ML20137C108
| ML20137C108 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/06/1986 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20137C081 | List: |
| References | |
| NUDOCS 8601160195 | |
| Download: ML20137C108 (29) | |
Text
_. -
9 TABLE OF CONTENTS
- .. Page No.
1.0 DEFINITIONS 1 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 1.1 1 FUEL CLADDING INTEGRITY 2.1 6 1.2 REACTOR COOLANT ~ SYSTEM INTEGRITY 2.2 22 LIMITING-CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.1 - REACTOR PROTECTION SYSTEM 4.1 26 3.2 PROTECTIVE INSTRUMENTATION 4.2 42
~ 3.3 REACTIVITY CONTROL 4.3 80 A. Reactivity Limitations A 80 B. Control Rods B 81 C. Scram Insertion Times C 83 D.' Control Rod Accumulators D 84 E. Reactivity Anomalies E 85 F. Alternate Requirements 85 G. Scram Discharge Volume G 85 3.4 . STANDBY LIQUID CONTROL SYSTEM 4.4 95 A. Normal System Avai~ ability A 95 B. Operation with. Inoperable Components B 96 C. Sodium Pentaborate Solution C 97 D. Alternate Requirements 97
-3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 103 A. Core Spray and LPCI Subsystems A 103 B. Containment Cooling Subsystem B 106 C. HPCI Subsystem C 107 D. RCIC Subsystem D 108 E. Automatic Depressurization System E 109 F. Minimum Low Pressure Cooling System and Diesel Generator Availability F 110 l G. (Deleted) G 111 H. Maintenance of Filled Discharge Pipe H 112 3.6 PRIMARY SYSTEM B0UNDARY 4.6 123 A. Thermal and Pressurization Limitations A 123 B. Coolant Chemistry B 124 C. Coolant Leakage C 125 D. Safety and Relief Valves D 126 E. Jet Pumps E 127 F. Recirculation Pumps F 127 l
! G. Structural Integrity G 127 H. High Energy Piping (Outside Containment) H 127b I. Shock Suppressors (Snubbers) I 137a Amendment No. 8601160195 84d3106
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o B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 90, are hereby incorporated in the license.
The licensee shall operate _ the facility in accordance with the Technical Specifications.
C. Records Boston Edison shall keep facility operating records in accordance with the requirements of the Technical Specifications.
D. Equalizer Valve Restriction - DELETED E. Recirculation Loop Inoperable - DELETED F. Fire Protection The licensee may proceed with and is required to complete the
, modifications identified in Paragraphs 3.1.1 through 3.1.19 of the NRC's Fire Protection Safety Evaluation (SE), dated Decemoer 21, 1978 for the facility. These modifications will be completed in accordance with the schedule in Table 3.1.
In addition, the licensee shall submit the additional information identified in Table 3.2 of this SE in accordance with the schedule contained therein. In the event these dates for submittal cannot be met, the licensee shall submit a report, explaining the circumstances, together with a revised schedule.
The licenses is required to implement the administrative controls identified in Section 6 of the SE. The administrative controls shall be in effect by December 31, 1978.
G. Physical Protection The licensee shall fully implement and maintain in effect all
- provisions of the following Commission approved documents, including 4
amendments and changes made pursuant to the authority of 10CFR50.54(p):
(1) " Security Plan for Pilgrim Nuclear Power Station," dated November 7, 1977 with Revision 2 dated May 26, 1978 and Revision 3 dated January 8, 1979.
r.
1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING D
~ 1.1 FUEL ~ CLADDING INTEGRITY 2.1 FUEL CLA00ING INTEGRITY
. Applicability: Applicability:
Applies to the interrelated Applies to trip settings of the variables associated with fuel instruments and devices which thermal behavior, are provided to prevent the reactor system safety limits from being exceeded.
Objective: Objective:
To establish limits below which To define the level of the the integrity of the fuel process variables at which cladding is preserved. automatic protective action is initiated to prevent the fuel cladding integrity safety limits from being exceeded.
Specification: Specification:
A. Reactor Pressure >800 psia and A. Neutron Flux Scram Core Flow >10% of Rated The existence of a minimum The limiting safety system trip critical power ratio (MCPR) less settings shall be as specified than 1.07 for two recirculation below:
loop operation (1.08 for single loop' operation greater than 24 1. Neutron Flux Trip Settings hrs.) shall constitute violation of the fuel cladding integrity a. APRM Flux Scram Trip safety limit. This MCPR of Setting (Run Mode) l 1.07 (or 1.08) is hereinafter referred to as the Safety Limit When the Mode Switch is MCPR. In the RUN position, the APRM flux scram trip B. Core Thermal Power Limit (Reactor setting shall be:
Pressure 1800 psia and or Core Flow 110%) S 1 58W + 62% - X When the reactor pressure is 1800 psia or core flow is less than or equal to 10% of rated, the steady state core thermal power shall not exceed 25% of design thermal power.
