ML20237J854
| ML20237J854 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/17/1987 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20237J824 | List: |
| References | |
| B12413, NUDOCS 8708260341 | |
| Download: ML20237J854 (36) | |
Text
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h' Docket No. 50-745 B12413 Attachment 1 Millstone Nuclear Power Station, Unit No.1 Proposed Revision to Technical Specifications Single Loop Operation l
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8708260341 870817 PDR ADOCK 05000245
'r PDR I
August 1987 L
.6 SAFETY LIMITS 2.1.1 FUEL CLADDING INTEGRITY Applicability:
Applies to the interrelated variables associated with fuel thermal behavior.
l Objective:
To establish limits below which the integrity of the fuel cladding is preserved.
Specification:
.. A. When the reactor pressure is greater than 800 psia and the core flow is greater than 10% of rated design, a minimum critical power ratio (MGPR) lens than 1.07 for two loop operation or 1.08 for single loop operation, shall constitute a violation of the fuel cladding integrity safety limit. l LIMITING SAFETY SYSTEM SETTINGS I 2.1.2 FUEL CLADDING INTEGRITY Applicability:
Applies to trip settings of the instruments and devices which are provided to prevent the reactor system safety limits from being exceeded. >
Objective:
To define the level of the process variables at which automatic protective action is initiated to prevent the safety limits from being exceeded.
Specification:
The limiting safety system settings shall be as specified below:
A. Neutron Flux Scram
- a. When the Mode Switch is in the RUN position, the APRM flux scram trip settir.g shall be as shown on Figure 2.1.2 and shall be:
S $ 0.58 W + 62 (2 loop operation)
S $ 0.58 W + 59 (1 loop operation)
Millstone Unit 1 2-1 1
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SAFETY LIMITS 2.1.1 FUEL CLADDING INTEGRITY B. When the reactor pressure is less than or equal to 800 psia or reactor flow is less than 10% of desi En, the reactor thermal power transferred to the coolant shall not exceed 25% of rated.
C. 1. To assure that the Limiting Safety System Settings established in Specifications 2.1.2A and 2.1.2B are not exceeded, each required scram shall be initiated by its primary source signal. The Safety Limit shall be assumed to be exceeded when scram is accomplished by a means other than the Primary Source Signal.
- 2. When the process computer is out of service, this safety limit shall be assumed to be exceeded if the neutron flux exceeds the scram setting established by Specification 2.1.2A and a control rod scram does not occur.
D. Whenever the reactor is in the cold shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be less than that corresponding to 12 inches above the top of the active fuel when it is seated in the core. This level shall be continuously monitored.
LIMITING SAFETY SYSTEM SETTINGS l
2.1.2.A.1.a. where:
S = Setting in percent of rated thermal power (2011 MWt) l W= Total active recirculation drive (pump) flow in percent of design.
See Note (1)
The trip setting shall not exceed 90 percent of rated power during generator load rejections from an initial generator power greater than 307 MWe. The APRM scram setdown shall be 90% of rated within 30 seconds after initiation of full load rejection.
Note (1) Design flow to be de{ined as the recirculation drive (pump) flow, not to exceed 33.48 x 10 lbs/hr. (86,000 gpm), needed to achieve 100% core flow.
Millstone Unit 1 2-3 L.
LIMITING SAFETY SYSTEM SETTINGS (Continued)
- b. In the event of operation with a maximum ~ L* action of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
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S f (0.58 W + 62) FRP_ 2 loop operation
_MFLPD_
S f (0.58 W + 59) - FRP ~ 1 loop operation MFLPD where, FRP = fraction of rated thermal power (2011 MWt)
Millstone Unit 1 2-3a
h LIMITING SAFETY SYSTEM SETTINGS (Continued) 2.1.2 FUEL CLADDING INTEGRITY B. 1. APRM Rod Block Trip Setting
- a. The APRM rod block trip setting shall be as shown in Figure 2.1.2 and shall be: (Run Mode)
RB g 0.58W + 50 S 2 loop operation SRB $ 0.58W + 47 1 loop operation where:
S RB Rod block setting in percent of rated thermal power (2011 MWt).
W = Total' recirculation flow in percent of design. (Note 1, Page 2-3).
- b. In the event of operation with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP),
the setting shall be modified as follows:
~
S RB g (0.58W + 50) FRP 2 loop operation MFLPD S
where:
FRP = fraction of rated thermal power (2011 MWt) l maximum fraction of' limiting 1
MFLPD =
power density where the limiting power density is 13.4 kW/ft for BP8x8R (GE-7B) fuel bundles and 14.4 kW/ft for GE8x8EB (GE-8B) fuel bundles. l l 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.
- c. During power ascensions with power levels less than or equal to 90%, APRM Rod Block Trip Setting adjustments may be made as described below, provided that the change in scram setting adjustment is less than 10% and a notice of the adjustment is posted on the reactor control panel:
Millstone Unit 1 2-5 c - -___-_-__ _
LIMITING SAFETY SYSTEM SETTINGS (Continued) 2.1.2 FUEL CLADDING INTEGRITY B.1.c. The APRM meter indication is adjusted by:
APRM =
f P where:
APRM = APRM Meter Indication P = % Core Thermal Power
- 2. The APRM rod block trip setting for the refuel and start-up/ hot standby mode shall be less than or equal to 12% rated thermal power.
