kW/ft" J c. Greater Than 125 EFPD ' H , But Less Than '
; 5 END-OF-CORE LIFE' 10.075 kW/ft i
APPLICABILITY: MODE 'l', above 40% of. RATED THERMAL POWER. ACTION: . , . . . .. .
. With the LHGR of any fuel rod exceeding the.Ilmits specified above, ,
i initiate corrective action within 15 minutes and continue corrective
- action so that the LHGR is.within the limit within 2 hours or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REQUIREMENTS
- l. LHGR's shall be determined to be 'within the above limits by~a . core power distribution measurement using the Moveable Incore Detector. e System _and in conside' ration of the factors listed in 2 below:
- a. At least once per 31 EFPD,
- b. Prior to THERMAL POWER exceeding 50% of RATED THERMAL POWER after each fuel loading, if the initial power ascension is performed with three loops operating.
; c.
4 . Aftef achieving 65% of RATED THERMAL POWER and achieving. . equilibrium xenon conditions af ter each refucijn6, if the initial power ascension is performed with three loops operating. 6 m 3. 31 -c m h
% 9 L
l
- POWER DISTRIBUTION LIMITS
' SURVEILLANCE REQUIREMENTS (Continued)
. - 2. Measured values of core power peaking factors used in determining .
LHGRs shallinclude the following allowances:
- a. Normal power peaking.* * *,
, . b'. Flux peaking augmentat. ion factors (Power Spike)*, Figure 3.17-3
- c. Measurement uncertainty of'l.05, d.. Statistical density factor of 1.012,
, . e. Engineering factor of 1.02, s
- f. Stack shortening / thermal expansion factor of 1.007, and
- g. . Power level uncertainty of 1.02.
- Items a. and b. are chosen at a core height t'o maximize the product.
** Determined using the thimble. location which yields the higher total core peaking .. p factor.
4 4
- j 1;
i 1 l,
')
l 3-31d 1
i .!i ! 0 _ 2 1
~
1 s n 6 9 s e r o a - C r C f o C R l e e l d - e o . e M3 . t 7 S e9 - k1 I 2 s s i 7 e s pl h c l e Si r n n rp I I i eA , a w o , n t S P5 i o 9 a t r 0 f o h 8 s - i s uP o y pA P t iC a 6 l SW 1 1 l a a . , i n P s x e e A P d r a e no aC k i p nd S aa a ml - l C - l - 3 el
- H e 7 e 1 .t a 4 JS 2 3
E e . R c a U n G I e r - F e f e a R 0 2 0 8 6 4 1 2 0 1 0 0 0 0 0 1 l 1 l . I l l s D %%m. a 2 E
!l !l L
i-I- POWER DISTRIBUTION LIMITS
~ ~
3.17.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FEH i FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.17.3.1 The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, FjH, shall i be limited by the following relationship: 1 FNH f 1.60[1.0 + 0.3 (1 -P)) where: p_ THERMAL POWER
,and RATED THERMAL POWER l
FyH, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR,is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. APPLICABILITY: MODE 1 l AC TION: With FhH outside of the above specified limit:
- a. Within 2 hours either:
- 1. Restore FNH to within the above limit, or
- 2. Reduce THERMAL POWER to less than 65% of RATED THERMAL POWER and reduce the Nuclear Overpower Trip Setpoint to less than or equal to 74% of RATED THERMAL POWER,
- b. Within 24 hours of initially being outside the above limit, verify through incore flux mapping that FhH is restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours; and
- 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 ACTIONS a.2. and/or b. above subsequent THERMAL POWER operation may proceed provided that F H is demonstrated, through incore flux mapping, to be within the a ove limit prior to exceeding the following THERM AL POWER levels:
- 1. A nominal 50% of RATED THERMAL POWER,
- 2. A nominal 75% of RATED THERMAL POWER, and
- 3. Within 72 hours of attaining greater than or equal to 95% of RATED THERMAL POWER.
3-31e
. )
~
- . POWER-DISTRIBUTION LIMITS z
SURVEILLANCE REQUIREMENTS
. i
. The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR shall be '
determined to be 'within the above limit:
- a. Prior to operation above 80% of RATED THERMAL POWER after ?
each fuel loading, and ,
- b. At least once per 31 Effective Full Power Days. .