Amendment No. 6
7-x
. 1.1 SAFETY LIMIT' 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 : FUEL CLADDING INTEGRITY (Cont'd) 2.1 FUEL CLADDING INTEGRITY (Cont'd)
C. . Power Transient Where:
L The safety limit shall be assumed S - Setting in percent of rated to be exceeded when scram is thermal power (1998 MWt) known to have been accomplished by a means other than the W - Percent of drive flow to expected scram signal unless produce a rated core flow of analyses demonstrate that the 69 M lb/hr.
fuel cladding integrity safety ,
i limits. defined in Specifications X - 0 for two recirculation loop 1.1A and 1.1B were not exceeded operation and single loop during the actual transient. operation less than 24 hrs.
D. Whenever the reactor is in the X - 2.9 for single loop operation cold shutdown condition with greater than 24 hrs.
Irradiated fuel in the reactor vessel, the water level shall not In the event of operation with a be less than 12 in. above the top maximum fraction of limiting power of the normal active fuel zone. density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
~ ~
FRP S < (0.58W + 62% - X) MFLPD Where, FRP - fraction of rated thermal power (1998 MWt)
MFLPD - maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for 8x8 and P8x8R fuel.
W and X are defined above (2.1.A.I.a).
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual
- operating value is less than the l design value of 1.0, in which case the actual operating value will be
! used.
l l
1 Amendment No. 7 i
F,
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. _ s 1.1= SAFETY LIMIT ~ 2 '.1 LIMITING SAFETY SYSTEM SETTING For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to
- exceed 120% of rated thermal power.
Standby Mode)
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.
shall be 1120/125 of scale.
B. APRM Rod Block Trip Setting The,4PRM rod block trip setting shall be:
S.. $ 0.58W + 50% - X Where, S,. - Rod block setting in percent of rated thermal power (1998 MHt)
H and X are defined in Spec.
2.1.A.1.a.
In the event of operating with a maximum fraction limitirg power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as-follows:
FRP S i (0.58W + 50% - X) MFLPD i A
\
l Amendment No. 8 k.
r-e 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING Where, FRP = fraction of rated thermal power MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for 8x8 and P8x8R fuel.
H and X are defined in Spec.
2.1.A.I.a.
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.
D. Turbine stop valve closure scram settings shall be i 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 1 23 in. Hg. vacuum.
G. Main steam isolation scram setting shall be 1 10 percent valve closure.
H. Main steam isolation on main team line low pressure at inlet to turbine valves. Pressure setting shall be 1 880 psig.
I. Reactor low-low water level initiation of CSCS systems setting shall be at or above -49 in.
Indicated level.
Amendment No. 9 (the next page is 11)
BASES:
1.1 . ~ A Fuel Cladding Integrity (Cont'd)
The required input to the statistical model are the uncertainties listed on Table 5-1, Reference 3, the nominal values of the core
. parameters listed in Table 5-2, Reference 3, and the relative assembly power distribution shown in Figures 5-1 and 5-1A of Reference 3. Tables 5-2A and 5-28, Reference 3, show the R-factor distributions that are input to the statistical model which is used to establish the safety limit MCPR. The R-factor distributions shown are-taken near the beginning of the fuel cycle.
The basis for the. uncertainties in the core parameters are given in 1 . NE00 20340<2. and the basis for the ur. certainty in the GEXL
. correlation is given in NED0-10958' * . The power distribution is Dased on a typical 764 assembly core in whicn tne rod pattern was aroltrarily chosen to produce a skewed power distribution naving the greatest number of assemblies at the nignest cower leveis. The worst destribution in Pilgrim Nuclear Power Station Unit I during any fuel cycle would not be as severe as the distribution used in the analysis.
For single loop operation the safety limit MCPR is increased by 0.01 as discussed in Reference 4.
- 8. Core Thermal Power Limit (Reactor Pressure < 800 psig or Core Flow
< 10% of Rated)
The use of the GEXL correlation is .1ot valid for the critical power calculations at pressures below 800 psig or core flows less than 10%
of rated. Therefore, the fuel cladding integrity safety limit is established by other means. This is done by establishing a limiting condition of core thermal power operation with the following basis.