C. The Reactor Low Water Level Scram trip setting shall be greater than or equal to 127 inches above the top of the active fuel.
D. The Reactor Low Low Water Level ECCS initiation trip point shall not be greater than 83 inches nor less than 79 inches.
E. The Turbine Stop Valve Scram trip setting shall be less than or equal to ten percent valve closure from full open.
F. The Turbine Control Valve Fast Closure Scram shall trip upon actuation of the acceleration relay in conjunction with failure of selected bypass valves to start opening within 280 milliseconds.
The maximum setting of the time delay relays which bypass this scram shall be 280 milliseconds.
G. The Main Steam Isolation Valve Closure Scram trip settings shall be less than or equal to ten percent valve closure from full open.
H. The Main Steam Line Low Pressure trip, which initiates main steam line isolation valve closure, shall be greater than or equal to 825 psig.
Millstone Unit 1 2-6
s 2.1.1 FUEL CLADDING INTEGRITY BASES The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the minimum critical power ratio (MCPR) is no less than 1.07 for two loop operation or 1.08 for single loop operation.
MCPR equal to or greater than 1.07 for two loop operation or 1.08 for single loop operation represents a conservative margin relative to the conditions
. required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission producc migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from
-reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with margin to the conditions which would produce onset of transition boiling (MCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation.
Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure. However, the existence of critical power, or boiling transition, is not a directly observable parameter in an operating reactor. Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.
The margin for each fuel assembly is characterized by the critical power ratio (CPR) which is the ratio of the bundle power which would produce onset of transition boiling to the actual bundle power. The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR). It is assumed that plant operation is controlled to the nominal protective setpoints via the instrumented variable, i.e., normal plant operation presented on Figure 2.1.2 by the nominal expected flow control line. The Safety Limit (MCPR of 107 for two loop operation or 1.08 for single loop operation) has sufficient conservatism to assure that in the event of an abnormal operational transient, initiated from a normal operating condition, more than 99.9% of the fuel rods in the core are expected to avoid boiling transition. The margin j between MCPR of 1.0 (onset of transition boiling) and the safety limit for 4 two recirculation loop operation (MCPR = 1.07) is derived from a detailed statistical analysis considering all of the uncertainties in monitoring the
- i. core operating state including uncertainty in the boiling transition L
Millstone Unit 1 B 2-1 L
i 2.1.1 . FUEL CLADDING INTEGRITY (continued)
BASES correlation as described.in Reference 1. The uncertainties employed in deriving the safety limit are provided at the beginning of each fuel cycle.
Adjustment to the safety limit (MCPR = 1.08) for single recirculation loop operation'was derived in Reference 2.
- 1. NED0-10958, General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application.
- 2. Millstone Point Nuclear Power Station, Unit 1 Single-Loop Operation, NEDO-211312, December 1980.
Millstone Unit 1 B 2-la
1 2.1.1 FUEL CLADDING INTEGRITY BASES Because the boiling transition correlation is based on a large quantity of full scale data, there is a very high confidence that operation of a fuel assembly at the condition of MCPR = 1.07 for two recirculation loop operation or 1.08 for single loop operation would not produce boiling transition. l 1
However, if boiling transition were to occur, clad perforation would not be expected. Cladding temperatures would increase to approximately 11000F l which is below the perforation temperature of the cladding material. This has been verified by tests in the General Electric Test Reactor (CETR) where fuel similar in design to Millstone operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation. Thus, although it is not required to establish the safety limit, additional margin exists between the safety limit and the actual occurrence of loss of cladding integrity. The limit of applicability of the boiling transition correlation is 1400 psia during normal power operation. However, the reactor pressure is limited as per Specification 2.2.1.
The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of maximum fraction of limiting power density and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1.2. A.1, when the maximum fraction of limiting power density is greater than the fraction of rated power.
If the APRM scram setting should require a change due to an abnormal peaking condition, it will be done by increasing the APRM gain thus reducing the slope and intercept point of the flow referenced scram curve by the reciprocal of the APRM gain change.
At pressures below 800 psia, the core elevation pressure drop (0 power, l 0 flow) is greater than 4.56 psi. At low power and all flows this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressuredropatlowpowerandallflowsgillalwaysbegreaterthan4.56 psi.
Analyses show that with a flow of 28 x 10 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3 5 psi. Tgus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 10 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 indicaee that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a core thermal power of more than 50%. Thus, a core thermal power limit of 25% for reactor pressures below 800 psia or core flow less than 10% is conservative.
Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 2.1.1 A or 2.1.1B will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a Millstone Unit 1 B 2-2
4 2.1.1 FUEL CLADDING INTEGRITY BASES scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure 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.
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 level shall be continuously monitored.
'The computer provided with Millstone Unit No. 1 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. Thus, computer information normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 2.1.1C.2 will be relied on to determine if a safety limit has t.en violated.
Millstone Unit 1 B 2-3
2.1.2 FUEL CLADDING INTEGRITY I BASES The transients expected during operation of Millstone Unit 1 have been analyzed up to the thermal power condition of 2011 MWt. The analyses were based upon plant operation in accordance with the operating map given in Figure 3 3 1 for two loop operation and Figure 3 3 2 for single loop operation.
In addition, 2011 MWt is the licensed maximum steady-state power level of i This maximum steady-state power will never be knowingly Millstone Unit 1.
exceeded.