0 4 i 3-31f
POWER DISTRIBUTION LIMITS THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.17.3.2 The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, F$H, shall ! be limited by the following relationship: l FMH f 1.64 1.0 + 0.3 (0.65 - P)) where: p= THERMAL POWER ,and : RATED THERMAL POWER- , FkH, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR,is defined l as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. APPLICABILITY: MODE 1 1 ACTION: With F[H outside of the above specified limit:
- a. Within 2 hours either:
- 1. Restore FNH to within the above limit, or
- 2. Reduce THERMAL POWER to less than 20% of RATED THERMAL POWER and reduce the Nuclear Overpower Trip Setpoint to less than or equal to 25% of RATED THERMAL POWER.
- b. Within 24 hours of initially being outside the above limit, verify through incore flux mapping that F H is restored to within the above limit, or reduce THERMAL POWER o less than 5% of RATED THERMAL POWER within the next 2 hours; and
- c. Identify and correct the cause of the out-of-limit condition prior to I increasing THERMAL POWER above the reduced THERMAL POWER {
limit required by ACTIONS a.2. and/or b. above subsequent THERMAL { POWER operation may proceed provided that F g is demonstrated, j through incore flux mapping, to be within the a ove limit prior to 1 exceeding the following THERMAL POWER levels:
- 1. A nominal 50% of RATED THERMAL POWER 1
I l 3-31g I i
s .
' POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS
.The NUCL' EAR ENTHALPY RISE HOT CHANNEL FACTOR shall be determined to be within the above limit:-
- a. Prior to operation above 30% of RATED THERMAL POWER 'af ter each fuel lpading if the initial power ascension was performed with three loops operating. ,
- b. At least once per 31 Effective Full Power Days.
l l e 3-31h
POWER DISTRIBUTION LIMITS 3'.17'.4 ' QUADRANT POWER TILT RATIO LIMITING CONDITIONS FOR OPERATION 3.17.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER. ACTION:
- a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT R iTIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED ! THERMAL POWER.
- 2. Within 2 hours either:
a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Nuclear Overpower Trip Setpoint within the next 4 hours.
- 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce i THERMAL POWER to less than 65% of RATED THERMAL POWER within the next 2 hours and reduce the Nuclear Overpower Trip !
Setpoint to less than or equal to 74% of RATED THERMAL POWER within the next 4 hours; and
- 4. Identify and correct the cause of the out-of-limit condition prior to 1 increasing THERMAL POWER; subsequent POWER OPERATION ;
above 50% of RATED THERMAL POWER may proceed provided l that the QUADRANT POWER TILT RATIO is verified within its ! limit at least once per hour for 12 hours or until verified l acceptable at 95% or greater RATED THERMAL POWER. j l l l l 3-311
h.. w-POWER DISTRIBUTION LIMITS
' LIMITING CONDITIONS FOR OPERATION
. ACTION (Continued)
- b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:
- 1. . . Calculate the~ QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or
. b) THERMAL POWER is reduced to less than 50% of RATED -
THERM AL' POWER. -
- 2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes;
- 3. Verify, that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 65% of RATED THERMAL POWER within the -
-next 2 hours and reduce the Nuclear Overpower Trip Setpoint to less than or equal to 74% of RATED THERMAL POWER within the next 4 hours; and 4.' identify and correct the cause of the out-of-limit condition 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 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.
- c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:
- 1. Calculate the QUADRANT POWER TILT RATIO at least on e per hour until either:
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. 3-31j ____________-_________________-____-___A
POWER DISTRIBUTION LIMITS LIMITING CONDITIONS FOR OPERATION ACTION (Continued)
- 2. Reduce THERMAL POWER to less than 20% of RATED THERMAL POWER within 2 hours and reduce the Nuclear Overpower Trip Setpoint to less than or equal to 25% of RATED THERMAL POWER within the next 4 hours; and
- 3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 20% 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 hours or until verified at 95% or greater RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS
- a. The QUADRANT POWER TILT RATIO shall be determined within 7 days af ter completion of the incore power distribution measurements used to determine the initial excore/incore correlation after each fuelloading or major change in excore Power Range instrumentation.
- b. The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by calculating the ratio at least once per 7 days.
- c. The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channelinoperable by using the Movable Incore Detector System at least once per 24 hours.
