ML19260C361
| ML19260C361 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 12/12/1979 |
| From: | Howard J BOSTON EDISON CO. |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19260C362 | List: |
| References | |
| NUDOCS 7912260242 | |
| Download: ML19260C361 (39) | |
Text
. _..
BOSTON EDISON COMPANY B00 BOYLSTON STREET BOSTON. MABBACHUBETTs 02199 J. EoWARD HOWARD viss possimaser e8WoLEAS December 12, 1979 BECo. Ltr. #79-269 Mr. Thomas A. Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission k'ashington, D. C.
20555 License No. DPR-35 Docket No. 50-293 Reload 4 Submittal and Request for Technical Specification Changes
Reference:
- a. " Generic Reload Fuel Application" NEDE-24011-P-A, July 1979
- b. " Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Unit 1 Reload 4",
NED0-24224, November 1979
- c. " Loss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station", NED0-21696, August 1977.
Dear Sir:
Thq fourth refueling outage for Pilgrim Nuclear Power Station, Unit #1 is scheduled to comence in January 1980. Analyses (Reference b.) supporting and justifying the operation of Pilgrim I during Cycle 5 are hereby submitted for your review.
Reference b supplements generic analyses previously submitted by General Electric by Reference a.
Applicable chesnges to Appendix A (Technical Specifications) to Facility Operating License (No. DPR-35) are also sub-mitted for your review, approval and issuance pursuant to Section 50 of Title 10, Code of Federal Regulations.
1617 149 7 912260 h[-
4
BDSTON EDISON COMPANY Mr. Thomas A. Ippolito, Chief December 12, 1979 Page 2 Proposed Technical Specification Changes (Including Reasons)
It is proposed to modify the existing Technical Specifications as described in the attached pages. The changes are designed to allow operation af ter Reload 4, reduce the need for Technical Specification changes for future reloads and to eliminate the power spiking penalty on Linear Heat Generation Rate.
1.
Specification 1.1A The Safety Limit MCPR has been changed to 1.07 based on the more conserva tive P8x8R fuel, which will be introduced into the core during Reload 4.
The phrase "The Safety Limit MCPR" replaces the former numerical value of 1.06 in the bases throughout the Technical Specifications.
2.
Specification 2.1A and B Specification 3.1 and 4.1 Specification 3.2C In the above specifications, the present trip reduction factor and criterion for the flow biased APRM scram and rod block trips are re-placed with a new reduction factor and a new criterion which are defined by quantities which are not bundle type dependent and which are directly available from the process computer.
3.
Specification 1.1 Specific values of input parameters to safety analyses have been replaced by references to the Generic Reload Licensing Submittal and to the Supple-mental Reload Licensing Submittal for the current cycle.
4.
Specification 3.11A A single curve of limiting APLHGR values is proposed which has been chosen conservatively below the lower envelope of the maximum allowable APLHGR values of all fuel types in the current reload.
5.
Specification 3.11B The reduction in permissible LHGR due to densification effects is deleted.
As noted in the Supplemental Reload Licensing Submittal (NED0-24224), the maximum calculated LHGR for the rod withdrawal error and the bundle loading error transients have been augmented by the potential effects of densification power spiking prior to comparison with the fuel damage limit LHGR.
The restriction on operating LHGR can therefore be removed.
6.
Specification 3.11C The operating limit MCPR is changed to 1.35.
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_ _ _ _ _. ~... _. _.. _ _ _
BDSTON EDISON COMPANY Mr. 'Ihomas A. Ippolito, Chief December 12, 1979 Page 3 Safety Considerations Generic information relative to the reload fuel design and analyses of BWR fuel is presented in GE Licensing Topical Report NEDE-24011-P-A, " Generic Reload Fuel Application", July, 1979 (Reference a).
This report is supple-mented by plant-specific information contained in Reference b.
(The reference core loading for Reload 4 is identified in this later document.) Together, these two documents provide the bases for the safety analysis and safety evaluation for Reload 4, and the proposed Technical Specification changes associated with the reload. The following narrative summarizes those safety aspects which are reload specific.