Conservatism was incorporated by conservatively estimating the controlling i factors such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, axial power shapes, etc. These factors are all 3 elected conservatively with respect to their effect on the applicable l transient results as determined by the current analysis model. This transient '
model, evolved over many years, has been substantiated in operation as a.
conservative tool for the evaluation of reactor dynamics performance.
Comparisons have been made showing results obtained from a General Electric boiling water reactor and the predictions made by the model. The comparisons and results are summarized in Reference 3.
The void reactivity coefficient utilized in the analysis is conservatively estimated to be about 25% larger that. the most negative value expected to occur during the core lifetime. The scram worth used has been derated to be equivalent to the scram worth of about 80% of the control rods. The scram delay time and rate of rod insertion are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifications. The effect of scram worth, scram delay time and rod. insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative l reactivity strongly turns the transient and the stated 5% and 20% insertion '
times conservatively accomplished this desired initial effect. The time for 50% and 90% insertion are given to assure proper completion of the insertion stroke, to further assure the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state ,
condition. I For analyses of the thermal consequences of the transients, MCPRs specified in Section 3 11.C are conservatively assumed to exist prior to initiation of the transients.
(3) Linford, R.B., " Analytical Methods of Plant Transient Evaluations for the l
General Electric Boiling Water Reactor," NED0-10802. '
(4) " Extended Load Line Limit Analysis, Millstone Point Nuclear Power Station, Unit 1", NED0-24366 and NEDO 24366-1. (See TS Page B2-5.)
Millstone Unit 1 B 2-4 l
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2.1.2 FUEL CLADDING INTEGRITY BASES This choice of using conservative values of controlling parameters and initiating transients at the licensed maximum steady state power level produces more pessimistic answers than would result by using 9xpected values of control parameters and analyzing at higher power leve13.
In summary:
1 - The licensed maximum steady-state power level is 2011 MWt.
2 - The abnormal operational transients were analyzed to a thermal power level of 2011 MWt.
3 - Analyses of transients employ adequately conservative values of the controlling reactor parameters.
4 - 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.
A. Neutron Flux Scram The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads percent of rated thermal power. _ Since 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 instantaneous neutron flux due to the time constant of the fuel. Therefore, during transients induced by disturbances.and with APRM scram settings as specified by Section 2.1.2A, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrate that, even with a fixed 120% scram trip setting, none of the postulated transients result in violation of the fuel safety limit and there is a substantial margin from
{
fuel damage. Therefore, use of a flow-biased scram provides even '
additional margin.
Increasing the APRM scram setting would decrease the margin present before the thermal hydraulic safety limit is reached. The APRM scram setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. A reduction in this operating margin would increase the frequency of spurious scrams which have an j adverse affect on reactor safety. Thus, the specified APRM setting was .
selected to provide adequate margin from the thermal-hydraulic safety S limit and allow operating margin to minimize the frequency of unnecessary scrams.
1 Millstone Unit 1 B 2-5 j i
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I 1
2.1.2 FUEL CLADDING INTEGRITY j BASES Analyses of the limiting transients show that no scram adjustment is required to assure MCPR. 251.07 for two recirculation loop operation or 1.08 for single loop operation when the transient is initiated from MCPR's specified in Section 3 11.c. In order to assure adequate core margin during full load rejections in the event of failure of the select rod insert, _it is necessary to reduce the APRM scram trip setting to 90% of rated power following a full load rejection incident. This is necessary because, in the event of failure of the select rod insert to function, the cold feedwater would slowly increase the reactor power level to the scram trip setpoint. A trip setpoint of 90% of rated has been established to provide substantial margin during such an occurrence. The trip setdown is delayed to prevent scram during the ini tial portion of the transient. The specified maximum setdown delay of 30 seconds is conservative because the cold feedwater transient does not produce significant increases in reactor power before approximately 60 seconds following the load rejection.
For operation in the REFUEL or STARTUP/ HOT STANDBY modes while the reactor is at low pressure,_the APRM reduced flux trip scram. setting of 5'15% of i rated power provides adequate thermal margin between the maximum power and the safety limit, 25% of rated power. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content are minor.
Cold water from sources available during startup is not much colder than that already in the syst;m. Temperature coefficients are small and control _ rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.
In an assumed uniform rod withdrawal approach to the scram level, the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The APRM reduced trip scram remains active until the mode switch is placed in the RUN position. This switch occurs when the reactor pressure is greater than 880 psig.
The IRM trip at 6120/125 of full scale remains as a backup feature.
The analysis to support operation at various power and flow relationships has considereu operation with either one or two recirculation pumps.
During steady-state operation, with one recirculation pump operating, the equalizer line shall be isolated. Analyses of transients from this operating condition are given in Reference 2.
l Millstone Unit 1 B 2-6
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2,1.2 FUEL CLADDING INTEGRITY BASES l
l B. APRM Control Rod Block Trips Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to i prevent rod withdrawal beyond a given point at constant recirculation flow l rate and thus to protect against a condition of MCPR ' 41.07 for two loop operation or 1.08 for single loop operation. This rod block setpoint, which is automatically varied with recirculation flow rate, prevents an increase in the reactor power level to excessive valuns due'to control rod withdrawal. The specified flow variable setpoint provides substantial margin from fuel damage, assuming steady-state operation at.the setpoint, over the entire recirculation flow range. The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow re]ationship. Therefore, the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip 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.