I f 3-31k j l _ _ _ - _ _ _ _ _ _ _
l L - POWER DISTRIBUTION LIMITS .
' DNB PARAMETERS
/!
LIMITING CONDITION FOR OPERATION 3.17.5 The following DNB-related parameters shall be maintained within the limits shown in Table 3.17-1: - .g a.- . Reactor Coolant System T Colde .
- b. Pressurizer Pressure, and
, c. Reactor Coolant System Flowrate.
APPLICABILITY: MODE l.' ACTION: With any of the above. parameters exceeding'its limit, restore the parameter to within
, 'Its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED .
THERMAL P'OWER within the next 4 hours. SURVEILL ANCE REQUIREMENTS j
- a. Each of the parameters of Table.3.2-1 shall be verified to be within its limit at least once per 12 hours.
1
- b. The RCS total flow rate shall be determined by heat balance within seven EFPD of achieving RATED THERMAL POWER af ter a refueling.
- c. The RCS total flow rate indicators shall be subjected to a CHANNEL ,
CALIBRATION at least once per 18 months. 1 l J
]
j l l 3-31L 1
)
d n A n od ie . t s ao rl eC p) Oe nv i i m d L
- p sV p ig g A -
op F s 0 M . Lt oo 0 p 0 RR S 6 0 0, E E e ep 0 0 4 HW r o 4 5 0 2 9 TO DP 1 h o TL ' 4 2 . 2 E L TA AM S RR T I fE S ol M / i R I
%T E L 1 - T 5D 7 E f E 1
M oT 3 sA s A eR E R L cf x o B A e P n m A i p n% T B s pn i'g g i 0 N oo F s 0 p1 D 0 p 0 oi 0 mf L t a 0 0, a o r r 2 0 6 r s s ue op 4 5 0 2 4 2 Re FO Ec x
/ ? 2 We O
Pi n L p At e e Ms t RR d a r EE l o w HW C o TO T F l aP rL hteA m m e e t t M s y e r s y i eR S u S gE t s nH s t n e n i rT a r a u l o P lo d a C o r e C o l e or z b e r i r r at t o u t o icun c s c lpi s R a e e a e pm E r a r T R P R t e E d d d op M e e e n t t t t R A a a a iE R ic c ic A d id d i mW P n n n LO I I I
- P
L3.17 POWER DISTRIBUTION' LIMITS g
, BASES-The specifications of this section provide assurance of fuelintegrity during Condition 1
. (Normal Operation) and !! (Incidents of Moderate Frequency) events by: :(1) main.taining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteri'a. In addition, limiting the peak linear power density during Condition i events provides -
assurance that the initial conditions assumed for the LOCA analyses are met and the
,. ECCS Interim acceptance criterion limit of 23000F peak cladding temperature is not exceeded. ..
3.17.1 - AXIAL OFFSET l i.' The AXIAL OFFSET specification provides continuous con'firmation of acceptable
' LINEAR HEAT GENERATION RATES (LHCR) during the time interval between incore -
. measurements.
l 3.17.2 ^ LINEAR HEAT GENERATION RATE' Limiting the' peak LINEAR HEAT GENER'ATION RATE (LHGR) during Condition ! I events provides assurance that the initial condition assumed for. LOCA analyses are met and the ECCS Interim acceptance criterion limit of 23000F peak cladding temperature , is not exceeded. These limits are based on a minimum inlet temperature of 5360F at i RATED. THERMAL POWER. q
- 3.17.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR-The limit on the NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Fys) ensures -
that the minimum DNBR limit is not exceeded. The FNH is measurable, but will normally on!y be determined periodically. This periodic surveillance is sufficient to insure that the limits are maintained provided:
- a. The control rod insertion limits of Specification 3.10.2.6 and 3.10.2.7 are maintained, and
- b. The AXIAL OFFSET limits of Specification 3.17.1.1 and 3.17.1.2 are maintained.
' The relaxation of F]H as a function of THERMAL POWER allows changes in the radia!
power shape for all permissible rod insertion limits. The full power limits (1.60 for four
.,. . loop, and 1.64 for threc ' sop) include a 4% incore measurement uncertainty.
1 I l 3-31m F
c . 3 .