Since the pressure drop in the bypass region is essentially all elevation head which is 4.56 psi the core pressure drop at low power and all flows will always be greater than 4.56 psi. Analyses show that with a flow of 28x10' Ibs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with.a 4.56 psi driving head will be greater than 28x10' lbs/hr. Irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to 800. psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors the 3.35 MWt bundle power correspeads to a core thermal power of more than 50%. Therefore a core thermal power limit of 25%
for reactor pressures below 800 psia, or core flow less than 10% is conservative.
Amendment No. 12
U 1.1' BASES:
1C. Power Transient Plant safety analyses have shown that the: scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 1.1A or 1.18 will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate.
The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux.following closures of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.
-The_ computer provided with Pilgrim Unit-I has a sequence annunciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc., occur. This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.
D. Reactor Water Level (Shutdown Condition)
During periods when the reactor is shutdown, consideration must also be given.to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above the top of the fuel provides adequate margin. This i
level will be continuously monitored.
References
- 1. General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, General Electric Co. BWR Systems Department, November 1973 (NED0-10958).
- 2. Process Computer Performance Evaluation Accuracy, General Electric Company BHR Systems Department June, 1974 (NEDO-20340).
- 3. General Electric Boiling Water Reactor Generic Reload Fuel Application, NEDE-240ll-P.
- 4. " Pilgrim Nuclear Power Station Single Loop Operation," NED0-24268, June 1980 with Errata and Addenda Sheet No. 1, September 1980.
Amendment No. 13
[
.2.1 BASES:
.In summary:
-1.~ The abnormal operational transients were analyzed to a power level of 1998 MWt.
- 11. The licensed maximum power level is 1998 MWt.
- 111. Analyses of transients' employ adequately conservative values of the controlling reactor parameters.
iv. The analytical procedures now used result in a more logical answer than the alternative method of assuming a higher starting power in
, conjunction with the expected values for the parameters.
As-discussed in Reference 2. the core wide transient analyses for one recirculation pump oceration is conservatively bounded by two-loop operation analyses anc tne flow-dependent rod block and scram setpoint equations are adjusted for one-oumo coeration.
The bases for individual set points are discussed below:
A. Neutron Flux Scram Trip Settings APRM The average power range monitoring (APRM) system, which is calibrated using neat balance data taken during steady-state conditions, reads in percent of design power (1998 MWt). Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantar.eous neutron flux due to the time constant of the fuel.
Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrated that with a 120 percent scram trip setting, none of the abnormal operational transients analyzed violate the fuel safety limit and there is a substantial margin fr 1 fuel damage. Therefore, the use of flow referenced scram trip provides 'even addltional margin.
I An increase in the APRM scram setting would decrease the margin present before the the fuel cladding integrity safety limit is reached. The APRM scram setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during i operation. Reducing this operating margin would increase the l frequency of spurious scrams, which have an adverse effect on reactor safety because ot the resulting thermal stresses. Thus, the APRM setting was selected because it provides adequate margin for the fuel cladding integrity safety limit yet allows operating margin that reduces-the possibility of unnecessary scrams.
Amendment No. 15
2.1 BASES
W setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system. As with.the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum fraction of limiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin. The definition of the single loop setpoints is given in the FSAR.
C. Reactor Water Low Level Scram Trip Setting (LLll The setpoint for low level scram is above the bottom of the
. separator skirt. This level has been used in transient analyses t dealing with coolant inventory decrease. The results show that scram at this level adequately protects the fuel and the pressure
. . barrier,- because MCPR remains well above the safety limit MCPR in all cases, and system pressure does not reach the safety valve settings. The scram setting is approximately 25 in. below the normal operating range and is thus adequate to avoid spurious scrams.
D. Turbine Stop Valve Closure Scram Trip Setting The turbine stop valve closure scram anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 110 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the safety limit lMCPR even during the worst case transient that assumes the turbine bypass is closed.
E. Turbine Control Valve Fast Closure Scram Trip Setting The turbine control valve fast closure scram anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection exceeding the capability of the bypass valves. The reactor protection system initiates a scram when fast closure of the control valves is initiated by the acceleration relay. This setting and the fact that control valve closure time is approximately twice as long as that for the stop valves means that resulting transients, while similar, are less severe than for stop valve closure. MCPR remains above the safety limit MCPR.
F. Main Condenser Low Vacuum Scram Trip Setting To protect the main condenser against overpressure, a loss of condenser vacuum initiates automatic closure of the turbine stop valves and turbine bypass valves. To anticipate the transient and automatic' scram resulting from the closure of the turbine stop valves, low condenser vacuum initiates a scram. The low vacuum scram set point is selected to initiate a scram before the closure of the turbine stop valves is initiated.