When the maximum fraction of limiting power density exceeds the fraction of rated thermal reactor power, the rod block setting is adjusted in accordance with the. formula in Specification 2.1.2.B. If the APRM rod block setting should require a change due to an abnormal peaking condition, it will be done by increasing the APRM gain, thus reducing the slope and intercept point of the flow referenced rod block curve by the reciprocal-of the APRM gain change.
The relationship between recirculation drive flow and core flow is slightly different for single loop vs. two loop operation. This difference, which is described and quantified in Reference 5, changes the intercept point of the flow referenced APRM rod block and scram curves.
The APRM rod block setpoint is reduced to $i12% of rated thermal power with the mode switch in REFUEL or STARTUP/ HOT STANDBY position.
C. Reactor Low Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the safety limit is maintained.
(5) " Millstone Nuclear Power Station, Unit 1 License Amendment Submittal for Single Loop Operation," NEDO-21119, May 1976.
Millstone Unit i B 2-7
2.1.2 FUEL CLADDING INTEGRITY BASES D. Reactor Low Low Water Level ECCS Initiation Trip Point The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the decay heat associated with the loss-of-coolant accident; to limit fuel clad temperature to well below the clad melting temperature; to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%. To accomplish this function, the capacity of each emergency core cooling system component was established based on the reactor low low water level. To lower the setpoint of the low water level scram would require an increase in the capacity of each of the ECCS components. Thus, the reactor vessel low l
water level scram was set low enough to permit margin for operation, yet will not be set lower because of ECCS capacity requirements.
The design of the ECCS components to meet the above criteria was dependent on three previously set parameters: the maximum break size, the low water level scram setpoint, and the ECCS initiation setpoint. To lower the setpoint for initiation of the ECCS would prevent the ECCS components from meeting their design criteria. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.
E. Turbine Stop Valve Scram
.The turbine stop valve scram, like the load rejection scram, anticipates the pressure, neutron flux, and heat flux increase caused by the rapid closure of the turbine stop valves and failure of the bypass. With a scram setting of i 10% of valve closure, the resultant increase in I surface heat flux is limited such that MCPR remains above 1.07 for two recirculation loop operation or 1.08 for single loop operation even during the worst case transient that assumes the turbine bypass is closed.
This scram is bypassed when turbine steam flow is < 45% of rated, as measured by the turbine first stage pressure.
F. Turbine Control Valve Fast Closure The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass; i.e., it prevents MCPR from becoming less than 1.07 for two recirculation loop operation or 1.08 for single loop operation for this transient. For the load rejection from 100% power, the heat flux increases to only 106.5% of its rated power value, which results in only a l
small decrease in MCPR. This trip is bypassed below a generator output of i 307 MWe because, below this power level, the MCPR is greater than 1.07 for i two recirculation loop operation or 1.08 for single loop operation j throughout the transient without the scram. '
Millstone Unit 1 B 2-8
2.1.2 FUEL CLADDING INTEGRITY BASES In order to accommodate the full load rejection capability, this scram trip must be bypassed because it would be actuated and would scram the reactor during load rejections. This trip is automatically bypassed for a maximum of 280 millisec following initiation of load rejection. After 280 millisec, the trip is bypassed providing the bypass valves have opened. If the bypass valves have not opened after 280 millisec, the bypass is' removed and the trip is returned to the active condition. This bypass does not adversely affect plant safety because the primary system pressure is within limits during the worst transient even if the trip
-fails. There are many other trip functions which protect the system during such transients.
G. ' Main Steam Line Isolation Valve Closure Scram The low pressure isolation of the main steam lines at 825 psig was provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the 1 STARTUP position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients, which occur during normal or inadvertent isolation valve closure. With the scrams set at 10% valve closure, there is no increase in neutron flux during valve closure.
H. Main Steam Line Low Pressure Initiates Main Steam Isolution Valve Closure The low pressure isolation at 825 psig was provided to give protection against fast reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed to provide for reactor shutdown so that operation at pressures lower than those specified in the thermal-hydraulic safety limit does not occur, although operation at a pressure lower than 825 psis would not necessarily constitute an unsafe condition.
Millstone Unit 1 B 2-9
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b' LIMITING CONDITION FOR OPERATION
'33 REACTIVITY CONTROL E. Reactivity Anomalies The reactivity equivalent of the difference between the actual critical rod configuration and the expected configuration during power operation shall not exceed 15 21 K. If this limit is exceeded, the reactor will be shutdown until the cause has been determined and corrective actions have been taken if such actions are appropriate.
F. Shutdown Requirements If Specification 3.3 A through D above are not met, a normal orderly shutdown shall be initiated and the reactor shall be 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 />.
G. Thermal Power Core Flow Allowable combinations of thermal power and total core flow shall be restricted to that shown in Figure 3 3 1 for two loop operation or Figure 3 3.2 for single loop operation.
- 1. In addition, the following power / flow restrictions apply during single loop operation:
- a. Drive (pump) flow in the active loop 11 not exceed rated flow. Rated flow is defined as the loop recirculation flow, not I to' exceed 16.74 x 106 lbs/hr (43,000 gpm) per loop, needed to I achieve 100% core flow with two recirculation loops in operation.
1
- b. Core plate dP pressure fluctuations shall not exceed 2.9 psi, '
peak-to-peak.
- c. APRM neutron flux noise levels shall not exceed three (3) times their established baseline value.
