ML20214Q111
| ML20214Q111 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/22/1987 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20214Q102 | List: |
| References | |
| NUDOCS 8706040236 | |
| Download: ML20214Q111 (18) | |
Text
.
1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING D.
Whenever the reactor is in the In the event of operation with a cold shutdown condition with maximum fraction of limiting power irradiated fuel in the reactor density (MFLPD) greater than the vessel, the water level shall not fraction of rated power (FRP), the be less than 12 in. above the top setting shall be modified as of the normal active fuel zone.
follows:
FRP S < (0.58W + 62%)
MFLPD 2 Loop
- Where, FRP - fraction of rated thermal power (1998 MHt)
MFLPD - maximum fraction of limiting power density where the limiting power density is 13.4 KH/ft for all fuel.
l The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
For no combination of loop recirculation flow rate and core thermal powe. shall the APRM flux scram trir setting be allowed to exceed l'_0% of rated thermal power.
F. APRM Flux Scram Trip Setting (Refuel or Start and Hot Standby Mode) i When the reactor mode switch is in the REFUEL or STARTUP position, the APRM scram shall be set at less than or equal to 15% of rated power.
1 shall be B.
APRM Rod Block Trip Setting The APRM rod block trip setting shall be:
~
8706040236 870522 PDR ADOCK 05000293 P
PDR Sn. 1 0.58H + 50% 2 Loop 7
1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING
- Where, S..
Rod block setting in percent of rated thermal power (1998 MWt)
Percent of drive flow W
=
required to produce a rated core flow of 69 Mlb/hr.
In the event of operating with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
FRP S < (0.58W + 507.)
MFLPD 2 Loop
limiting power density where the limiting power density is 13.4 KW/ft for all fuel.
l The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
C.
Reactor low water level scram setting shall be 1 9 in, on level Instruments, l
l D.
Turbine stop valve closure scram settings shall be 1 10 percent valve closure.
E.
Turbine control valve fast closure setting shall be 1 150 psig control oil pressure at acceleration relay.
F.
Condenser low vacuum scram setting shall be > 23 in. Hg. vacuum.
i G.
Main steam isolation scram setting shall be i 10 percent valve closure.
8 I
NOTES FOR TABLE 3.2.A 1.
Whenever Primary Containment integrity is required by Section 3.7, there shall be two operable or tripped trip systems for each function.
2.
Action If the first column cannot be met for one of the trip systems, that trip system shall be tripped.
If the first column cannot be met for both trip systems, the appropriate action listed below shall be taken.
A.
Initiate an orderly shutdown and have the reactor in Cold Shutdown Condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have Main Steam Lines isolated within eight hours.
C.
Isolate Reactor Water Cleanup System.
D.
Isolate Shutdown Cooling.
3.
Instrument set point corresponds to 128.26 inches above top of active fuel.
4.
Instrument set point corresponds to 77.26 inches above top of active fuel.
5.
Not required in Run Mode (bypassed by Mode Switch).
6.
Two required for each steam line.
7.
These signals also start SBGTS and initiate secondary containment isolation.
8.
Only required in Run Mode (interlocked with Mode Switch).
9.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of hydrogen injection test with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the test.
The background radiation level and associated trip setpoints may be adjusted during the test based on either calculattens or measurements of actual radiation levels resulting from hydrogen injection.
The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal radiation levels after completion of hydrogen injection and prior to establishing reactor power levels below 20% rated power.
46
. NOTES FOR TABLE 3.2.8 1.
Whenever any CSCS subsystem is required by Section 3.5 to be operable, there shall be two (Note 5) operable trip systems.
If the first column cannot be met for one of the trip systems, that system shall be repaired or the reactor shall be placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after this trip system is made or found to be inoperable.
2.
Close isolation valves in RCIC subsystem.
3.
Close isolation valves in HPCI subsystem.
4.
Instrument set point corresponds to 77.26 inches of active fuel.
l 5.