.t 4
POW $R DISTRIBUTION LIMITS , i BASES ~ ; 3.17.4 QUADR ANT POWER T1LT RATIO
- The QUADRANT POWER TILT RATIO limit assures that the rad.ial power distribution-
; satisfies the design values used in power capability analysis. Radial power distribution -
measurements are made during STARTUP testing and periodically during power -
. operation. :
- The liinit of 1.02, at which corrective. action is required, provides DNB and LINEAR
' HEAT GENERATION RATE protection with x-y plane power tilts. A limit of 1.02 was
. selected to provide an allowance fer the uncertainty associated with the indicated i
- power tilt.
- The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less !
; than 1.09 is provided to allow identification and correction of a dropped or misaligned !
.7 control rod. In the event such action does not correct the tilt,.the margin for .
uncertainty on the limiting LHGR is reinstated by reducing the maximum allowed power - l
. ,, by 3% for each percent of tiit. - 3 For purposes of m'onitoring QUADRANT POWER TILT RATIO when one excore detector
. it inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power. distribution is consistent with the QUADRANT POWER TILT RATIO. '
The incore detector monitoring is done with at least one set of four quadrant syminetric thimbles. . 9 e 3-31n
s- - .g POWER DISTRIBUTION LIMITS BASES 3.17.5 - DNB PARAMETERS . The limits on the DNB-related parameters assure tnat each of the parameters are maintained within the normal steady-state envelope of operation assumed in the
- 4 transient and accident analyses. The limits are consistent with the accident
, assumptions.'The indicated values of $42.00F and 2000 psig correspond to analytical-limits of 544.10F and 1960 psia, respectively,.with allowances for. measurement
~
uncertainty. The four and three loop flow rates are the values assumed 'in the analyses. ,i
' The measured values are reduced by the appropriate measurement uncertainty prior to
. comparison with the limits., The four loop measurement and uncertainty are based on a heat balance and the results are used to calibrate the RCS flow indicators. The t.hree Lloop flow rat.e will be deemed acceptable if 0.76 times the measured four loop flow rate is greater than the;three loop flow rate requirement. Previous flow rate m_ measurements -
have shown that the three loop flow rate is 0.77 times the four loop flow rate with measurement uncertainties included. . t 9 P O e 3-31o
a i ll ,1 I
- i, 3
l J i Intentionally Lef t Blank i J i 1 1 1
)
i
)
I 1 3-32 through 3-36 l l I
-)
1
\
i l l l 4 j i L l , a l :) l l l* . l \ t. l I t, i l Intentionally Lef t Blank l l I
-i i
3-39 and 3-40
4
' 3.24 SPECIAL TEST EXCEPTIONS ' ,
3.24.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.24.1 The SHUTDOWN MARGIN requirement of Specification 3.10.1.1 or 3.10.1.4 may be suspended for measurement of control rod worth and. SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated rod worth is'available for trip insertion from -
. OPERABLE control rod (s).
APPLICABILITY: -MODE 2. A CTION: -
- a. With any control rod not fully inserted and_ with less than the above reactivity equivalent available for trip insertion, immediately Initiate and continue bort.tlon'at greater than or equal to 30 gpm'of (14000 ppm boric acid solution or its equivalent until the
. SHUTDOWN MARGIN required by Specification 3.10.1.1 or 3.10.1.4
. Is restored.
b'. 'With all control rods fully inserted and the reactor subcritical by ! less than the above reactivity equivalent, immediately initiate and ! continue boration at greater than or equal to 30 gpm of 14000 ppm boric acid solution or its equivalent until.the -SHUTDOWN MARGIN-required by' Specification 3.10.1.1 or 3.10.1.4 is restored. SURVEILLANCE REQUIREMENTS
- a. The position of each control rod either partially or fully withdrawn shall j be determined at least once per 2 hours. '
- b. Each control rod not fully inserted shall b'e demonstrated capable'of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.10.1.1 or .3.10.1.4 i
3-47
a P g i i SPEC 1AL TEST EXCEPTIONS -
~
3 '.24'.2 PHYSICS' TESTS LIMITING CONDITION FOR OPERATION:
?3.24.2 ThE limitations of Specifications 3.10.1.5, 3.10.1.6, 3.10.2.1, 3.10.2.5, 3.10.2.6 and 3.10.2.7 may be suspended ouring the performance of PHYSICS TESTS provided:
a.' The THERMAL POWER does not exceed 5% of RATED. THERMAL P O W E R ,-
- b. The Reactor Trip Setpoints on the OPERABLE' Power Range . ,
channels are set at less than or equal to 25% of RATED THERMAL ' PO.WER, and '
- c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 5150F. .