The Reload 4 fuel assemblies are similar in mechanical design to 3x8 assemblies previously licensed and operated in the Pilgrim 1 reactor. Minor dimensional changes have been made to fuel-rods and a second water rod has been added to improve operational and transient response.
The helium backfill pressure has been increased from 1 atmosphere to 3 atmospheres prepressurization.
The fuel is designated P8x8R (i.e. prepressurized 8x8 retrofit design).
Revisions to Reference a. involving the P8x8R fuel have been previously reviewed and ac-cepted by the NRC.
The Safety Limit MCPR for Reload 4 is established by the retrofit fuel. The value of 1.07 corresponding to the P8x8R fuel is therefore required in Technical Specifications and is included in the proposed Technical Specification changes.
All transients which are the basis of the Pilgrim license were~ reviewed for Reload 4.
Those transients which are critical with respect to safety margins and sensitive to the core reload parameter changes were reanalyzed.
The mos t restrictive condition is calculated to occur as a result of a Fuel Loading Error in which a fuel bundle is mis-rotated.
Analysis of this error results in a minimum permissible MCPR operating limit of 1.29 for P8x8R fuel at steady state conditions to avoid violating the Safety Limit MCPR at any time during Cycle 5.
A more restrictive MCPR operating limit of 1.35 is proposed in the Technical Specifications with the objective of establishing an operating limit MCPR which is fuel type and cycle independent, so that new Technical Specification will not need to be proposed each time the core is reloaded.
Reload 4 uses a different criterion for peak pressure during anticipated tran-sients than that used in previous reloads.
In the pas t, a 25 psi margin was recommended between the calculated peak pressure of the most limiting abnormal operational transient and the lowest spring safety valve (SSV) setpoint.
The purpose of this margin was to preclude SSV actuation during abnormal operational transient because of operational rather than safety considerations. To better reflect the operational nature of this concern, the criterion has been reformulated to incorporate consideration of expected frequency of transients.
Thus a larger margin (60 psi) has been applied to more frequent events (those which may occur more than once per plant lifetime).
The safety of reactor operation is not decreased by this change in criterion.
Under the new criterion, the most limiting transient with respect to pressure margin is the MSIV closure at end of cycle.
The peak pressure of 1158 psig shows that the 60 psi margin to the safety valve set point (1240 psig) is maintained.
1617 151
BOSTON EDIBON COMPANY Mr. Thomas A. Ippolito December 12, 1979 Page 4 The reactor vessel overpressure protection is verified by the analysis of the clos are of all main steam line isolation valves with an indirect (flux) scram.
At the end of Cycle 5 with all safety relief valves operating and an indirect scram the peak vessel pressure remains 34 psi below the peak allowable A9fE overpressure of 1375 psig at the vessel bottom. This pressure margin is re-duced from previous analysis due to the ef fect of the prompt recirculation pump trip on high pressure signal for ATWS protection.
Values of MAPLHGR for the new fuel bundles have been calculated using the methods described in References a. and c.
The curves of MAPLHGR versus exposure for the new fuel type as well as those for the 8x8 fuel fall above the values proposed as Technical Specifications which are thus conservative limits for operation of Pilgrim Cycle 5.
The Technical Specifications associated with this reload set MCPR and APLHGR limits more conservatively than required by the Reload 4 core.
Thus, safety margins are actually increased.
The potential effects of power spiking have been included in the reload analysis by augmenting the LGER's for rod withdrawal error and the bundle loading error.
Therefore, removal of the power spiking penalty from the LHGR Technical Specifi-cation does not reduce safety margins.
Substitution of the ratio FRP/MFLPD for A/MTPF in the APRM scram and rod block tr'r setting involves no safety question since the two expressions are equiv-alent. The fact that the new ratio is cycle independent and can be monitored easily from process computer output makes its use highly desirable.