RCIC and HPCI have only one trip system for these sensors.
53
PNPS TABLE 3.2-G INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIP AND ALTERNATE R00 INSERTION Minimum Number of Operable or Tripped Instrument Channels Per Trip System (1)
Trip Function Trip Level Setting 2
High Reactor Dome 1175 1 15 PSIG Pressure 2
Low-Low Reactor
> 77.26 inches l
Water Level above the top of the active fuel Actions (1)
There shall be two (2) operable trip systems for each function.
(a)
If the minimum number of operable or tripped instrument channels for one (1) trip system cannot be met, restore the trip system to operable status within 14 days or be in at least hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(b)
If the minimum operability conditions (l.a) cannot be met for both (2) trip systems, be in at least hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
59a
BASES:
3.2 In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.
This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.
The objectives of the Specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.
When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses.the conditions for which isolation is required.
Such instrumentation must be available whenever primary containment integrity is required.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
The low water level instrumentation set to trip at 128.26 inches above l
the top of the active fuel closes all isolation valves except those in Groups 1, 4 and 5.
Details of valve grouping and required closing times are given in Specification 3.7.
This trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve closing time.
Required closing times are less than this.
The low low reactor water level instrumentation is set to trip when reactor water level is 77.26 inches above the top of the active fuel I
(-49" on the instrument).
This trip closes Main Steam Line Isolation 68
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
- 2. The SRM shall have a minimum Spiral Reload of 3 cps except as specified in 3 and 4 below.
During spiral reload, SRM operability will be verified by
- 3. Prior to spiral unloading, the using a portable external source SRM's shall have an initial every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until the required count rate of'>3 cps. During amount of fuel is loaded to spiral unloading, the count maintain 3 cps. As an rate on the SRM's may drop alternative to the above, up to below 3 cps.
two fuel assemblies will be loaded in different cells
- 4. During spiral reload, each containing control blades around control cell shall have at each SRM to obtain the required 3 least one assembly with a cps. Until these assemblies have minimum exposure of 1000 loaded, the cps requirement is MHD/ST.
not necessary.
l C.
Spent Fuel Pool Water Level C. Spent Fuel Pool Water Level Whenever_trradiated fuel is Whenever irradiated fuel is stored in the spent fuel pool, stored in the spent fuel pool, the pool water level shall be the water level shall be maintained at or above 33 feet.
recorded daily.
D.
Multiple Control Rod Removal D. Multiple Control Rod Removal Any number of control rods and/or Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the control rod drive mechanisms may start of removal of control be removed from the reactor rods and/or control rod drive pressure vessel provided that at mechanisms from the core least the following requirements and/or reactor pressure vessel are satisfied until all control and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> rods and control rod drive thereafter until all control mechanisms are reinstalled and rods and control rod drive all control rods are fully mechanisms are reinstalled and inserted in the core.
all control rods are fully inserted in the core, verify a.-The reactor mode switch is that:
operable and locked in the Refuel position per
- a. The reactor mode switch is Specification 3.10.A, except operable and locked in the that the Refuel position "one Refuel position per rod out" interlock may be Specification 3.10.A.
bypassed, as required, for those control rods and/or
- b. The SRM channels are control rod drive mechanisms operable per Specification to be removed, after the fuel 3.3.8.4.
assemblies have been removed as specified below.
- c. The Reactivity Margin requirements of
Specification 3.3.B.4.
- c. The Reactivity Margin requirements of Specification 3.3.A.1 are satisfied.
203
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 8.
Linear Heat Generation Rate (LHGR) 8.
Linear Heat Generation Rate (LHGR)
Ouring reactor power operation The LHGR as a function of core the linear heat generation rate height shall be checked dail:.
(LHGR) of any rod in any fuel during reactor operation at assembly at any axial location
>25% rated thermal power.
shall not exceed 13.4 kw/ft for all fuel.
l If at any time during operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore. operation to within the prescribed limits.