APPLICABILITY: MODE 2. : . ACTION:
- a. . With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers..
- b. W1.th a Reactor Coolant System operating loop temperature (Tavg) less than 5150F, restore Tava to within its limit within 15 minutes or be in at least HOT STANDBY within the next '15 minutes.
SURVEILLANCE REQUIREMENTS
- a. The THERMAL POWER shall be determined to be less than or equal to j 5% of RATED THERMAL POWER at lea ~st once per hour during PHYSICS TESTS.
- b. Each Power Range channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating PHYSICS TESTS.
- c. The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 3150F at least once per 30 minutes during PHYSICS TESTS.
3-48 1 L _ _ _ _ _ _ - - -
r- . e , , r ,
'SPECIAL TEST EXCEPTIONS 3.24.3 POSITION INDICATION SYSTEM - SHUTDOWN-LIMITING COtIDITION FOR OPERATION
'3.24.3 The limitations of Specification 3.10.2.3 may be suspended during the a performance of individual shutdown and control rod drop time measurements provided; : ,
- a. Only one dutdown or control bank is withdrawn from the fully
- inserted position at a time, and
- b. ' The analog rod position indicator is OPERABLE during the withdrawal of the rods.*
APPLICABILITY: -MODES 3,4, and 5 during performance of rod drop time
.racasurements. .
ACTION: .
~
With the Position InJication Systems inoperable, or more than one bank of rods-withdrawn, immediately open the reactor trip breakers. SURVEILLANCE REQUIREMENTS
, The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours prior to the start of and at least once per 24 hours thereafter during rod drop time measurements by verifying the Bank Average Analog Rod Position Indication System and the Digital Rod Position Indication System agree within 16 steps.
*This requirement is not applicable during the initial calibration of the Analog Rod
- Position Indication Syst'e m provided: (1) Keff si maintained less than or equal to 0.94, and (2) only one shutdown or. control rod bank is withdrawn from the fully inserted uposition at one time, i
l
~
3-49 l ________-_-_________-___-_---_____L
3.24 SPECIAL TEST EXCEPTIONS '- s BASES ' 3.24,1 SHUTD3WN MARGIN ' J l This Special Test Exception provides that a minimum amount of control rod i worth is immediately available for reactivity control when tests are performed
. for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity !
condition occuring as a esult of fuel burnup or fuel cycling operations. 3.24.2 PHYSICS TESTS This Special Test Exception permits PHYSICS TESTS to be performed at les's than or equal to 5% of RATED THERMAL POWER with the RCS Tave slightly lower than normally allowed so that the fundamental nuclear charact'6ristics of the core and related Instrumental. ion can be verified. In order for various characteristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights,which may not normally be allowed by Specification 3.10.2.6 or Specification 3.10.2.7 which in torn may cause the
3.10.1.6.g RCS Tav to fall slightly below the minimum temperature of Specification 3.24.3 POSITION INDICATION SYSTEM-SHUTDOWN This Special Test Exception permits the Position Indication Systems to be inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefo.e, cannot be observed if the Position Indication Systems remain OPERABLE. l I
e 3-50
L L .- , 4.0.--MAIN STEAM ISOLATION VALVES l Applicability: Applies to periodic testing o'f'the main steam isolation. valves. I L Objective: To verify the ability of the main steam isolation valves to close upon signal. L . l E Specification: The main steam isolation valves will be tested each quarter for movement of the valve disc through a distance of approximately one and one-ha'lf inches. This quarterly test may be conducted while-the station is on line. Simul-taneous closure of.all four valves within ten seconds shall be verified each cold shutdown if it had not been done in the previous three months. i Basis: The-main steam isolation valves serve to limit the reactor coolant syst,em cooldown rate and resultant reactivity insercion, whenever the
- moderator coefficient of reactivity is negative, following a main steam break incident. A l closure time of ten seconds has been shown to !
yield acceptable results. The time period between verification of the ten second closure requirement is based on ASME Boiler and Pr' essure Vessel Co.de Section XI, Article IWV-3410(a)(b) and Westinghouse Standard Technical Specifications, for Pressurizer Water Reactors (5/15/76).