Review of the nuclear design of the Pilgrim core with the Reload 4 fuel in place shows that the minimum shutdown margin with the strongest control rod fully withdrawn is calculated to be greater than 2% A K/K, which exceeds the 0.25% A K/K required by the Technical Specifications of the Pilgrim Nuclear Power Station plus the 0.04% AK/K allowance for inverted tubes in the control rod bla?.es.
The maximum incremental control rod worth using bank position withdrawal sequences in Cycle 5 is 0.95% A K.
This is below the Technical Specification limit of 1.0% AK and assures tha peak fuel enthalpy during a rod drop accident will be less than the 280 cal /gm design limit (Reference a).
The new fuel can be safety stored in the spent fuel pool since the maximum fuel loading and bundle average enrichment are within present Technical Specification limits of 16.0 gm U-235 per em and 3.0 w/o U-235, respectively.
1617 152
BOSTON EDISDN COMPANY Mr. Thomas A. Ippolito December 12, 1979 Page 5 k
Conclusions Based on the evaluation presented herein and the contents and analyses presented in Reference a. and b., it can be concluded that there is reason-able assurance that the health and safety of the public will not be endangered by operation of the Pilgrim Nuclear Power Station, Unit #1
'following Reload No. 4.
'Ihis proposed amendment has been reviewed and approved by the Operations Review Committee and reviewed by the Nuclear Safety Review and Audit Conmittee.
Schedule The Boston Edison Company tentatively plans to commence the refueling outage for Pilgrim I on January 5, 1980. Therefore, an expeditious review and approval of this submittal is requested.
Fee Consideration In accordance with Section 170.12 of the Commission's Regulations, Boston Edison proposes this license change es Class III since it utilizes NRC approved topical reports as referenced. Accordingly, a check for Four Thousand Dollars ($4,000) is enclosed.
Should there be any questions regarding this submittal, please contact us.
Very truly yours, mu e
l) 3 signed originals and 40 copies Attachments - (1) Proposed Changes to Appendix A (Technical Specifications)
(2) NEDO 24224, November, 1979 1617 153 Commonwealth of Massachusetts)
County of Suffolk
)
Then personally appeared before me J. Edward Howard, who, being duly sworn, did state that he is Vice President - Nuclear of Boston Edison Company, the applicant herein, and that he is duly authorized to execute and file the submittal contained herein in the name and on behalf of Boston Edison Company and that the statements in said submittal are true to the best of his knowledge and belief.
9 Li I)l9 O I ntt( A 3
My Commission expires:
t NotaryPublicjG
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1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITI 2.1 FUEL CLADDING INTEGRITY Applicability:
Applicability:
Applies to the interrelated Applies to trip settings of the variables associated with fuel instruments and devices which are provided to prevent the reactor thermal behavior.
system safety limits from being exceeded.
Objective:
Objective:
To establish limits below which To define the level of the process variables at which automatic pro-the integrity of the fuel tective action is initiated to cladding is preserved.
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 limiting safety system trip The existence of a minimum settings shall be as specified critical power ratio (MCPR) less below:
than 1.07 shall constitute vio-lation of the fuel cladding Nuetron Flux Trip Settings integrity safety limit. A MCPR 1.
of 1.07 is hereinafter referred to as the Safety Limit MCPR.
a.
APRM Flux Scram Trip Setting (Run Mode)
B.
Core Thermal Power Limit (Reactor Pressure 1800 psia and/or Core When the Mode Switch is Flow 110%)
in the RUN position, the APRM flux scram trip When the reactor pressure is 5800 setting shall be:
psia or core flow is less than or equal to 10% of rated, the steady S1 65W + 55% 2 loop state core thermal power chall not exceed 25% of design thermal power.
Where:
C.
Power Transient S = Setting in percent of rated thermal The safety limit shall be assumed power (1998 MWt) to be exceeded when scram is known to have been accomplished by a W = Percent of drive means other than the expected flow to produce scram signal unless analyses a rated core flow demonstrate that the fuel of 69 M lb/4c cladding integrity safety limits defined in Specifi-cations 1.lA and 1.lB were not exceeded during the actual transient.
6
i 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEH SETTING D.