If the LHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
TABLE 3.11-1 OPERATING LIMIT MCPR VALUES A.
MCPR Operating Limit from Beginning of Cycle (B0C) to BOC + 7,513 MHD/ST.
P8x8R/BP8x8R For all values of t 1.36 B.
MCPR Operating Limit f'om BOC + 7,513 MHD/ST to End of Cycle.
T P8x8R/BP8x8R T1 0 1.39 0.0.<T 1 0.1 1.40 0.1 < T 1 0.2 1.40 0.2 <T 1 0.3 1.41 0.3 < T 1 0.4 1.41 0.4 <T 1 0.5 1.42 0.5 <T i_C.6 1.42 0.6.< T 10.7 1.43 0.7 <T 1 0.8 1.43 0.8 < T 10.9 1.44 0.9 <T 1 1.0 1.44 2058-2
BASES:
- 3. llc MINIMUM CRITICAL POWER RATIO (MCPR)
Operating Limit MCPR For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.1.
l The difference between the specified Operating Limit MCPR in Specification 3.11C and the Safety Limit MCPR in Specification 1.1A defines the largest reduction in critical power ratio (CPR) permitted during any anticipated abnormal operating transient.
To ensure that this reduction is not exceeded, the most limiting transients are analized for each reload and fuel type to determine that transient which yields the' l
largest value of ACPR.
This value, when added to the Safety Limit MCPR must be less than the minimum operating limit MCPR's of Specification 3.11.C.
The result of this evaluation is documented in the
" Supplemental Reload Licensing Submittal" for the current reload.
The evaluation of a given transient begins with the system input parameters shown in Table 5-4, 5-6 and 5-8 of NEDE-240ll-P(,
Supplemented by reload unique inputs given in the current Supplemental Reload Licensing Submittal. These values are input to a GE core dynamic behavior transient computer program described in NED0-10802(*'.
The transient code used for all pressurization events is described in NEDE-24154-P (Reference 5).
The MCPR analysis for pressurization events is done in accordance with the procedures given in Reference 6.
1
)
l l
205C-3 l
FIGURE 3.11-1 MAPLHGR Versus Planar Average Exposure Fuel Type 8DB219L Core Flow < 90% rated Core Flow >,= 90% rated 13 1
12.3 12.1 12.1 l
11.9 I
12 11.3 Maximum 11.5
(
1 11.3 -
l Average Planar g
\\
10.2 10.8 10.9 Linear Heat 10.7 l
Generation
\\
9.6 0
f Rate (kw/ft) l 1000 9.0 l
9.7 9
f 200 9I
> 8.6 m
8 h
0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 i
Planar Average Exposure (MWD /ST) i
FIGURE 3.11-2 MAPLHGR Versus Planar Average Exposure Fuel Type 8DB219H Core Flow < 90% rated
- Core Flow >,= 90% rated 13 I
12.3 I
1 12.2 12.1 11.8
/
Nl u
m
.l 12 33 g
11.3 l
11.2
[11.6 11 s
11.5
\\
Maximum
)
~
Average Planar 11.2 h
10.2
)
Linear Heat 10.7 Generation
\\
9.6 Rate (kw/ft) h \\l y1000 9,g gy 200 9,1
' 8.6 8
i g
m 0
5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 6
Planar Average Exposure (MWD /ST) i
FIGURE 3.11-3 MAPLHGR Versus Planar Average Exposure Fuel Type 8DB262 Core Flow < 90% rated Core Flow >,= 90% rated l
3 I
12.2 I
i 12.1 i
12.1 11.9 l
l
^
l 11.6 1
11.3 l
l Maximum
/
/
.5 11.5 10.7 I
Average Planar
[
h l
10.6 i/
Linear Heat 10.7 9.8 Generation
(
Rate (kw/ft) 9 10.2 9.2 1000 i
(
200 9.3 b 8.7 8
g 0
5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 '.5,000 i
m Planar Average Exposure (MWD /ST)
FIGURE 3.11-4 MAPLHGR Versus Planar Average Exposure Fuel Types P8DRB265L and BPDRB265L Core Flow < 90% rated
- Core Flow >,- 90% rated
'3 I
I I
12.1 12.1 12.1 33,9 12-11.3 11.6 l
7 c
Maximum 1.5 11.5 11.5
\\
10.7 4
Average Planar 11-11.0 1.0 10.2 Linear Heat I
l Generation 10.7
(
\\
96 Rate (kw/ft)
V j
1000 10.2 9.7 200 ro 8
l S
0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000
[
Planar Average Exposure (MWD /ST)
FIGURE 3.11-5 i
MAPLHGR Versus Planar Average Exposure Fuel Types P8DRB282 and BP8DRB282 Core Flow < 90% rated
- Core Flow >,= 90% rated 13 g
g l
11.8 12.1 11.8 I
I l
12 l
11.