Reference:
(1) FDSA -- Section 8.1. (2) ASME Boiler and Pressure Vessel Code Section XI, Article IWV-3410(a) and (b). (3) Standard Technical. Specification 4.0.5. O h 4-18
Docket No. 50-213 B12604 i Attachment 3 Haddam Neck Plant Change-Out Instructions l l I 1 July 1987
i c f i c e p S l a i c t n h n c e e N m T O m d r I o a T C d C n U a R S t T e S s N u I o T h g U n O i. t t E s G ea ~ N Wmr A no I f H C N O I T ) s A i t C I m - i F y t I L t n C y f e a E t a l o P ) s f e S o S t n a C L r e o S m r s i u o A n i t 2 m C I n 2 i) t c i ( I f x s a) N e ~ 1 2 J ag em H D - - Mnt i Re C 2 2 E 0 1 2 2 4 2 t e 3. t 3 y s T ( e e ( S ( S r r D ~ . u u ~ , ' a b c d e f A 3 6 1
- i g
i g 2 3 5 7 8 3 4 4 4 4 4 4 4 O 1 1 2 F F 2 2 2 2 2 3 3 3 3 3 3 3 3 L E R 5 1 E L C Y C t e l e e D 4 0 6 1 2 2 6 2 2 1 f Bo 2 2 r1 e e e r e r t t g u u a b c d 3 6 1 g g 2 3 5 7 8 3 4 4 4 4 4 ea - - - i i - - - - - - - - - - - LP 1 1 2 F F 2 2 2 2 2 3 3 3 3 3 3 - -
i _ f i c e p S l a _ i c t n h n c e e N m T O m d I o r T a C C d n U t a R S T e S s N u I o T h g U n O i t t
. ~
s E e a G N Wmr A I f no H C N y O e . t min I T m e N e un f a rt oo l C o t y s i S I Vt l F cS r a r a I C at en et n o E t n R a )t ae i
. t P
S t 3 l on oo Wcn a r r e L e (3CC mo uC p A s n
) O C I S m n )T I C i nroS 9.
N ~ 1 -. V i oW 3 ( I H - C MBR C 3 E 3 5
- 7. d e T e 3
( (3 at nh D 1 j k 1 r u m n o p q r s t 1 -
]a 5
A 4 4 4 4
- i g 4 4
4 4 4 4 4 4
'3 9 ,
1 1 O 3 3 3 3 F 3 3 3 3 3 3 3 3 3 3 3 3 L E R 5 1 E L C Y C - t e l e e D 4 0 6 2 6 1 f Bo r2 e t e a t g 1 5 ea 3 9 1 1 LP - - - - - - - - - - - - - 3 3 3 3
i c ncnd aaer 3ee we e t eit d oy f i c~ s eRe eman d o ob h a u r t e e i r niPst rbi i t a fi at l S p out c8,h opt af or ne mt qS e 1 ,r r s l a on ne er eS sd 3 oe l ois cDt y c rd ni r n e n i i n Ait c uef not toni
-_' t h yi f eh t n oanw i n c e a ga oCme oe de tf l e e i v oivi nmi e r b Pte
{ N m T. i t rtd O m d tCuso c qef r no ae edl e a eWn g h at v ee r al p I o a l d T C Rmd eio A rn (T hi n d C n r n t i 8 ne t U a 5 eah.t al o cot 0 1 3 ee t R S 1 h d i 1 b T 3 so etoir t f n . w s S e t o3 n o a nI ,lein ce noMot s nh t i s N u o i I o i6 ed pC n d nt s8 i T h t d eS 1 U g n ce31 n al t e e hc3t yo i O i. er ai t r mi) t s t Sneo cvur iib e ef r h s . E e a : . obpnt i rh cr s u7 i i wt
-N G Wmr t et e o o cv c eeis a c at n q1 m i