Whenever the reactor is in the In the event of operation with a cold shutdown condition with maximum fraction of limiting power density (MFLPD) greater than the irradiated fuel in the reactor vessel, the water level shall not fraction of rated power (FRP),
be less than 12 in, above the top the setting shall be modified as of the normal active fuel zone.
follows:
FPP 2 Toop S $ (0.65W + 55% )
- 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/f t for 8x8 and P8x8R fuel.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
For no combination of loop recircula-tion flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.
b.
APRM Flux Scram Trip Setting (Refuel or Start and Hot 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.
c.
B.
APRM Rod Block Trip Setting The APRM rod block trip setting shall be:
SRB 1 0.65W + 42%
2 Loop 1617 155 7
1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING
- khere, SRB = Rod block setting in percent of rated thermal power (1998 MWt)
W = Percent of drive flow required to produce a rated core flow of 69M lb/hr.
In the event of operating with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
_. FRP ~
P SRB $ (0.65 W + 42%) _MFLPD
- Where, FRP = fraction of rated thermalpower MFLPD= maximum fraction of limiting power density where the limiting power density is 13.4 KR/ft for 8x8 and P8x8R fuel.
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 be2 9 in. on level instruments.
D.
Turbine stop valve closure scram setting shall be $; 10 percent valve closure.
E.
Turbine control valve fast closure setting shall be21 150 psig con-trol oil pressure at acceleration relay.
F.
Condenser low vacuum scram setting shall be jy 23 in. Hg. vacuum.
G.
Main steam isolation scram setting shall be $ 10 percent valve clo-sure.
8 1617 156
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i 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING H.
Main steam isolation on main steam line low pressure at inlet to turbine valves. Pressure setting shall be j> 880 psig.
1.
Reactor low-low water level initiation of CSCS systems set-ting shall be at or above -49 in.
indicated level.
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9 1617 158
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 5-2B, 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(2) and the basis for the uncertainty in the GEXL the uncertainties in the core parameters are given in NEDO 20340 correlation is given in NEDD-10958(1).
The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution in Pilgrim Nuclear Power Station Unit 1 during any fuel cycle would not be as severe as the distribution used in the analysis.
B.
Core Thermal Power Limit (Reactor Pressure < 803 psig or Core Flow
< 10% of Rated)
The use of the GEXL correlation is not valid for the critical power calculations at pressures below 800 psig or core flows less than 10% of rated.
Thereforc, 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 powerandallflowswillalwaysbegreagerthan4.56 psi.
Analyses show that with a flow of 28x10 lbs/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 3
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 0.35 MWt bundle power cor-responds 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.
1617 159 12
C.
Power Transient Plant safety analyses have shown that the scrams caused by ex-ceeding any safety setting will assure that the Safety Limit of Specification 1.lA or 1.lB will not be exceeded.
Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scrcm 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 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 cxists 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 level will be continuously monitored.
References General Electric Thermal Analysis Basis (GETAB): Data, 1.
Correlation and Design Application, General Electric Co.
BWR Systems Department, November 1973 (NEDO-10958).
Process Computer Performance Evaluation Accuracy, General 2.
Electric Company BWR Systems Department, June,1974 (NEDO-20340).
General Electric Boiling Water Reactor Generic Reload 3.
Fuel Application, NEDE-240ll-P.
1617 1'60 23 e
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2.1 BASES
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 pover density (MFLPD)and reactor core i
thermal power. The scram setting is adjusted in accordance l
with the formula in Specification 2.1. A.1 when the MFLPD is greater than the fraction of rated power (FRP).
Analyses of the limiting transients show that no scram adjustment is required to assure MCPR greater than the Safety Limit MCPR when the transient is initiated from MCPR above the operating l
limit MCPR.
For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated.
The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressute at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, 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 case of significant power rise.
Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally the heat flux is in the near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than five percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before power could exceed the safety limit. The 15% APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 880 psig.
The analysis to support operation at various power and flow re-lationships has considered operation with either one or two re-circulation pumps.
