ML20099K512
| ML20099K512 | |
| Person / Time | |
|---|---|
| Site: | Monticello, 05000000 |
| Issue date: | 09/07/1976 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20099K024 | List: |
| References | |
| FOIA-84-105 NUDOCS 8411290455 | |
| Download: ML20099K512 (9) | |
Text
'
4 Bases Continued:
temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimiser. Worth of individual rods is very law in a uniform rod pattern. H us, of all possibir sources of reactivity input, uniform control rod withdraws! is the most probable cause of significant power rise. Recesse the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods meat be moved to change power by a significant percentage of rated power, the rate of power rise is very alow. Generally, the heat flux is in near equilibrisma with the fission rate. In an assumed
. uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated power per minute, and the IRM system would be amore than adequate to assure a scram before the power could exceed the safety limit. S e IRM screa remains active until the mode switch is placed in the run position. %1s switch occurs when reactor pressure is greater than 850 psig.
Operation with one recirculation pump and the equalizer line closed causes flow to bypass the core through idle jet pumps. nis loss of core flow is accounted for in the specified setting adjust-ments for the one-pump, equalizer closed case. Reference the September 7, 1976 License A:aendment Request from L 0 Mayer (NSP) to Victor Stello (USNRC).
S e operator will set the APRM neutrou flux trip setcing no greater than thn: shown in Figure 2.3.1.
Navaver, the actual setpoint can be as unsch as 31 greater than that shuwn on Figure 2.3.1 for recirculation driving flows less than 50% of design and 27, greater than that shown for recirculation driving flows greater than 50% of design due to the deviations discussed on page 18.
B.
APWI Control Rod Block Trips Reactor power level may be varied by moving control rods or by varying the recirculation flow rate..ne APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate, and thus to protect against the condition of a HCPR less than the Safety Limit (T.S.2.1.A).
His rod block trip setting, wtaich is automatically varied with recirculation to<,p flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. H e flow variable trip setting provides substantial margin froma fuel damage, assianing a steady-state operation at the trip setting, 2.3 BASES 20 8411290455 840419 PDR FOIA BELL 84-105 PDR
2 Bases Continued:
l over the entire recirculation flow range. n e margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship; 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. %e actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPlet system. Operation with one recirculation pump and the equalizer line closed causes flow to bypass the core through the idle jet pumps. nis loss of core flow is accounted for in the specified setting adjustments for the one-pump, equalizer-closed case.
Reference the September 7,1976 License Amendment Request frcan L 0 Mayer (NSP) to Victor Stello (USNRC). When the maximum total peaking factor exceeds the design value, the rod block setting is adjusted in accordance with the formula in Specification 2.3.B.
If the APlui rod block setting should require a change due to an abnormal peaking condition, it will be done by increasing the APRM gain and thus reducing the slope and Intercept point of the flow referenced rod block curve by the reciprocal of the APRM gain change.
H e operator will set the APIOt rod block trip settings no greater than that shown in Figure 2.3.1.
However, the actual setpoint can be as much as 3% greater than that shown on Figure 2.3.1 for recirculation driving flows less than 50% of design and 2% greater than that e me.. for recirculation driving flows greater than 507. of design due to the deviations discouaed on r. ige 18.
C.
Reactor Low Water Level Scram The reactor low water icvel scram 1:: ec at a point which will assure that the water levet used in the bases for the safety limit is mainta.ned.
h e operator will set the low water level trip setting no lower than 10'6" above the top of the active fuel. Ilowever, the actual setpoint can be as much as 6 inches lower due to the deviations discussed on page 18.
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 energy associated with the loss of coolant accident and to limit fuel clad temperature to well below the clad melting temperature to assure that core gennetry remains intact and to limit any clad netal-water reaction to less than 1%.
The design of the ECCS components to meet the above criterion was dependent on three previously set parameters:
the raaximum break size, the low water level scram setpoint, and the ECCS initiation set-point. To lower the setpoint for initiation of the ECCS could prevent the ECCS components from 2.3 BASES 21
1 C
C C
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D.3 - Continued
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hat Initiates Rod Block
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ich Min. No. of Oper-
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'Ibtal No. of able or Operating W}. g @j[ pi dch*
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Channels per Per Trip System Bequired ow-Instrument Instrument Channels
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Ihan Trip System (Notes 1,6)
Conditions *
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for one of the tuo t.rtp ny ;tums, this condition
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- v iecirculation driving flow in percent.