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Amendment No. la
0 f , ci :
s 4 L2.1. BASES:
qTransient and accident analyses demonstrate that these conditions result in adequate safety margins for the fuel.
References:
- 1. Linford,~R. B..._" Analytical-Methods of Plant Transient Evaluat!ons for the General Electric-Boiling Water Reactor," NEDO-10802, Feb., 6973.
?2. " Pilgrim Nuclear Power Station Single Loop Operation," NED0-24268, June 1980 with Errata-and Addenda Sheet No. 1, September 1980.
I i
p
' Amendment No. 20 (the next page is 22)
TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT
' Minimum Number Modes in Which Function .
Operable Inst. Trip Function Trip' Level Setting Must-Be Operable' Action
Channels per. Refuel (7) Startup Hot Run Trip (l) System Standby 1 Mode Switch in Shutdown X X X A 1 Manual Scram X X X A IRM 3 High Flux 1120/125 of full scale X 'X (5) A 3 . Inoperative .
X X (5) A APRM 1 (.58W + 62 - X) ~FRP' MFLPD 2 High Flux (14) (15) (19) .(17) ( 17) .X A or B 2 Inoperative .
X X(9) X A or B 2 Downscale >2.5 Indicated on Scale (11) (11) X(12) A or B 2 High Flux (15%) 115% of Design Power X X (16) A or B 2 High Reactor Pressure 11085 psig X(10) ,X X A 2 High Drywell Pressure 12.5 psig X(8) X(8) X A 2 Reactor Low Water Level >9 In. Indicated Level X X X A 2 High Water Level in Scram Discharge Tank 139 Gallons X(2) X X A 2 Turbine Condenser Low Vacuum >23 In. Hg Vacuum X(3) X(3) X A or C 2 Main Steam Line High 17X Normal Full Power Radiation Background (18) X X X(18) 'A or C 4 Main Steam Line Isolation Valve Closure 110% Valve Closure X(3)(6) X(3)(6) X(6) A or C 2 Turb. Cont. Valve Fast >150 psig Control Oil Closure Pressure at Acceleration Relay X(4) X(4) X(4) A or D 4 Turbine Stop Valve Closure 1 10% Valve Closure X(4) X(4) X(4) A or D Amendment No. 27
m.-
NOTES FOR TABLE 3.1.1 (Cont'd)
- 10. Not_ required to be operable when the reactor pressure vessel head is not bolted to the vessel.
- 11. The APRM downscale trip function is only active when the reactor mode 1
switch is in run.
- 12. The APRM downscale_ trip is automatically bypassed when the IRM instrumentation is operable and not high.
- 13. An APRM will be considered inoperable if there are less than 2 LPRM inputs per level or there is less than 50% of the normal complement of LPRM's to an APRM.
- 14. W-is percent'of drive flow required to produce.a rated core flow of 69 Mlb/hr. Trip level setting'in percent of design power (1998 MWt).
- 15. See Section 2.1.A.l.
The APRM flow biased high flux' scram is bypassed when in the refuel or
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-17.
startup/ hot standby modes.
-18. 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 excected during the test. The background radiation level and associated trip setpoints may be adjusted during the test based on either calculations 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.
- 19. X - O for two loop operation and single loop operation less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
. X = 2.9 for single loop operation greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
e Amendment No. 29
z3 9 4 PNPS-TABLE 3.2.C INSTRUMENTATION THAT INITIATES ROD BLOCKS Minimum # of Operable Instrument
' Channels Per Trip Systems (1) Instrument Trio Level Setting 2 APRM Upscale (Flow (0.58 W + 50% - X) ~ FRP' (2)
Blased) MFLPD 2 APRM Downscale 2.5 indicated on scale.
1 (7) Rod Block Monitor (0.65 W + 42% - 1.12X) ' FRP' (2) (10)
(Flow Blased) ,
MFLPD 1 (7) Rod Block Monitor 5/125 of full scale Downscale 3 IRM Downscale (3) 5/125 of full scale 3 IRM Detector not in (8)
Startup Position 3 IRM Upscale 11 08/125 of full scale 2 (5) SRM Detector not in (4)
Startup Position 5
2 (5) (6) SRM Upscale 10 1 counts /sec.
1 (per tank) (9) Scram Discharge Volume 118 gallons Water Level-High Amendment No. 54
NOTES FOR TABLE 3.2.C
- 1. For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function.