A no I f NShiTRDt ea n r e r3L i H - C N O I y T t i A i v C t I c F I a e )l C Ro r
.E P 0t S
n t r 1 o L e 3C ( A s n C I I N ~ 1 H 0 C 1 E 3 T - - e D a b c d e g h i j k 1 m n o r 6 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 u
~
A 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
- i g
O 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 F L - E R 5 1
.-E L
C Y C _ t e l e e D - r# 1 0 - 6 0 2 6 1 1 f Bo 3 r3 c e ee a e b d r t t g 6 7 7 7 7 7 7 u 1 1 1 1 1 1 1 g ea - - - - - - - - LP 3 3 3 3 3 3 3 _ - - - - - - - - - F
t n , t n mdthin ee tai o oin nbi r e0 e 0 t t oc t i t t i la1 l a1 s et i c cisd i v3 i v3 _yl Sei nf e iS c S eDe t d u qon u q n l t
.nnepoi d efe le n r e e *o a aet S nd qwdn ei t t c l ea u oePe d c d c i ob rl
- h e aS ne aS n o spct o ai r
d
,e t
n e ~ d n d n h c Car rh o i nol pc18b ei ei e o enc hc n s N m t t T a c ri eAi3 t e de ede s .h O m l l d r a su sT 1 i I o et a et d ar a es t 7l T C C ,dno r no d n Rer ne d7h r a 1.i c1 o 3t r U ep ep a 0 ,Pmd 3h n
- 2. d e n nwn o erl erl t R oo b oc o T
S wbs ncr
, ain t s ai nno tr S
e 3n auSt nit r a of.oC i ot i N o s u ,q cgencd I s h n sC t h nt sC o i oee r re sSi e nsa T n o n ty ctut uodiinngn h t U o ey t ie g ic miv t t n e a nh ea c miv i O e ei eei i t . Sr egt md eotr l E s rt s rt s e a t
- e a nt aedt e e pvt i
ic G suicq ae sua Wmr oeq uere cotui mis mr n no9tNT l N i iqe h eR Tr h e TrR eWf oh nM A I f t i H sf C ~ N O I y T i t
).
A vt n C i r n c o t I eo F aC wit I el ou
)
C Ro r t P b )s E n ' i t P 0 t e ) 7 tr i S n g 1 si m t r 1 o m n n, 3 i L e 3 C ( i i l ( DL A s a e n t u C I n f A B A B I ~ _ o e N 2
- C R 1
1 2 2 H 0 1 3 7 8 7 7 C 1 1 1 1 1 1 1 E 3 3
/
3 3 3 3 3 T e ( ( e e e e r q r s j r r r r D u 7 7 7 '8 9 3 4 7 3 0 u u 1 u u A g 1 1 1 1 1 2 2 2 2 3 g g 3 g g i - - - - - - - - - - i i - i i O F 3 3 3 3 3 3 3 3 3 3 F F 3 F F L E R 5 - s 1 t
)
s n E e e L i 3 i ic l C a f Y m f C o e n o e A C t y l l e t a e i v m )y D i t r t ei c v a ht i 4 e o s t c 0 R I a 6 6 6 e 2 - 5 R _ 1 f 1 1 Bo 3 3 f _ rto ( ( o e e ] ] t t g 3 9 3 4 7 3 0 1 2 2 2 2 3 1 1 3 ea - - - - - - - - - LP _ - - 3 3 3 3 3 3 3 - - 3 -
yr
- t o ep g:.
r r o i m t n c ei n m_ n - ~ i ts n e e t nb g i e n d e t n ev N t a eh O m l I o est T C d n . C ne7 U em1 R e e3 b r T sid N S ha qa un I e T nr0
, ot 1 U .