IRM The IRM system consists of 8 chambers, 4 in each of the reactor protection system logic channels.
The IRM is a 5-decade instrument which covers the range of power level
}h 7 b
16
2.1 BASES
between that covered by the SRM and the APRM.
The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size.
The IRM scram setting of 120/125 of full scale is active in each range of the IRM. For example, if the instrument were on range 1, the scram setting would be a 120/125 of full scale for that range; likewise, if the instrument were on range 5, the scram.would be 120/125 of full scale on that range.
Thus, as the IRM is ranged up to accommodate the increase in power level, the scram setting is also ranged up.
The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For in-sequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that heat flux is in equilibrium with the neutron flux, and an IRM scram would result in a reactor shutdown well before any safety limit is exceeded.
In order to ensure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale.
This condition exists at quarter rod density. Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed.
The results of this analysis show that the reactor is scrammed and peak core power limited to one percent of rated power, thus maintaining MCPR above the Safety Limit MCPR. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
B.
APRM Control Rod Block 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 prevent rod withdrawal beyond a given point at constant recirculation flow rate, and thus to protect against the condition of a MCPR less the Safety Limit MCPR.
This rod block set point, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal.
The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range.
The 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 107%
of rated thermal power because of the APRM rod block trip 1617 16,4 17
2.1 BASES
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 f raction of limiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin.
C.
Reactor Water Low Level Scram Trip Setting (LL1)
The set point for low level scram is above the bottom of the separator skirt.
This level has been used in transient analyses dealing with coolant inventory decrease. The results show that scram at this level adequately protects the fuel and the pressure barrier, because MCPR l
remains well above the safety limit MCPR in all cases, and syst3m 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 $ 10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the safety limit MCPR l
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 f rom 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.
1617 165 18
3.1 LIMITING CONDITION FOR OPERATION 4.1 SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM REACTOR PROTECTION SYSTEM Applicability:
Applicability:
Applies to the instrumentation Applies to the surveillance of i
the instrumentation and associ-l and associated devices which ated devices which initiate re-initiate a reactor scram, actor scram.
Objective:
Objective:
To assure the operability of the To specify the type and frequency of surveillance to be applied to reactor protection system, the protection instrumentation.
Specification:
Specification:
A.
Instrumentation systems shall The setpoints, minimum number of be functionally tested and trip systems, and minimum number calibrated as indicated in of instrument channels that must Tables 4.1.1 and 4.1.2 re-be operable for each position of the reactor mode switch shall be spectively.
as given in Table 3.1.1.
The B.
Daily during reactor power system response times from the operation, the maximum frac-opening of the sensor contact up tion of limiting power density to and including the opening of shall be checked and the scram the trip actuator contacts shall and APRM Rod Block settings not excead 100 milli-seconds.
given by equations in Specification 2.1. A.1 and 2.1.B shall be calculated if maximum fraction of limiting power density exceeds the fraction of rated power.
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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 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 M1b/br.
Trip level setting in percent of design power (1998 MWt).
15.
See Section 2.1.A.1.
16.
The APRM (15%) high flux scram is bypassed when in the run mode.
17.
The APRM flow biased high flux scram is bypassed when in the refuel or startup/ hot standby modes.
1617 168 i
29 i
4.1 BASES (Cont'd)
B.
The Maximum Fraction of Limiting Power Density (MFLPD) shall be checked once per day to determine if the APRM scram requires adjustment.
This will normally be done by c*iecking the LPKH readings. Only a small number of control rods are moved daily and thus the MFLPD is not expected to change significantly and thus a daily check of the MFLPD is adequate.
The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate.
This is compensated for in the APRM system by calibrating every three days using heat balance data and by calibrating individual LPRM's every 1000 effective full power hours using TIP traverse data.
1617 169 40
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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 IBM 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 condi-tion may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereaf ter; 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.
W is percent of drive flow required to produce a rated core flow of 2.
69 M 1b/hr. Trip level setting is in percent of design power (1998 MWt).
3.
IEM downscale is bypassed when it is on its lowest range.