The SRM and IRM blocks need not be operable in "Run" mode, and the APRM and RBM rod blocks need not be operable in "Startup" mode. If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter. If this condition lasts longer than seven days, the system shall be tripped. If the first column cannot be met for both trip systems, the systems shall be tripped.
- 2. H is percent of drive flow required to produce a rated core flow of 69 Mlb/hr. Trip level setting is in percent of design power (1998 MWt).
For' flows of 100% or greater, the rod block monitor maximum trip level setting shall be 107% power.
X - 0 for two recirculation loop operation and for single loop operation less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
X = 2.9 for. single loop operation greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 3. IRM downscale is bypassed when it is on its lowest range.
- 4. This function is bypassed when the count rate is > 100 cps.
- 5. One of the four SRM inputs may be bypassed.
- 7. The trip is bypassed when the reactor power is 1 30%.
- 8. This function is bytassed when the mode switch is placed in Run.
- 9. If the number of nnt.rable channels is less than required by the minimum number of operable instrument channels per trip system requirement, place the inoperable channel in the tripped condition within one hour.
- 10. A factor of 1.12 (ratio of the slope of the R8M flow bias to that of the APRM flow blas) adjusts X for the slope difference.
Amendment No. 55
e LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.D JSafety Relief Valves (Con't) 4.6.E. Jet Pumps from the initial discovery of Whenever there is recirculation discharge pipe temperatures in flow with the reactor in the excess of 212*F for more than 24 startup or run mode, jet pump hours without prior NRC approval operability shall be checked dally of the engineering evaluation by verifying that the following delineated in 3.6.D.3. conditions do not occur simultaneously.
- 5. The limiting conditions of operation for the 1. The two recirculation loops have a instrumentation that monitors flow imbalance of 15% or more when tail pipe temperature are given the pumps are operated at the same in Table 3.2.F. speed.
E. Jet Pumps 2. The indicated value of core flow rate varies from the value derived
- 1. Whenever the reactor is in the from loop flow measurements by startup or run modes, all jet more than 10%.
pumps shall be operable. If it is determined that a jet pump is 3. The diffuser to lower plenum inoperable, an orderly shutdown differential pressure reading on shall be initiated and the an individual jet pump varies from reactor shall be in a Cold established jet pump AP Shutdown Condition within 24 characteristics by more than 10%.
hours.
F. Recirculation Pumps F. Recirculation Pumps
- 1. Whenever both recirculation Recirculation pump speeds shall be pumps are in operation, pump checked and logged at least once speeds shall be maintained per day.
within 10% of each other when power level is greater than 80%
and within 15% of each other when power level is less than or equal to 80%.
- 2. If Specification 3.6.F.1 is exceeded immediate corrective action shall be taken. If recirculation pump speed mismatch is not corrected within 30 minutes, declare the recirculation loop having the pump with the slower speed inoperable and initiate an orderly process to reduce power to 50% or less and to secure the inoperable pump.
Amendment No. 126a i
C:
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.F Recirculation Pumps (Cont'd) 4.6.F Recirculation Pumps (Cont'd)
- 3. Single' loop reactor operation is not permitted for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the reactor is at less than or equal to 50% of rated thermal power and the following designated adjustments are made for APRM rod block and scram setpoints (Tech. Spec.
2.1.A and B), RBM setpoint (Table 3.2.C), MCPR fuel
-cladding integrity safety limit and operating limits (Tech.
Spec. 1.1.A and 3.II.C respectively), and MAPLHGR
-(Tech. Spec. 3.ll.A).
G. Structural Integrity G. Structural Integrity
- 1. The structural integrity of the The nondestructive inspections primary system boundary shall be listed in Table 4.6.1 shall be maintained at the level required performed as specified. The by the ASHE Boller and Pressure results obtained from compliance Vessel Code,Section XI, " Rules with this specification will be of Inservice Inspection of evaluated after 5 years and the Nuclear Power Plant Components", conclusions of this evaluation 1974. will be reviewed with AEC.
Amendment No. 127 I
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11 REACTOR FUEL ASSEMBLY 4.11 REACTOR FUEL ASSEMBLY Appilcability: Applicabillity:
The Limiting Conditions for The surveillance requirements Operation associated with fuel apply to the parameters which rods apply to those parameters the fuel rod operating which monitor the fuel rod conditions.
operating conditions.
Objective: Objective:
The Objective of the Limiting The Objective of the Conditions for Operation is to Surveillance Requirements is to assure the performance of the specify the type and frequency fuel rods. of surveillance to be applied to the fuel rods.