(,
- p h3)'Only one of the four SlH channels may be bypassed.
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(4) 'Diere must be at least one operable or operating IIM channel monitoring each core quadrant.
(5) One of the two RIMS may be bypusced for maintenance c. a/or testing for pericxis not in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30 day period. An RIN channel will be conside..ed inoperable if there am less than half the total ntunber of normal inputs from any LPfN level.
3 2A.2 i
u
Bases Continued:
3.2 he HPCI and/or RGIC high 11sw and temperature instrumentation is provided to detect a break in the HPCI and/or RCIC piping. Tripping of this instrumentation results in actuation of HPCI and/or RCIC isolation valves; i.e., Group 4 and/or Group 5 valves. he trip settings of 200 F and 1507. of HPCI and 3007, of RCIC design flows and valve closure times are such that the core will not be uncovered and fission product release will not exceed 10 CFR 100 guidelines.
he instrumentation which initiates ECCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion the Specification preserves the effectiveness of the systesa even during periods when maintenance or testing is being performed.
he control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR remains above the Safety Limit (T.S.2.1.A).
He 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. he miniums instrument channel requirements for the IRM and RBH may be reduced by one for a shott period of time to allow for maintenance, testing, or calibration. See Section 7.3 FSAR.
%e APRM rod block trip is referenced to flow and prevents a significant reduction in MCPR especially during operation at reduced. flow. The APRM provides. gross core 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 MCPR is maintained greater than the Safety Limit.
i Tb RBM provides local protection of the core; i.e.,
the prevention of critical power in a local region of the core, for a single rod withdrawal error from a limiting control rod pat tern.
The trip point is referenced to flow. he worst case single control rod withdrawal error haa neen analyzed and the j
results show that with the specified trip settings rod withdrawal ic blocked at MCPR greater than the Safety Limit, thus allowing adequate margin.
Below 607. power, MCIR i eniains above the Safety Limit for the worst case withdrawal of a single control rod without rod block action, thus below this level it is not required. His subject -is discussed in General Electric INR Thermal Analysis Basis (CETAB):
Data Correlation and Design Application, NEDO-10958. Requiring at least half of the nomal LPRM inputs
{
from each level to be operable assures that the RBH response will be adequate to prevent rod with-l drawal errors. Operation with one recirculation pump and the equalizer line closed causes flow to bypass the core through idle jet pumps. His loss of core flow is at. counted for in the specified setting adjustments for the one-pump, equalizer closed case.
Reference the September 7,1976 License Amendment Request from L 0 Mayer (NSP) to Victor Stello (USNRC).
%e IRM rod block function provides local as well as gross core protection. he scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst case accident results in rod block action before MCPR approaches the Safety Limit (T.S.2.1.A).
A downscale indication of an APRM or 1RM is an indication the instrument has failed or the instrument is not sensitive enough.
In either case the instrument vill not respond to changes in control rod motion and thus control rod motion is prevented. E c downscale trips are set at 3/125 of full scale.
3.2 BASES 67 i
9 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS I.
R: circulation System I.
Recirculation System 1.
Except as specified in 3.5.1.2 below, whenever 1.
Once per month, when;irriated) fuel is in the irradiated fuel is in the reactor, with reactor reactor with reactor coolant tamperature greater coolant tamperature greater than 212*F and both than 212 F and both reactor recirculation reactor recirculation pumps operating, the pumps operating, the recirculation system cross recirculation system cross tie valve interlocks tie valve interlocks shall be demonstrated to shall be operable.
l be operable by verifying that the cross tie valves cannot be opened using the normal control 2.
1he recirculation system cross tie valve inter-switch.
locks may be inoperable if at least one cross tie valve is maintained fully closed.
2.
When a recirculation system cross tie valve interlock is inoperable, the position of at least one fully closed cross tie valve shall 3.
Operation with one recirculation pump be recorded daily.
(equalizer valve closed) is pennitted provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time one pump operation consnences the appropriate adjustments to limits specified in Specifications 2.3. A.1, 2.3. B, 3.2.C and 3.11. A are incor-porated. If the settings cannot be adjusted or two-pump operation restored by the end of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an orderly reactor shutdown shall be initiated.
3.5/4.5 lO8A l
Bases continued 3.5 C.
Emergency Cooling Availability The purpose of Specification C is to assure that sufficient core cooling equipment is available at all times.