4.
This function is bypassed when the count rate is 1100 cps.
5.
One of the four SRM inputs may be bypassed.
6.
This SRM function is bypassed when the IRM range switches are on range 8 or above.
7.
The trip is bypassed when the reactor power is <30%.
8.
This function is bypassed when the mode switch is placed in Run.
I617 171 s
55
3.2 BASES (Cont' d)
The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to the Safety Limit MCPR.
The trip logic for this function is 1 out of n:
e.g.,
any trip on one of six APRM's, eight IRM's, or four SRM's will result in a rod block.
The minimum instrument channel requirements assure sufficient instru-mentation to assure the single failure criteria is met.
The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration.
This time period is only 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.
The APRM rod block function is flow biased 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 MCPR.
The RBM rod block function provides local protection of the core, for a single rod withdrawal error from a limiting control rod pattern.
The IRM rod block function provides local as well as gross core pro-tection.
The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level.
A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.
The downscale trips are set at 2.5 indicated on scale.
The flow comparator and scram discharge volume high level components have only one logic channel and are not required for safety.
The refueling interlocks also operate one logic channel, and are re-quired for safety only when the mode switch is in the refueling position.
For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time.
The automatic pressure relief function is provided as a backup to the 1617 172 71
. - -... -.... -..... - ~. -.. -. - -. -. -
3.3 and 4.3 BASES:
During reactor operation with certain limiting control rod patterns, the withdrawal of a desig-nated single control rod could result in one or more fuel rods with MCPR's less than the Safety Limit MCPR. During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur. It is the responsibility of the Reactor Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns.
C.
Scram Insertion Times The control rod system is designed to bring the reac-tor subcritical at : rate fast enough to prevent fuel damage; i. e., to prevent the MCPR from becoming less than the Safety Limit MCPR. Analysis of the limiting power transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above Specification, provide the required protection, and MCPR remains greater than the Safety Limit MCPR.
i The scram times for all control rods will be deter-mined at the time of each refueling outage. A re-presentative sample of control rods will be scram tested during each cycle as a periodic check against deterioration of the control rod performance.
1617 173 91
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11 REACTOR FUEL ASSDiBLY 4.11 REACTOR FUEL ASSEMBLY Applicability Applicability The Limiting Conditions for Operation The surveillance Requirements associated with the fuel rods apply apply to the parameters which to those parameters which monitor the the fuel rod operating condi-fuel rod operating conditions.
tions.
Objective Objective The Objective of the Limiting Condi-The Objective of the Surveil-tions for Operation is to assure the lance Requirements is to performance of the fuel rods.
specify the type and frequency 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, the fuel as a function of average APLHGR for each type of fuel as a planar exposure shall be function of average planar exposure determined daily during shall not exceed the applicable reactor operation at ]t25%
limiting value shown in Figure rated thermal power.
3.11-1.
When core flow is less than 90% of rated core flow, the APLHGR shall not exceed 95%
of the limiting value showa in Figure 3.11-1.
If at any time during operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR is not re-turned 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.
1617 174 205A
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS B.
Linear Heat Generation Rate (LHGR)
B.
Linear Heat Generation Rate (LEGR)
During reactor power operation The LHGR as a function of core the linear heat generation rate (LHGR) height shall be checked daily of any rod in any fuel assembly at during reactor operation at j> 25%
any axial location shall not exceed rated thermal power.
13.4 kw/ft for 8x8 and P8x8R fuel.
If at any time during operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to reatore 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.
1617 175 205A-1
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS C.
Minimum Critical Power Ratio (MCPR)
C.
Minimum Critical Power Ratio (MCPR)
During power operation MCPR shall bej!
MCPR shall be determined daily 1.35 for 8x8 and P8x8R f uel.
If at during reactor power operation at any time during operation it is deter-
> 25% rated thermal power and mined by normal surveillance that the following any change in power limiting value for MCPR is being ex-level or distribution that would ceeded, action shall be initiated cause operation with a limiting within 15 minutes to restore operation control rod pattern as described to within the prescribed limits.