SURVEILLANCE REQUIREMENT 4.3 REACTIVITY CONTROL E. Reactivity Anomalies i
During startups following refueling outages, the critical rod <
configurations will be compared to the expected configurations at selected I l
Millstone Unit 1 3/4 3-8
SURVEILLANCE REQUIREMENT-4.3~ REACTIVITY CONTROL (continued) operating conditions. These comparisons will be used as base data for reactivity monitoring during subsequent power operation throughout the fuel cycle. At specific power operating conditions, the critical rod configuration will be compared to the configuration expected based upon appropriately. corrected past data. This comparison will be made at least every equivalent full power month.
G. Thermal Power Core Flow During single recirculation loop operation, the following shall be checked at least once per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s: 1
- 1. Conformance to the allowable operating region of Figure 3 3 2.
- 2. Drive (pump) flow in the active recirculation loop is less than rated flow.
- 3. Core plate dP pressure fluctuations less than 2 9 psi, peak-to-peak.
- 4. APRM neutron flux noise levels less than three (3) times their established baseline value. The beseline value for single loop operation must be established before exceeding 50% of rated thermal power in single loop operation.
Millstone Unit 1 3/4 3-8a ;
I-l-
=____--_________-__.
O ALLOWABLE COMBINATIONS OF TOTAL CORE FLOW AND POWER LEVEL FOR TWO LOOP OPERATION FIGURE 3 3 1 (See Note 4 TS Page B 2-4) 140 1003 Power Line 120 1003 Intercept g Point (100/87) T l 1 i APRM Rod Block Line (103/100)
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M LIMITING CGNDITION FOR OPERATION L ,
H l L ~3.6 PRIMARY SYSTEM BOUNDARY G. ' Jet Pumps
- 1. Whenever the reactor is'in the STARTUP/ HOT STANDBY or.RUN modes, all jet. pumps shall.be intact and all operating jet pumps shall be operable. .If it is. determined that a jet pump 1s inoperable, an
~
orderly shutdown shall be initiated and the reactor shall be in the
' COLD SHUTDOWN or REFUEL condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
'2. Flow indication ' from each of tlie twenty jet pumps (except that flow indication from only 19 jet pumps is acceptable prior to and during Cycle '12 operation) shall be verified prior to initiation of reactor startup from a cold shutdown condition.
3 For two recirculation loop operation, the indicated core flow is the sum of theLflow indication from each of the twenty jet pumps. If-l flow indication failure occurs for two or more jet pumps, immediate corrective action shall be taken. If. flow indication cannot. be obtained for at least nineteen jet pumps, an orderly shutdown shall be initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the reactor shall be in the COLD SHUTDOWN or REFUEL condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 4. For single recirculation loop operation, actual core flow is calculated as:
~ ~
Active Loop ' Inactive Loop ~
Actual Core Flow = Indication .95 Indication Where loop indication is the sum of the flows indicated by the single tap jet pumps in each loop.
SURVEILI.ANCE REQUIREMENT 4.6 PRIMARY SYSTEM BOUNDARY G. Jet Pumps
- 1. Whenever there is a recirculation flow with the reactor in the STARTUP/ HOT STANDBY or RUN modes, jet pump integrity and operability shall be checked daily by verifying that the following two conditions do not occur:
- a. The recirculation pump flow differs by more than 10% from the established speed-flow characteristics; or
- b. The indicated total core flow is more than 10% greater than the core flow value derived from established power-core flow relationships.
Millstone Unit 1 3/4 6-13
SURVEILLANCE REQUIREMENT 4.6; PRIMARY SYSTEM BOUNDARY (continued)
- 2. Additionally, when operating with one recirculation pump with the l equalizer line isolated, the diffuser to lower plenum differential pressure shall be checked daily, and the differential pressure of any jet pump in both loops shall not vary by more than 10% from established patterns.
3 The baseline data required to evaluate the conditions in Specification 4.6.G.1 will be acquired each operating cycle for two loop operation and data for Specification 4.6.G.1 and 4.6.G.2 will be acquired when required for single loop operation.
Millstone Unit 1 3/4 6-13a
A LIMITING CONDITION FOR OPERATION 36 PRIMARY SYSTEM BOUNDARY H. Recirculation Pump Flow Mismatch
- 1. Whenever both recirculation pumps are in operation, pump speeds shall be maintained within 10% of each other when power level is greater than 80% and within 15% of each other when power level is less than 80%.
- 2. If Specification 3 6.H.1 cannot be met, one recirculation pump shall be tripped. Operation with a single recirculation pump is permitted with the designated adjustments for APRM rod block and scram setpoints (Specification 2.1.2), RBM setpoint (Table 3 2 3), MCPR safety and. operating limits (Specifications 2.1.1 and 3 11.C, ,
respectively), and APLHGR (Specification 3.11.A). These adjustments '
shall be implemented within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after commencing single recirculation loop operation. If these adjustments are not made, the reactor shall be in the COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3 The reactor shall not be operated unless the equalizer line is isolated.
- 4. With the mode switch in the STARTUP/ HOT STANDBY or RUN MUDE, with no forced circulation, immediately initiate action to reduce thermal power to less than or equal to 40% of rated, and if at least one recirculation pump cannot be started, initiate measures to place the unit in STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENT 4.6 PRIMARY SYSTEM BOUNDARY H. Recirculation Pump Flow Mismatch
- 1. Recirculation pump speed shall be checked daily for mismatch.