11.3 l
11 1 j d f,4 Maximum M
11, q
11.2
[0.6 11.2 \\
Average Planar l
s Linear Heat 10.6 H
9.8 Generation 10.7 10.
10 N
l Rate (kw/ft) 3I1000 i
9.9 i
9-r 9.3 l
200 l
i m
8 8
o 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 m
6, Planar Average Exposure (MWD /ST)
I f
FIGURE 3.11-6 MAPLHGR Versus Planar Average Exposure j
Fuel Types P8DRB265H and BP8DRB265H i
Core Flow < 90% rated
- Core Flow >,= 90% rated I
]
13 l
l 12.1 12.1 11.9 11.9 I
I J
12 -- 1 1.6 l
[
11.3 l
Maximum
/
1.5 11.
\\
\\ 10.7 j
l Average Planar Linear Heat 11.0 11.3 h N
10.2
'F Generation 10 9.6 Rate (kw/ft) 9 10.2 1000 I
9.7
)
9 7
I 200 91 l
to 8
h 0
5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 Planar Average Exposure (MWD /ST)
4 FIGURE 3.11-7 MAPLHGR Versus Planar Average Exposure Fuel Type BP8DRB300 Core Flow < 90% rated Core Flow >,= 90% rated 13 i
12.3 12.0 8
12 1
/
k Maximum 5
10.4
/
10.8 10 Linear Heat Generation 10.5 N
\\
Rate (kw/ft)
D N
3 1000 9.8 7200 8
0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 m
Planar Average Exposure (MWD /ST) s
_. - =
5.0 MAJOR DESIGN FEATURES 5.1 SITE FEATURES Pilgrim Nuclear Power Station is located on the Mestern Shore of Cape Cod Bay in the Town of Plymouth, Plymouth County, Massachusetts. The site is located at approximately 41'51' north latitude and 70*35' west longitude on the Manomet Quadrangle, Massachusetts, Plymouth County 7.5 Minute Series (topographic) map issued by U.S. Geological j
Survey. UTM coordinates are 19-46446N-3692E.
, The reactor (center line) is located approximately 1800 feet from the nearest property boundary.
5.2 REACTOR A.
The core shall consist of not more than 580 fuel assemblies.
B.
The reactor core shall contain 145 cruciform-shaped control rods. The control materials shall be either boron carbide
~
powder (B C) compacted to approximately 70% of theoretical 4
density or a combination of boren carbide powder and solid hafnium.
5.3 REACTOR;1ESSEL The reactor vessel shall be as described in Table 4.2.2 of the FSAR.
The applicable design codes shall be as described in Table 4.2.1 of the FSAR.
5.4 CONTAINMENT A.
The principal design parameters for the primary containment shall be as given in Table 5.2.1 of the FSAR. The applicable design codes shall be as described in Section 12.2.2.8 of the FSAR.
B.
The secondary containment shall be as described in Section 5.3.2 of the FSAR.
~
C.
Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.
y l
t 206m 1