If in the bases for Specification the steady state MCPR is not returned 3.3.B.5.
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.
For core flows other than rated the MCPR shall be j> 1.35 for 8x8 and P8x8Rfuel times Kg, where Kg is as shown in Figure 3.11-8.
As an alternative method ?roviding equivalent thermal-hydraulic protec-tion at core flows other than rated, the calculated MCPR may be divided by K, where Kf is as shown in f
Figure 3.11-8.
1617 176 205B
BASES 3.llA Average Planar Linear Heat Generation Rate (APLHGR)
This specifications 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 the 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 results of the LOCA analysis for Pilgrm are given in Reference 2.
The maxi-mum allowable APLHGR valua for the present fuel types are presented in the Supplemental Reload Licensing Submittal for the current reload.
The operating limit MAPLHGR as a function of average planar exposure has been chosen conserva-tively below the lower envelope of the maximum allowable APLHGR values for all fuel types in the current reload.
1617 177 205C
3.11A BASES (Cont' d)
This provides operating limit MAPLHGR's which are fuel type independent for the current reload, and which also are fuel cycle independent as long as the maximum allowable MAPLHGR values for any future fuel type are bounded by the current operating limit MAPLHGRs. The operating limit MAPLHGR as a function of average planar exposure is shown in Figure 3.11-1.
t Reference 3 demonstrates that for lower initial core flow rates the potential exists for earlier DNB during postulated LOCA' s.
Therefore, a more restrictive limit for APLHGR is required during reduced flow conditions.
The ECCS analysis presented in Reference 4 assumed an initial MCPR of 1.24 for reduced flow conditions.
1617 178 205C-1 e
=w
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6 REFERENCES 1.
General Electric BWR Generic Reload Fuel Application, NEDE-240ll-P.
2.
Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NEDO-21696, August 1977.
3.
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566 (Draf t), August 1974.
1617 179 205C-2 0
en s.
s
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BASES:
- 3. llc MINIMUM CRITICAL POWER RATIO (MCPR)
Operating Limit MCPR For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any tLae during the transient assuming instrument trip setting given in Specification 2.1.
The required operating limit MCPR at steady state conditions in Specification 3.11.C was chosen conservctively at a value higher than MCPR's of past analysis with the objective of establishing an operating limit MCPR which is fuel type and cycle independant.
The difference between the specifie Operating Limit MCPR in d
Specification 3. llc and the Safety Limit MCPR in Specification 1.lA defines the largest reduction in critical power ratio (CPR) permitted during any anticipated abnormal operating transient.
To ensure that this reduction is not exceeded, the most limiting transients are analized for each reload and fuel type (8x8 and P8x8R) to determine that transient which yields the largest value of Ji CPR. This value, when added to the Safety Limit MCPR must be less than the minimum operating limit MCPR's of Specification i
3.11.C.
The result of this evaluation is documented in the
" Supplemental Reload Licensing Submittal" for the current reload.
parametersshowninTables5-4,5-6and5-8ofNEDE-240ll-P{gyut The evaluation of a given transient begins with the system Supplemented by reload unique inputs given in the current Supplemental Reload Licensing Submittal. These values are input to a GE core dp)amic behavior transient computer program described
~
in NEDO-10802(4 Also, the void reactivity coefficients that were input to the transient calculational procedure are based on a new mthod of calculation termed NEV which provides a better agreement between the calculated and plant instrument power distributions. The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic SCAT code described in NEDE-20566(3). The principal result of this evaluation is the reduction in MCPR caused by the transient.
1617 180 e
s Two codes are used to analyze the rod withdrawal error transient, The first code simulates the three dimensional BWR core nuclear and thermal-hydraulic characteristics. Using this code a limiting control rod pattern is determined; the following assumptions are included in this determination:
(1) The core is operating at full power in the xenon-free condition.
(2) The highest worth control rod is assumed to be fully inserted.
(3) The analysis is performed for the most reactive point in the cycle.