i t n3 O cele n E s ao vi G sit N i h qe uc A TeS H _C N I O T A ) _C r nt I eon F I wio t ouC C Pb i s t E r P 7 t i S t 1 i s m r i L e 3DL
$0AC 1
^ I s
n ( 6 3 I ' g 3 N 3
- 1 h H 7 - g C 1 7 u E 3 1 o r
T e 3 h t D b t d r e f g h 1 j ~k 1 le m n o 2 1 1 1 u 1 1 1 1 1 1 1 1 1 1 1 2 wA 3 3 3
- i g 3 3
3 3 3 3 3 3 b a 3 3 3 3 e .C 3 3 3 F 3 3 3 3 3 3 3 3 T 3 3 3 3 - L E R 5 1 d E n L a C r eog n) Y C win t our i e Pb i ot t 8 ri le 1 sit no e D - ( 3DM 4 1 - 0 - 6 2 6 7 1 1 f Bo 3 r5 e te e r ' t g u g 2 3 3 3 _ 1P 2a 3_ - - - i F - - - - - - - - - - - - 3 3
;l
, 1lll l;
-_ i n nz s e i d d ol a ne i t n, ata am cr -. yr i o t o f f _ ep i _ r r cd e pan i t o - n c n S ei l w s. _ i ts ne aet cnne e i t i nb hln ym n d e cer e e ev ei r i N m t a eh Ttn uq _ O m l est de e I o r asr T C d n i on C d ne nn U ? em aoi R e ti t br e S t p T ee cc ss c si S N
- - a u q x I
h e . u se nr7 oi T ot 1 hht U i t n3 gT n es O , ce n l i t .t E ea s vi o s e tl aa G sit uc Wmic ~ N ,4 i h qe r e ' - n op - A - TeS if s - H +:c- r
, C, N
O I T A
't
- C F
I I 1 C t E s P e s S t r T mv)e a , L e l el A s a) S Va t n i s cn C I I eo n n i 0 pi N - 4 S t p ao H 3 4 e Mita C 2 c E d x 9. o l n 3 E ( 4 s Js D .- 9 a I
'7 8 9 0
( I 8 - A 3 4 4 4 5 1 O - - - - - - - - - - - 3 3, 3 3 4 4 L E R _, 5 ) m 1 dt e nn - t s f, l aC o y C . rn eog Sd n Y t n a C - win t i a e _ our l e Pbo or t 8 i t ri ou e t n Ct' a l e 1 s o i y r r D 3DM ( o e
]'
~
d , a a j t p cm) a
, 4 b l
c 2 b 2 2 c eee 0 6 ., l I
- - - - R Tr ,su -
2 6 8 8 8 8 8 8 -
. 1 f 1 1 1 1 1 1 0 ws -
. B o g 'm 3 3 3 3 3 3 2. loer 3
r6 e e e e e e (FP te e a r r r r r r t g 4 4 5 . 6 ug ug u g u g u g u g 9 0, 3 3 3 3 . 3 3 4 1 ea - - - - i i i i i i - - - LP 3 3 3 3 F F F F F F 3 3 - - - - 4
_a cr yr t o i o
%. w ep r r i
f cd f i t o e n __ 1 n c pa - ei n S - l w s. t s n aet
~ i e c nne n e
; i nym t i b hl
= n d e cer e
~ : e ev ei r T_tn uq i
N m t a => O m l eh dr e e - I o est asr T C d n i C ne d n U nn o em 1 aoi R V- e b er ti t t p T Sc mN S I si a u h e . nr7 q ee s s c u s e oi e x
~
T ot 1 hht U i t n3 gT es n O ce n l i t .t E eao s vi s e t l aa G N i sit h qe uc Wmi r e c nop
- A -
TeS if s
.H C .
.N
,O I
T A _. _C I _h. FI C E ' t s P e s m v)e S t T r a L e l - el a) S Va t A s n i s
.'C I cn n n
' I eo N 0 4
pi S tp iao H ~ 4 e Mita C 3 . E - d 2. cx - 9. o l
. T n 3
( E 4 s a ( I D 9 7 8 9 0 8 A - 3 4 4 4 5 1 O - - - - - - - - - - 3 3 3 3 3 4 L - E R - 5 )
. m 1
d t o e E n aC n o t s L y C r eogn Sd n Y t win n a,
,C t our i a
e
~
l e Pb ito or
._ t 3 ri ou e t n Ct a l
e 1 i so r r
- - D 3DM
( o t p e 1 ca m )' 4 0 l a b l e. 2 a b 2 2 c e ee 6 - - l
- - - - R Tr ,su 2 6.
1 f 8 1 8 1 3 1 3 1 8 1 3 1 0 ws Bo 3 3 3 3 3 3 2 oe r l 3 FP
-e t e r6 a
e r e r e r e e e ( r r r
~
t g 4 4 5 6 u g u g u g u g u g u g 9 0 8 3 3 3 3 3 4 ea - - - - i i i i i i - - 1 I LP -
- 3 3 -
3 3 F F F F F F 3 3 - - - - 4
^
^}}