Specifications: Specifications:
A. Average Planar Linear Heat A. Average Planar Linear Heat Generation Rate (APLHGR) Generation Rate (APLHGR)
During power operation with both The APLHGR for each type of recirculation pumps operating, fuel as a function of average the APLHGR for each type of fuel planar exposure shall be as a function of average planar determined daily during reactor exposure shall not exceed the operation at >25% rated thermal applicable limiting value shown power.
in-Figures 3.11-1 through 3.11-6. The top curves are appilcable for core flow-greater than or equal to 90% of rated core flow. When core flow is less than 90% of rated core flow, the lower curves shall be limiting. For greater than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> operation with one recirculation pump out of service, values from these lower curves are to be multiplied by 0.86 for 8x8 and 8x8R fuel.
(see Table 3.11-1, Reference pages 205El-6) If at any time during operation it is determined by normal survelliance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the Amendment No. 205a
r
- O
~
LIMITINGCONDITIONSFOROPEkATION SURVEILLANCE REQUIREMENTS
, j .: '.
3.II'.A. Average Plandr Linear Heat Generation Rate (APLHGR)
(Cont'd) prescribed limits. If the APLHGR is not returned to within the prescribed limits 1.,
within two (2). hours, the reactor shall be brought to Lthe Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
- Surveillances and corresponding action'shalI continue until reactor operation is within the prescribed limits.
'B. -Linear Heat Generation Rate 'B. Linear Heat Generation Rate (LHGR) , (LHGR)
During reactor power operation the 1inear.' heat generation The LhGR as a function of core rate (LHGR) of any rod in any height shall be checked daily fuel assembly at any axial curing reactor operation at location shall not exceed 13.4 >257, rated thermal power.
kw/ft for 8x8 and P8x8R fuel.
If at any time during operation it is determined by normal surveillance that'the ilmiting value for LHGR is being exceeded, action saall 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.
Amendment No. 205a-1
m LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 6 C. Minimum Critical Power Ratio (MCPR) C. Minimum Critical Power Rat'q_(MCPR)
- 1. During power operation MCPR shall 1. MCPR shall be determined daily be > the MCPR operating limit during reactor power operation specified in 3.11.C.2. If at any at >25% rated thermal power and
' time during operation it is following any change in oower determined by normal surveillance level or distribution that that the limiting value for MCPR would cause operation with a is'being exceeded, action shall limiting control rod pattern as be initiated within 15 minutes to described in the bases for restore operation to within the Specification 3.3.B.S.
prescribed limits. If the steady state MCPR is not returned to 2. The value of T in Soecification within the prescribed limits 3.11'.C.2 shall be eaual to 1.0 within two (2) hours, the reactor unless determined from the shall be brought to the Cold result of surveillance testing Shutdown condition within 36 of Specification 4.3.C as hours. Surveillance and follows:
corresponding action shall continue untti reactor operation a) T is defined as:
is within the prescribed limits.
T r- ave 76 For core flows other than rated the 1.275 78 MCPR limits shall be the limits identified above times K, where K, b) The average scram time to is as shown in Figure 3.11-8. the 30% insertion position is determined as follows:
As an alternative method providing equivalent thermal-hydraulic n protection at core flows other than E N. T, T
rated, the calculated MCPR may be ave 1 divided by Kr.where K, is shown in n Figure 3.11-8. E N.
1-1
- 2. a. The operating limit MCPR values for two recirculation where: n - number of
" loop operation as a function surveillance tests of T are given in fable performed to date in the 3.11-1 where -f is given in cycle.
specification 4.11.C.2. y
- b. For one recirculation loop operation of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the MCPR limits at rated flow are 0.01 higher than comparable two-loop operating limits. (See 3.11.C.2.a and Table 3.11-1)
Amendment No. 205b
g TABLE 3.11-1
- 0PERATING LIMIT MCPR VALUES A. ' MCPR Operating Limit from Beginning of Cycle (B0C) to BOC + 6000 MWD /T.
For all values of T 8x8 P8x8R/BP8x8R Two Loop Operation 1.36 .1,40 For all values of T 8x8 P8x8R/BP8x8R Single Loop Operation 1.37 1.41
Two Loop Operation Single Loop Operation T 8x8 P8x8R/8P8x8R 8x8 P8x8R/BP348R T10 1.38 1.40 1.39 1.41 0.0 < T 10.1 1.39 1.41 1.40 1.42 0.1 < r< 0.2 1.39 1.41 1.40 1.42
-0.2 < 71 0.3 1.40 1.42 1.41 1.43 0.3 < r i 0.4 1.40 1.42 1.41 1.43 0.4 < ri 0.5 1.41 1.43 1.42 1.44
- 0.5 < T 10.6 1.41 1.43 1.42 1.44 0.6 <r 1 0.7 1.42 1.44 1.43 1.45 0.7 < r i 0.8 1.42 1.44 1.43 1.45 0.8'< r i 0.9 1.43 1.45 1.44 1.46
~ 0.9 < r i 1.0- 1.43 1.45 1.44 1.46 Amendment No. 205b-2
BASES:
4 3.11.A Average Planar Linear Heat Generation Rate (APLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K.