(4) The control rods are assumed to be the worst possible pattern without exceeding thermal limits.
(5) A bundle in the vicinity of the highest worth control rod is assumed to be operating at the maximum allowable linear heat generation rate.
(6) A bundle in the vicinity of the highest worth control rod is assumed to be operating,the minimum allowable critical power ratio.
The three-dimensional BWR code then simulates the core response to the control rod withdrawal error.
The second code calculates the Rod Block Monitor response to the rod withdrawal error.
This code simulates the Rod Block Monitor under selected failure conditions (LPRM) for the core rod use (calculated by the 3-dimensional BWR simulation code) for the control rod withdrawal.
1617 181 205C-4
-. ~
-.. ~.
e The analysis of the rod withdrawal error for Pilgrim Unit 1 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-240ll-P.
MCPR LIMITS FOR CORE FIDWS OTHER TRAN RATED The purpose of the Kf 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 factor provides and the Kg f actor. Specifically, the Kf the required thermal margin to protect against a flow in-crease transient.
The most limiting transient initiated from less than 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 g
assure that the operating limit MCPR given in Specification 3. llc will not be violated should the most limiting transient occur at less than rated flow.
In the manual flow control mode, the Kg f actors assure that the Safety Limit MCPR will not be violated for the same postulated transient event.
The K f actor curves shown in Figure 3.11-8(4) were developed generkcallywhichareapplicabletoallBWR/2,BWR/3,andBWR/4 reactors.
The Kr f actors were derived using the flow control line corresponding to rated thermal power at rated core flow.
For the manual flow control mode, the Kf 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 Kr.
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.
1617 182 205C-5
6-The K factors shown in Figure 3.11-8(4) are conservative for the Pilgrkm Unit 1 operation because the operating limit MCPR given in Specification 3. llc is greater than the original 1.20 operating limit MCPR used for the generic derivation of Kg.
4.ll.C MINIMUM CRITICAL POWER RATIO (MCPR) - SURVEILLANCE REQUIREMENT At core thermal power levels less thea or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level with minimum recirculation pump speed.
The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily re-quirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been.significant power or control rod changes.
The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.
I617 183 205C-6
REFERENCES 1.
General Electric BWR Generic Reload Fuel Application, NEDE-240ll-P.
2.
R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NEDO-10802).
3.
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566 (Draft), August 1974.
4.
Letter from J. E. Howard, Boston Edison Company to D. L. Ziemann USNRC, dated October 31, 1975.
1617 184 i
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- F IG U R E 3.11 -1 MAXIMUM AV ERAGE PLANAR LINE AR HE AT GENER ATION RATE VERSUS PLANAR AVERAGE EXPOSURE 8x8 AND P8X8R FUEL 7
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5.0 MAJOR DESIGN FEATURES 5.1 SITE FEATURES Pilgrim Nuclear Power Station is located on the Western Shore of Cape Cod Bay in the Town of Plymouth, Plymouth County, Massachu-setts. The site is located at approximately 41 51' north latitude and 70035' west longitude on the Manomet Quadrangle, Massachusetts, Plymouth County 7.5 Minutu Series (topographic) map issued by U.S.
Geological Survey. UTM coordinates are 19-46446N-3692E.
The reactor (center line) is located approximately 1800 feet from the nearest property boundary.
5.2 REACTOR A.
The core shall consist of not more than 580 fuel assemblies of 8x8 (63 fuel rods) and P8x8R (62 fuel rods).
B.
The reactor core shall contain 145 cruciform-shaped control rods. The control material shall be boron carbide powder (B C) compacted to approximately 70% of theoretical density.
4 5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2.2 of the FSAR.
The applicable design codes shall be as described in Table 4.2.1 ef the FSAR.
5.4 CONTAINMENT A.
The principal design parameters for the primary containment shall be as given in Table 5.2.1 of the FSAR. The applicable design codes shall be as described in Section 12.2.2.8 of the FSAR.
B.
The secondary containment shall be as described in Section 5.3.2 of the FSAR.
C.
Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.
1617 187 206m
.