Millstone Unit 1 3/4 6-14
e t
' LIMITING CONDITION FOR OPERATION' 3.11 REACTOR FUEL ASSEMBLY-
' Applicability
.The Limiting Conditions for. Operation associated with the fuel rods apply to those parameters which monitor the fuel rod operating conditions.
.0bjective The Objective of the Limiting Conditions for Operation is to assure.the performance of the fuel rods.
Specifications A. Average Planar Linear Heat Generation Rate (APLHGR)
- 1. During power operation,.the APLHGR, i.e., the LCO, for each type of fuel as a function of axial location and average planar exposure, shall-not exceed limits based on applicable APLHGR limit values that have been approved for the respective fuel and lattice types, as determined by the approved methodology described in GESTAR II. (This approval is based.on and limited to the GESTAR II methodology.) If hand calculations are required, the APLHGR for each type of fuel as a function of average planar exposure shall not exceed the limiting value for the most limiting lattice (excluding natural U) shown in Figure 3.11.1.
During operation with a single recirculation loop, the APLHGR shall not exceed 86% of the two loop rated core flow values in Figure 3.11.1.
- 2. If at any time during operation it is determined by normal surveillance .that the limiting value for APLHGR specified in Section 3.11.A.1 is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR is not returned to within. the prescribed limits within two (2) hours, tne 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.
SURVEILLANCE REQUIREMENT 4.11 REACTOR FUEL ASSEMBLY Applicability The Surveillance Requirements apply to the parameters which monitor the fuel rod operating conditions.
Millstone Unit 1 3/4 11-1
M t
SURVEILLANCE REQUIREMENT 4.11 REACTOR FUEL ASSEMBLY (continued)
Specifications
!- A. Average Planar Linear Heat Generation Rate (APLHGR)
The APLHGR for each type of fuel, as a function of average planar exposure shall be determined daily during reactor operation at 125% RATED THERMAL POWER.
l Millstone Unit 1 3/4 11-1a
_ m
LIMITING CONDITION FOR OPERATION l
3 11 REACTOR FUEL ASSEMBLY C. Minimum Critical Power Ratio (MCPR)
During power operation, MCPR shall be as shown in Table 3 11.1 for two recirculation loop operation. For one loop operation, the MCPR limits at rated flow increase by 0.01. If at any time during operation it is
. determined by normal surveillance that the limiting value for MCPR is baing exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought ta 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.
For core flows other than rated, the MCPRs in Table 3 11.1 shall be multiplied by Kp , where Kg is as shown in Figure 3 11.2.
D. If any of the limiting values identified in Specifications 3 11.A, B, or C, are exceeded, even if corrective action is taken, as prescribed, a Reportable Occurrence report shall be submitted.
SURVEILLANCE REQUIREMENT 4.11 REACTOR FUEL ASSEMBLY C. Minimum Critical Power Ratio (MCPR) ;
1
- 1. MCPR shall be determined daily during reactor power operation at j 2 25% RATED THERMAL POWER and following any change in power level or j distribution that would cause operation with a limiting control rod j pattern as described in the bases for specification 3 3.8.5. )
l
- 2. Utilization of Option B Operating limit MCPR values requires the I scram time testing of 15 or more control rods on a rotating basis every 120 operating days.
t a
4 a
b Millstone Unit 1 3/4 11-6
e a
TABLE 3.11.1 OPERATING LIMIT MCPRS FOR CYCLE 12#
(OPTION B)
BOC 12 TO EOC EOC 12 TO 70% COASTDOWN FUEL TYPE 1.34 1.34 BP8 x 8R (GE-7B) 1.34 1 34 GE8 x 8EB (GE-8B)
OPERATING LIMIT HCPRS FOR CYCLE 128 (OPTION A)
BOC 12 TO EOC EOC 12 TO 70% COASTDOWN FUEL TYPE 1.39 1.39 BP8 x 8R (GE-7B)
I 1.39 1.39 GE8 x 8EB (GE-8B) ,
Based on two recirculation loop operation. Limits increase by 0.01 for single loop operation.
Millstone Unit 1 3/4 11-7
32 PROTECTIVE INSTRUMENTATION EjSES The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to < 1.07 for two loop operation or 1.08 for single loop operation. The trip logic for this function is 1 out of n; e.g., any trip on one of the six APRM's, eight IRM's, or four SRM's will result in a rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the IRM and RBM may be reduced by one for a short period of time to allow for maintenance testing and calibration.
The APRM rod block trip is flow biased and prevents significant approach to MCPR=1.07 for two loop operation or 1.08 for single loop operation especially during operation at reduced flow. The APRM provides gross core l protection, i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence, The trips are set so that fuel damage limits are not exceeded.
The RBM provides local protection of the core, i.e., the prevention of fuel damage in a local region of the core, for a single rod withdrawal error.
The trip point is flow biased. The worst case single control rod withdrawal error has been analyzed for the initial core and also prior to each reload; the results show that, with specified trip settings, rod withdrawal is blocked within an adequate margin to fuel damage limits. This margin varies slightly from reload to reload and, thus, each reload submittal contains an update of the analysis. Below es70% power, the withdrawal of a single control rod results in MCPR > 1.07 for two loop operation or 1.08 for single loop operation l without rod block action, thus requiring the RBM system to be operable above 30% of rated power is conservative. Requiring at least half of the normal LPRM inputs from each level to be operable assures that the RBM response will be adequate to prevent rod withdrawal errors.