The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent, secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than thc design LHGR. This LHGR times 1.02 is used in the heat-up code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factors. The limiting value
' for APLHGR is this LHGR of the highest powered rod divided by its local peaking factor.
The calculational procedure used to establish the APLHGR limit for each fuel type is based on a loss-of-coolant accident analysis. The emergency core cooling system (ECCS) evaluation models which are employed to determine the effects of the loss of coolant accident (LOCA) in accordance with 10CFR50 and Appendix K are discussed in Reference 1. The models are identified as LAMB, SCAT, SAFE, REFLOOD, and CHASTE. The LAMB Code calculates the short term blowdown response and core flow, which are input into the SCAT code to calculate blowdown heat transfer coefficients. The SAFE code is used to
' determine longer term system response and flows from the various ECC systems. Where appropriate, the output of SAFE is used in the REFLOOD code to calculate liquid levels. The results of these codes are used in the CHASTE code to calculate fuel clad temperatures and maximum average planar linear heat generation rates (MAPLHGR) for each fuel type.
The significant plant input parameters and the MAPLHGR's for the present fuel types calculated by the above procedure are included in Reference 2.
Reduction factors for one recirculation loop operation were derived from Reference 3.
REFERENCES
- 1. General Electric GWR Generic Reload Fuel Application, NEDE-24011-P.
- 2. Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NED0-21696, August 1977 as amended.
- 3. " Pilgrim Nuclear Power Station Single Loop Operation," NED0-24268, June 1980 with Errata and Addenda Sheet No. 1, September 1980.
Amendment No. 205c (next page is 205-3)
TL
-9 BASES:
3.11.C (Cont'd)
The analysis of the rod withdrawal error for Pilgrim Unit I considers the continuous withdrawal of the maximum worth control rod at its maximum drive speed _from the reactor. A summary of the analytical methods used to determine the nuclear characteristics is given in Section 5.2.1.5 of NEDE-24011-P.
For single loop operation, the operational MCPR limit is increased oy 0.01 to account for the increase in the fuel cladding integrity safety limit (Specification 1.1.A).
MCPR LIMITS FOR CORE FLOWS OTHER THAN RATED The-purpose of the K, factor is to define operating limits at other than rated flow conditions. At less than 100% flow the required MCPR is the product of the operating limit MCPR and the K, factor. Specifically, the K, factor provides the required thermal margin to protect against a flow increase transient. The most limiting transient initiated from less tnan rated flow conditions is the recirculation pump speed up caused by a motor-generator speed control failure.
For operation in the automatic flow control mode, the K, factors assure that the operating limit MCPR given in Specification 3.11C will not be violated should the most limiting transient occur at less than rated flow. In the manual flow control mode, the K, factors assure that the Safety Limit MCPR will not be violated for the same postulated transient event.
The K, factor curves shown in Figure 3.11-8(*' were developed generically which are applicable to all BWR 2, BWR 3, and BWR 4 reactors. The K, factors were derived using the flow control line corresponding to rated thermal power at rated core flow.
For the manual flow control mode, the K, factors were calculated such that at the maximum flow state (as limited by the pump scoop tube set point) and the corresponding core power (along the rated flow control line), the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPR's were calculated at
-different points along the rated flow control line corresponding to different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR determines the K,.
'For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at rated power and flow.
Amendment No. 205c-5
M APLHGR Versus Planar Average Exposure Fuel Type 8DB219L 13 +- Core Flow > or = 902 rated l
I i
II 11.9 12.1
_4- ~[N4 12.1 + Core Flow < 902 rated D' D
- 33,4 - 11,3
- 1 11.7
\ \
Maximum io.e 9 i g,\ 4 2
Av-== .-y~.