It is during refueling outages that major maintenance is performed and during such time that all core and containment cooling subsystems may be out of service. Specifi-cation 3.5.G.3 allows all core and containment cooling subsystems to be inoperable provided no work is being done which has the potential for draining the reactor vessel. Thus events requiring core cooling are precluded.
Specification 3.5.G.4 recognizes that concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chamber for maintenance or for the inspection required by Specification 4.7.A.I.
In this situation, a sufficient in-ventory of water is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal integrity.
H.
Deleted I.
Recirculation System The capacity of the Emergency Core Coolant System is based on the potential consequences of a double ended recirculation line break. Such a break involves 3.9 sq. ft, when the cross tie valves are closed and 5.3 sq. ft, when the cross tie valves are open.
Specification 3.11.A is based on an ECCS evaluation assuming a break area of 3.9 sq. tt.; the limitations of 3.11. A do not ' apply to the larger break area. Therefore, at least one cross tie valve must remain closed at all times to reduce the potential break area as required by Specifications 3.5.1.1 and 2.
An analysis of one-pump operation (equalizer valve closed) identifies certain limitations peculiar to that mode of operation. Reference the September 7,1976 License Amendment Request from L 0 Mayer (NSP) to Victor Stello (USNRC). Operation with only one pump is not a normal mode; it will generally involve a forced outage of equipment. There may be insufficient time to make adjustments to the RBM and APRM flow referenced rod block and scram prior to commencing one-pump operation. The reduction in power with the reduced core flow will cause the APLilGR to reduce accordingly, naturally moving in the direction of the new limit. Specification 3.5.I.3 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before these new Ilmits are required to be implemented.
~ 3.5 BASES 113 I
I e
3.0 LIMITING CONDITIONS FOR OPER[flait 4.0 SURVEILIANCE REQUIREENTS 3.11 REAC'IOR E11EL ASSEMBLIES 4.11 REACTOR FUEL ASSEMBLIES Applicability Applicability "Ihe Limiting Conditions for Operation The Surveillance Requirements apply to associated with the fuel rods apply to the parameters which monitor the fuel those parameters which monitor the fuel rod operating conditions.
red operating conditions, t
Objective Objective The objective of the Surveillance Require-4 The objective of the Limiting Conditions ments is to specify the type and for Operation is to assure the perfor-frequency of surveillance to be applied a
mance of the fuel rods.
to the fuel rods.
Specifications Specifications i
A.
Average Planar Linear lleat Generation A.
Average Planar Linear Heat Generation Rate (APIJIGR)
Rate (APIJ!Cril l
During two-pump power operation, the APUICR The APijlGts-for each type of fuel as a for each type of fuel as a function of f.metion of average planar exposure average planar exposure shall not exceed the shall be determined daily during limiting value shown in Figures 3.11.1; reactor operation at 2 257. rated i
during one-pump operation (equalizer valve thermal power.
closed) the respective limit shall be 0.86 times the value shown in these figures. If at any time during operation it is determined that the limiting value for APUIGR is being i
exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. Surveillance and correspond-ing action shall continue until reactor opera-tion is within the prescribed limits. If the APIJICR is not returned to within the pre-scribed 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 />.
3.11/4.11 189B i
r 3
- a Bases 3.11 A.
Average Planar Linear IIeat Generation Rate (AP!JICR)
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 10CFR50, Appendix K.
The peak cladding temperature following a postulated loss-of-cool' ant 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 cladding temperature by less than + 20 F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure Lliat calculated temperatures are within the 10CFR50 Appendix K limit. The limiting value for APLIIGR is given by this specification.
Operation with one reactor recirculation pump and the equalizer line closed causes flow to bypass the core through the idle jet pumps. This loss of core flow results in an increased peak clad temperature during a LOCA. The limits on APillGR are reduced correspondingly.
Reference the September 7, 1976 License Amendment Request fran L 0 Mayer (NSP) to Victor Stello (USNRC).
Those abnonmal operational transients, analyzed in FSAR Section 14.5, which rer. ult in an automatic reactor scram are not considered a violation of the LCO.
Exceeding APLIICR limits in such cases need not be reported.
B.
LHGR This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of Reference 1 and in References 2 and 3, and assumes a linearly increasing. variation and axial gaps between core bottom and top and assures with a 957 confidence, that no more than one fuel rod exceeds the design linear heat generation rate due to power spiking.
Those abnormal operational transients, analyzed in FSAR Section 14.5, which result in an automatic reactor scram are not considered a violation of the LCO.
Exceeding LilGR limits in such cases need not be reported.
3.11 BASES 189 E i
.