The IRM rod block functions assure proper upranging of the IRM system, and reduce the probability of spurious scrams during startup operations.
A downscale indication of an APRM or IRM is an indication the instrument has failed, the instrument is not sensitive enough, or the neutron flux is below the instrument response threshold. In these cases the instrument will not respond to changes in control rod motion and thus control rod motion is prevented. The downscale trips are set at 3/125 of full scale.
To prevent excessive fuel clad temperature for the small pipe break, the FWCI or Isolation Condenser systems must function, since for these breaks reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function and Isolation Condenser system are provided as back-ups to the FWCI in the event the FWCI does not operate. The arrangements of the tripping contacts are such as to provide these functions when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criterion is met. Ref. Section VI-2.0 FSAR. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.
Millstone Unit 1 B 3/4 2-4
E .
33 REACTIVITY CONTROL BASES Power / flow restrictions applied during single recirculation loop operation serve several purposes. Limiting loop flow and pump spep assuresequipmentandvibrationlimitsarenotexceeded.tg)torated Limiting total core flow to 56% and core thermal power to 65% of rated ensures plant operation within the bounds analyzed in Reference 9 In addition, power and flow limits were chosen to maximize the power attainable in single recirculation loop operation while keeping vibration and system oscillations (e.g., jet pump flow, APRMs and core plate dP fluctuations).
at acceptable levels. The maximum power / flow attainable occurs when any of these limits is reached.
l (8) Single recirculation pump operation, 22A7456, Revision 0. I (9) Millstone Point Nuclear Power Station, Unit No. 1 Single Loop Operation, NEDO-24312, December 1980.
k Millstone Unit 1 B 3/4 3-7
3.6 and 4.6 PRIMARY SYSTEM BOUNDARY BASES A break in a jet pump decreases the flow resistance characteristic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared with previous operation.
The change in flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop.
Comparison of the data with a normal relationship or pattern provides the indication necessary to detect a failed jet pump.
For Cycle 12 operation, flow indication for jet pump "K" will be determined by flow indication from the adjacent, paired jet pump "J,"
based upon historical jet pump performance characteristics.
The jet pump flow deviation pattern derived from the diffuser to lower plenum differential pressure readings will be used to further evaluate jet pump operability in the event that the jet pumps fail the tests in Sections 4.6.G.1 and 2.
Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive or broken jet pumps. This bypass flow is reverse with respect to normal jet pump flow. The indicated total core flow f a summation of the flow indications twenty individual jet pumps.
The tt .1 core flow measuring instrumentation sums reverse jet pump flow as though it were forward flow. Thus, the indicated flow is higher than actual core flow by at least twice the normal flow through any backflowing pump. Reactivity inventory is known to a high degree of confidence so that even if a jet pump failure occurred during a shutdown period, subsequent power ascension would promptly demonstrate abnormal control rod withdrawal for any power-flow operating map point.
A nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible.
The equation for calculating core flow during single recirculation loop operation is the result of a conservative analysis to modify the single-tap flow coefficient for reverse flow.(l) The 0.95 factor is defined as the ratio of " Inactive Loop True Flow" to " Inactive Loop Indicated Flow."
" Loop Indicated Flow" is the flow indicated by the single-tap jet pump loop flow summers and indicators which are set to indicate forward flow correctly. Thus core flow during one-pump operation is determined in a conservative way, and its uncertainty has been conservatively evaluated.
t (1) Millstone Point Nuclear Power Station, Unit 1, Single Loop Operation, NEDO-24312, December 1980. !
l Millstone Unit 1 B 3/4 6-6
( __ _ _ _ _ _ _ _ _ _ _ ___ _____ _ _ _
F~ . .. ,
h.
3 6'and 4.6 PRIMARY. SYSTEM BOUNDARY!(continued)
BASES H.- Recirculation Pump' Flow Mismatch The LPCI loop selection logic is' described in the FSAR, Section 6.2.4.2.
For some limited, low probability accidents, with the recirculation loop operating with large speed differences, it is possible for the logic to select the wrong loop for injection.
I For these limited conditions the core spray itself is adequate to prevent fuel temperatures from exceeding allowable limits. However, to limit.the probability even further, a procedural limitation has been placed on the allowable variation in speed between the recirculation pumps.
The analyses for Quad. Cities indicate that above 80% power the loop select logic could not be expected to function at a speed differential of 15%.
Below 80%. power the. loop select logic would not be expected to function at a speed differential of 20%. This specification provides a margin of 5%
in pump speed differential before a problem could arise. If the reactor
'is operating on one pump, the loop select logic trips that pump before making the loop selection.
The impact of operation with one recirculation loop upon plant safety analyses was assessed.(2) This assessment showed that single loop-operation is permitted if the MCPR fuel cladding integrity safety limit and operating limits are increased as noted in Specifications 2.1.1 and 3 11.C, APLHGR limits are decreased by the factors given in Specification 3.11.A, and APRM rod block and scram setpoints and the RBM setpoint are adjusted as noted in Specification 2.1.2 and Table 3 2 3 Immediately initiating action to reduce thermal power to less than or equal to 40% of rated without forced circulation forces operation in a region of the power / flow map where stability problems are minimal, i.e.,
decay ratio less than 0.6, while attempts to restart at least one recirculation pump are made. This approach was found acceptable in Reference 3 (2) " Millstone Point Nuclear Power Station, Unit 1, Single Loop Operation,"
-NEDO-24312, December 1980.