20 y ~qe i ar 9.8 9.8
,,, a 9 , g,
\
9.2 9
@ ation 9.1 i
Rate l (kw/ft) 8 -
%0 'N e,3 N 8.6 i
1 I 7.8 i
II
! 7 7.3 0 5000 10000 15000 20000 25000 30000 35000 40000 Planar Average Exposure (MW d/t)
Amendment NO. Figure 3.11-1 Page 205E-1
MAPLHGR Versus Planar Average Exposure Fuel Type 8DB219H 13 -*- Core Flow > or = 90X rated l
12 2 G3 12.1
+ Core Flow < 902 rated 11.8 / -4
' m 4' o_ -" single loop > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
" 3,,3* 3
.6 11 7 N
~
11.2 Maximum i)g/ Y -G,2 w '" io,7
^yr'8' y p:j '"i~,N d Q~ 9.,6
- ,r Generation
,.,., 1 9.1
,2 9.'2
.s JNNef 1 9,3 Rate 8.6
, i[
(kw/ft) a v 3,3 x ,,
W*** y,'e ,
2M I I 7.3 0 5000 10000 15000 20000 25000 30000 35000 40000 Planar Average Exposure (MW d/t)
Anendment NO. Figure 3.11-2 Page 205E-2
O 1 MAPLHGR Versus Planar Average Exposure Fuel Type 8DB262 I3 + Core Flow > or - 905 rated l
I l 11.9 [ -
i + Core Flow < 905 rated 12 .- 11.6 1 n- 4 + Single loop > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> II
.,/
/[11.3 11 116 ;j l
N ,10./
Maximum 11 0 F l 10.6 10.7 6s Average in y
['m"*"
Heat 9 - d 9.2
,1 ei el 1N. g' N
NiN4 l\ i Generation 9.1 "
9.3 Nf ;
Rate 8.7 a,7
. (kw/ft) a Y ,
V1000 7.g ,
i 7 7.4 0 5000 10000 15000 20000 25000 30000 35000 40000 Planar Average Exposure (MW d/t)
Amendment No. Figure 3.11-3 Page 205E-3 i
c .
MAPLHGR Versus Planar Average Exposure Fuel Types P8DRB265L BPDRB265L
'* Core Flow > or - 90X rated 12.1 12.1 12.1 11.9 .o- Core Flow < 90R rated 12 - 11.6 -4
+-+ 0 ..
oN 1.3 u- Single loop > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 11.6 -
v 4 15 11.5 11.5 I 11 jk >
11.3 K 10.7 s i.o 11.0 Maximum l 1 .2 10.7 10 m :: E-a 10.2 d in ar 98 g,a f
B-l' 9.8 9,7 9,7 Heat l ,N 9 9 g-9A 9.z N g'i Generation ,
n Rate a.7 yN (kw/ft) a y e.3 N a V 7.8 200 o 5000 10000 15000 2o000 25000 30000 35000 40000 Planar Average Exposure (MW d/t)
Anendment No. Figure 3.11-4 Page 205E-4
MAPLHGR Versus Planar Average Exposure Fuel Types P8DRB282,BP8DRB282 13 *' Core Flow > or - 902 rated 12.1 + Core Flow < 902 rated 12
^ 11 0 b d E- Single loop > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> I j j c; N d11.1 Maximum
( ,[ 1 l'.2 i
11.2 l
u- j 10.4 Average 10.6 10.6 10.7 l
10.6 , \ 9.8 Planar 3o
' l **
Linear gh IX, 99 f
Heat 9
"- "fl 9.6 9.s 9.6 y 9.2
--it U
l Generation 9.1 9.1 g,j Rate i (kW/ft) 8 y 8~[ si i 1000 1 7.9 I 7 0 5000 10000 15000 20000 25000 30000 35000 40000 Planar Average Exposure (MW d/t)
Amendment No. Figure 3.11-5 Page 205E-5
- m. .,
MAPLHGR Versus Planar Average Exposure Fuel Types P8DRB265H.BPSDRB265H 13 .*- Core Flow > or - 902 rated 12.1 12.1 + Core Flow < 90X rated 11.9 e +% 11.9 I .6
>j, .m. Single loop > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 11.s <
j p ,,
II ~Y 11,3 11.5 11.5 3 y ,3 % 10,7 "0
Maximum io.g I 10.2 Average io 10 7 . W Planar :
-s 10.2 N N9.6 Linear a l I I
,7 9.8 9.8 9,7 g,7
, N, Heat 9 J -- 9 -
Rate Generation 9.3 gh N',' I 9.1 8.7 y (kw/ft) a "
8.3 N a 1000 1 3 y 200 7 )
o 5000 10000 15000 20000 25000 30000 35000 4o000 Planar Average Exposure (MW d/t)
Anendment No. Figure 3.11-6 Page 205E-6