(3) Fafety Evaluation of the General Electric Topical Report, NEDE-24011, General Electric Standard Application for Reload Fuel, Amendment 8, Thermal Hydraulic Stability, March, 1985 Millstone Unit 1 B 3/4 6-6a b
3 11 REACTOR FUEL ASSEMBLY !
l BASES ;
A. Average Planar Linear Heat Generation Rate (APLHOR)
This specification assures that the peak cladding temperature following the postulated design basis locs-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K.
The peak cladding temperature 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. Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than 1200F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendiv K limit.
The limiting value for APLHGR is shown in Figure 3 11.1. The calculational procedures used to establish the APLHGR limiting values are documented in References 1 and 2.
Conservative LOCA calculations predict that nucleate boiling will be maintained for several seconds following a design basis LOCA. This results in early removal of significant amounts of stored energy which, if present later in the transient, when heat transfer coefficients are considerably lower, would result in higher peak cladding temperature. As core flow is reduced below about 90%, the time of onset of boiling transition makes a sudden change from greater than about 5 seconds to less than 1 second. The approved ECCS evaluation model requires that at the first onset of local boiling transition, the severely reduced heat transfer coefficients must be applied to the affected planar area of the bundle, and thus exaggerates the calculated peak clad temperature. The effect is to significantly reduce the energy calculated to be removed from the fuel during blowdown. This results in an increase in calculated peak clad temperature of about 1000F which can be offset by a 5% reduction in MAPLHGR. For flows less than 90% of rated, a 5% reduction in the MAPLHGR limits in Figure 3 11.1, derived for 100% flow will assure that the plant is operated in compliance to 10 CFR 50.46 at those lower flows. The models for assessing reduced core flow effects, including single loop operation, are documented in References 1, 3 and 4. As shown in Reference 1, a reduction in full power / full flow MAPLHGR using a 0.86 multiplier assures compliance with 10 CFR 50.46 for single loop operation.
B. Linear Heat Generation Rate (LHGR)
This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation rate. The LHGR shall be checked daily during reactor operation at 2 25% power to determine if fuel burn-up, or control rod movement, has caused changes in power distribution. For LHGR to be a limiting value below 25% RATED THERMAL POWER, the MTPF would have to be greater than 10 which is precluded by a considerable margin when employing any permissible control rod pattern.
Millstone Unit 1 B 3/4 11-1
e 3 11 REACTOR FUEL ASSEMBLY BASES C. The steady state value for MCPR was selected to provide a margin to accommodate transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation itself.
This value ensures that:
- 1. For the initial conditions of the LOCA analysis, a MCPR of 1.18 is satisfied. For the low flow ECCS analysis, an initial MCPR of 1.24 is assumed, and
- 2. For any of the special transients, or disturbances, caused by single operator error or single equipment malfunction the value of MCPR is conservatively assumed to exist prior to the initiation of the transient or disturbance.
At core thermal power levels 25%, the reactor will be operating at minimum recirculation pump speed, and moderator void content will be very small. For all designated control rod pritterns which may be employed at this power, thermal hydraulic analysis indicates that the resultant MCPR value is in excess of requirements. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for calculation of MCPR at greater than 25% RATED THEPMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. Changes in MCPR for single loop operation are discussed in Reference 1. During such operation, MCPR operating limits are increased by 0.01 compared to two loop rated flow conditions.
The use of the Option B operating limit MCPR requires additional SCRAM time testing and verification in accordance with GE letter, A. D. Vaughn to R. M. Matheny, April 21, 1987, regarding Potential Technical Specification Changes for Implementation of Advanced Methods.
D. Reporting Requirements The LCOs associated with -monitoring the fuel rod operating conditions are required to be met at all times or corrected to within the limiting values )
of MAPLHGR, LHGR, and MCPR within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the time the plant is determined to be exceeding them. It is a requirement, as stated in Specifications 3.11. A, B, and C, that if at any time during power operation it is determined that the limiting values for MAPLHGR, LHGR, or MCPR are exceeded, action is then initiated to restore operation to within the prescribed limits. This action is to be initiated within 15 minutes !
if normal surveillance indicates that an operating limit has been reached. -
Each event involving operation beyond a specified limit shall be logged I and a reportable occurrence issued. It must be recognized that there is l always an action which would return any of the parameters (MAPLHGR, LHGR, I or MCPR) to within prescribed limits, namely power reduction. Under most circumstances, this will not be the only alternative.
Millstone Unit 1 B 3/4 11-2 l
L_ _ _ _ _ _ _ _ _ ___-.._____--_______________U
311 REACTOR FUEL ASSEMBLY BASES
~E. References
- 1. " Millstone Point Nuclear Power Station, Unit 1, Single Loop Operation," NED0-24312, December 1980.
- 2. " Generic Reload Fuel Application," NEDE-24011-P-A* and NEDO-24011-P-A-US.#
3 " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K, July 1978, NEDO-20566."
- 4. " Loss-of-Coolant Accident A'salysis Report for Millstone Unit 1 Nuclear Power Station, Juli 1980, NED0-24085-1. "
- Approved version at time reload analyses are performed.
i l
Millstone Unit 1 B 3/4 11-3
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ -