ML19309F781

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Amend 59 to License DPR-49,changing Tech Specs to Include Prepressurized 8 X 8 Retrofit Fuel,Revising Operating Limit Min Critical Power Ratios & Incorporating Administrative Improvements.Revised Tech Spec Pages Encl
ML19309F781
Person / Time
Site: Duane Arnold NextEra Energy icon.png
Issue date: 04/10/1980
From: Ippolito T
Office of Nuclear Reactor Regulation
To:
Shared Package
ML19309F782 List:
References
NUDOCS 8005010206
Download: ML19309F781 (58)
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{{#Wiki_filter:(.G p mas d o, UNITED STATES 8 NUCLEAR REGULATORY COMMISS;ON 3 E WASHINGTON, D. C. 20655 8 s,...../ IOWA ELECTRIC LIGHT AND POWER COMPA E CENTRAL IOWA POWER C00PERATIVr, CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 59 License No. DPR-49 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Iowa Electric Light and Power Company, Central Iowa Power Cooperative, and Corn Belt Power Cooperative (the licensee) dated January 22, 1980 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the roles and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of tha public; and E. The issuance of this amendment is in accordance with in CFR Part 51 of the Connission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 59, are hereby incorporated in the l license. 'The licensee shall operate the facility in accordance l with the Technical Specifications. 8005010 1

l l l 3. This license amendmelt is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION -. s <M, Thomas Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: April 10, 1980 1 l \\ I I

ATTACHMENT TO LICENSE AMENDMENT NO. 59 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 "aplace the following pages of the Appendix "A" Technical Specifications with the enclosed identically numbered pages. The revised pages are identified by Anendment number and contain vertical lines indicating the area of change. 3.1-4 y 3.1-7 yj l 3.1-10 yjj

  • 3.1-27/3.1-28 1.0-11 3.2-6

,'f_^" 3.2-7/3.2-8* 3.2-15 1.1-2 3.2-16 1.1-3 3.2-17 1.1-5

  • 3.2-35/3.2-36 1.1-6 3.2-42 1.1-7 3.3-9 1.1-8 3.3-19 1,j_g
  • 3 4-3/3.4-4 1.1-10 3.6-24 1.1-11 3.12-1 1.1-13 3.12-2 1.1-14 3.12-4 1.1-15 3.12-5 1.1-17 3.12-6 1.1-19 3.12-9 1.1-22 3.12-9a 1.1-24 3.12-11 3.1-1/3.1-2*

3.1-3 Add pages: 3.2-36a 3.12-5a 3.12-18

  • No change, provided for convenience l

OAEC-1 TECHNICAL SPECIFICATIONS LIST OF TABLES Table Number Title Page 1.1-1 Deleted 1.1-2 Deleted 1.1-4 Deleted 3.1-1 Reactor Protection System (SCRAM) Instrumentation 3.1-3 Requirements 4.1-1 Reactor Protection System (SCRAM) Instrument 3.1-8 Functional Tests 4.1-2 Re,4ctor Protection System (SCRAM) Instrument 3.1-12 Calibration 3.2-A Instrumentation that Initiates Primary Containment 3.2-5 Isolation 3.2-8 Instrumentation that Initiates or Controls the Core 3.2-8 and Containment Spray Systems 3.2-C Instrumentation that Initiates Control Rod Blocks 3.2-16 3.2-0 Radiation Monitoring Systems that Initiate and/or 3.2-19 Isolate Systems 3.2-E Instrumentation that Monitors Drywell Leak Detection 3.2-20 3.2-F Sur-veillance Instrumentation 3.2-21 3.2-G Instrumentation that Initiates Recirculation Pump Trip 3.2-23 4.2-A Minimum Test and Calibration Frequency for PCIS 3.2-24 4.2-8 Minimum Test and Calibration Frequency for CSCS 3.2-26 4.2-C Minimum Test and Calibration Frequency for Co5 trol 3.2-38 Rod Blocks Actuation Amendment No. 59 y

DAEC-1 s Table Number Title Page 4.2-0 Minimum Test and Calibration Frequency for Radiation 3.2-29 Monitoring Systems 4.2-E Minimum Test Calibration Frequency for Drywell 3.2-30 Leak Detection 4.2-F Minimum Test Calibration Frequency for Surveillance 3.2-31 Instrumentation

4. 2-G Minimum Test and Calibration Frequency for 3.2-34 Recirculation Pump Trip 3.6-1 Number of Specimens by Source 3.6-33 4.6-3 Snubbers Accessible Ouring Normal Operation 3.6-41 4.6-4 Snubbers Inaccessible Ouring Normal Operation 3.6-43 4.6-5 Snubbers in High Radiation Area During Shudown and/or 3.6-44 Especially Diffigult to Remove 3.7-1 Containment Penetrations Subject to Type "B" Test 3.7-20 Requirements 3.7-2 containment Isolation Valves Subject to Type "C"

3.7-22 Test Requirements 3.7-3 Primary Containment Power Operated Isolation Valves 3.7-25 3.12-1 Deleted l 3.12-2 MCPR Limits 3.12-9a 3.13-1 Fire Detection Instruments 3.13-11 3.13-2 ' Required Fire Hose Stations 3.13-12 1

6. 2-1 Minimum Shift Crew Personnel and License 6.2-3 Requirements 6.9-1 Protection Factors for Respirators 6.9-8 6.11-1 Reporting Summary - Routine Reports 6.11-12 6.11-2 Reporting Summary - Non-routine Reports 6.11-14 l

Anendnent No. 59 g l

l OAEC-1 l l l TECHNICAL SPECIFICATIONS l LIST OF FIGURES Figure l Number Title l 1.1-1 Power / Flow Map l l 1.1-2 Deleted 2.1-1 APRM Flow Biased Scram and Rod Blocks 2.1-2 Deleted 4.1-1 Instrument Test Interval Determination Curves 4.2-2 Probability of System Unavailability vs. Test Interval 3.4-1 Sodium Pentaborate Solution Volume Concentration Requirements 3.4-2 Saturation Temperature of Sodium Pentaborate Solution 3.6-1 OAEC,0perating Limits 6.2-1 OAEC Nuclear Plant Staffing 3.12-1 K as a Function of Core Flow 7 3.12-2 Limiting Average Planar Linear Heat Generation Rate (Fuel Type 2) 3.12-3 Limiting Average Planar Linear Heat Generation Rate (Fuel Type 3) 3.12-4 Limiting Average Planar Linear Heat Generation Rate (Fuel Type 4, 70230) 3.12-5 Limiting Average Planar Linear Heat Generation Rate (Fuel Type 80274L) 3.12-6 Limiting Average Planar Linear Heat Generation Rate i (Fuel Type 80274H) 3.12-7 Limiting Average Planar Linear Heat Generation Rate c j (Fuel Type P80P8289) l l i Amendment *10. 59 Yk

DAEC-1 21. Thermal Parameters a. Minimum Critical Power Ratio (MCPR) - The value of critical power ratio (CPR) for that fuel bundle having the lowest CPR. b. Critical Power Ratio (CPR) - The ratio of that fuel bundle power which would produce boiling transition to the actual fuel bundle power. Transition Boiling - Transition boiling means the boiling c. regime between nucleate and film boiling. Transition boiling is the regi.ne in which both nucleate and film boiling occur intermittently with neither type being completely stable. d. Deleted. e. Linear Heat Generation Rate - the heat output per unit length of fuel pin. f. Fraction of Limiting Power Density (FLPD) - The fraction of limiting power density is the ratio of the linear heat genera-tion rate (LHGR) existing at a given location to the design LHGR for that bundle type. g. Maximum Fraction of Limiting Power Oensity (MFLPD) - The maximum fraction of limiting power density is the highest value existing in the core of the fraction of limiting power density (FLPD). h. Fraction of Rated Power (FRP) - The fraction of rated power is the ratio of core thermal power to cated thermal power of 1593 MWth. Anendnent No. 59 L

DAEC-1 l l l 22. Instrumentation a. Instrument Calibration - An instrument calibration means l the verification or adjustment of an instrument signal out-put so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument. monitors. The acceptable range and accuracy of 1 an instrument and its setpoint are given in the system design control document and these setpoints are used in the Technical Specifications. l l i i l Amendment No. 59 1, o.11, a i ~-

DAEC-1 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLA00ING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability: Applicability: Applies to the inter-related Appifes to trip settings of variables associated with the instruments and devices fuel thermal behavior, which are provided to prevent the reactor system safety limits from being exceeded. Objective: Objective: To establish limits which To define the level of the ensure the integrity of the process variables at which fuel cladding. automatic protective action is initiated to prevent the fuel cladding integrity safety limits from being exceeded. Soecifications: Specifications: The limiting safety system settings shall be as speci-fied below: A. Reactor Pressure > 785 psig A. Neutron Flux Trips and Core Flow > 10% of Rated. 1. APRM High Flux Scram The existence of a minimum When In Run Mode, critical power ratio (MCPR) less j than 1.07 shall constitute For operation with the violation of the fuel cladding fraction of rated power integrity safety limit. (FRP) greater than or equal to the maximum 8. Core Thermal Power Limit (Reactor fraction of limiting Pressure < 785 osic or Core Flow power density (MFLPD), < 10% of Rated the APRM scram trip set-point shall be as shown 2 When the reactor pressure is < 785 on Fig. 2.1-1 and shall be: psig or core flow is less than 10% of rated, the core thermal power S< (0.66W + 54) shall not exceed 25 percent of rated thermal power, with a maximum setpoint of 120% rated power at 100% rated recirculation flow or greater. j 1.1-1 Anendnent No. 59 l \\

DAEC-1 SAFETY LIMIT 8.IMITING SAFETY SYSTEM SETTING 16 C. Power Transient _ Where: 5 = Setting in percent of rated power To ensure that the Safety (1,593 MWt) Limits established in Speci-fication 1.1.A and 1.1.8 are W = Recirculation loop not exceeded, each required flow in percent of scram shall be init'iated by rated flow. Rated its primary source signal. A recirculation loop Safety Limit shall be assumed flow is that recircu-to be exceeded when scram is lation loop ficu which accomplished by a means other-corresponds to 49x10s other than the Primary Source ib/hr core flow. Signal. D. With irradiated fuel in the reac-tor vessel, the water level shall not be less than 12 in. above the For a MFLPD greater than FRP, l top of the normal active fuel the APRM scram setpoint shall zone. Top of the active fuel be: ) zone is defined to be 344.5 inches above vessel zero (see Bases 3.2). 51(0.66W+54)MfLfD NOTE: These settings assume operation within the basic thermal design criteria. These criteria are LNGR < 18.5 KW/ft (7x7 array) or 13.4 KW/ft (8x8 array) and MCPR > values as indicated in Table 3.12-2 times K, 7 where K,is defined by Figure 3.12-1. Therefore, at full power, operation is not allowed with MFLPD greater than unity even if the scram setting is reduced. If it is a determined that either of these design criteria is being vio-lated during operation, action must be taken immediately to return to operation within these criteria. 2. APRM High Flux Scram When in the REFUEL or STARTUP and HOT STANOBY MODE. The APRM scram shall be set at less than or equal to 15 percent of rated power. ( Anendnent No. 59 1,12

DAEC-1 LIMITING SAFETY SYSTEM SETTING SAFETY LIMIT 3. APRM Rod Block When in Run Mode. For operation with MFLPD less than or equal to FRP the APRM Control Red Block setpoint shall be as shown on Fig. 2.1-1 and shall be-5 5 (0.66 W + 42) The definitions used above for the APRM scram trip apply. For a NFLPD greater than FRP, the APRM Control Rod Block setpoint shall be: S $ (0.66 W + 42) g"FLPD 4. IRM - The IRM scram shall be set at less than or equal to 120/125 of full scale. ,513.5 B. Scram and Iso- ) lation on inches reactor low above water level vessel zero (+12" on level instru-ments) C. Scram - turbine 1 10 percent stop valve valve closure closure D. Turbine control valve' fast closure shall occur within 30 milliseends of the start of turbine control valve fast closure. l l Amendment No. 59 1.1-3 l

DAEC-1.

1.1 SASES

FUEL CLADDING INTEGRITY A. Fuel Cladding Integrity Limit at Reactor Pressure 1 785 psig and Core Flow 1 10% of Rated The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated.. Since the parameters which result in fuel damage are not directly observable during reactor operation the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedure used to calculate the critical power result in an uncertainty in the value of, the critical power. Therefore the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. The Safety Limit MCPR is generically determined in Reference 1. l l Amendment No. 59 1.1-5 l s

DAEC-1 e 1 l I DTL7ED i 1.1-6 Amendment No. 59 t' ^ - - - - - ~ - r-.

E DAEC-1 B. Core Thermal Power Limit (Reactor Pressure 4 785 psig or Core Flow 6 10% of Rated) At pressures t Jow 785 psig, the core evaluation pressure drop (0 power, 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 pressure drop at low power and all flows will always be greater than 3 4.56 psi. Analyses show that with a flow of 28 x 101bs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be 3 greater than 28 x 101bs/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 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. C. Power Transient Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 1.1.A or 1.1.B will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following close 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. Amendment No. 59 1,j_7

DAEC-1 The computer provided with Ouane Arnold 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 informa-tion 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, Specifica-tion 1.1.C will be relied on to determine if a Safety Limit has been violated. D. Reactor Water Level (Shutdown Condition) During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water levei 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 perforatian. 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.

  • Top of the active fuel zone is defined to be 344.5 inches above vessel zero (See Bases 3.2).

1.1-8 Amendment No. 59

DAEC-1

1.1 REFERENCES

1 1 I 1. " Generic Reload Fuel Application," NEDE-24011-P-A and NE00-24011-A.* r i

  • Approved Revision at time reload analyses are performed.

1.1-9 Anendment No. 59

DAEC-1. I t l i i DELETED 4 c l

1. 1-10 Anendment No. 59 1

DAEC-1 l l OELETED. l t t 1.1-11 Anondnent No. 59

~ ~ -~ DAEC-1 l j l l OELETED i l l l 1.1-13 Amendment No. 59 l l 48

DAEC-1

2.1 BASES

LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INTEGRITY The abnormal operational transients appifcable to operation of the Quane Arnold Fr.ergy Center have been analyzed throughout the spectrum of pl.nnea operating conditions up. to the thermal power condition of 1658 MWt. The analyses were based upon plant operation in accord-ance with the operating map given in Figure 3.7-1 of the FSAR. In t f addition, 1658-MWt is the licensed maximum power level of the Ouane Arnold Energy Center, and this represents the maximum steady state power which shall not knowingly be exceeded. Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis mode. Conservatisms incorporated into the transient analysis is documented in Reference 1. The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to.,e about 25% greater than the nominal maximum absoluta value expected to occur during the core lifetime. The scram worth used has been derated to be equiva-lI lent to approximately 80% of the total scram worth of the control I rods. The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifications. The Amendment No. 59

1 OAEC-1 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 reactivity is assured by the time requirements for 25.6% and 46.4% insertion. By the time the rods are 67.2% inserted, l approximately four dollars of negative reactivity have been inserted which strongly turns the transient, and accomplishes the desired effect. The times for 4.7% and 88.1% insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition. For analyses of the thermal consequences of the transients the MCPRs stated in section.l.12 as a limiting condition of operation bound those which are conservatively assumed to exist prior to initiation of the transielts. This choice of using cont,ervative values of controlling parameters and initiating transients at the design power level produces more conservative results than would be obtained by using expected values of control parameters and analyzing at higher power levels. Steady-state operation without forced recirculation will not be permitted, except during special tasting. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps, j l In summary: 1. The abnormal operational transients have been analyzed to a power level of 1658 MWt. ii. The licensed maximum power level is 1658 MWt. iii. Analyses of transients employ adequately conservative values of the controlling reactor parameters.

1. 1-15 Amendment No. 59

DAEC-1 during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cl' adding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams. The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of MFLPD and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1.A.1, when the maximum fraction of limiting power density is greater than the fraction of rated power. Analyses of the limiting transients show that no scram adjustment is required to assure MCPR greater than or equal to safety limit when the transient is initiated from MCPR > values as indicated in Table 3.12.2. APRM High Flux Scram (Refuel or Startup & Hot Standby Mode) 2. For operation in these modes the APRM scram setting of 15 percent of rated power and the IRM High Flux Scram provide adequate thermal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accommodate anticipated saneuvers 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 system. f l temperature coefficients are small, and control rod patterns are con-strained to be uniform by operating procedures backed up by the rod I f

1. 1-17 Amendment No. 59 i

DAEC-1 l l l l as the flow decreases for the specified trip setting versus flow relationship; therefore the worst case MCPR which could occur l 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. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted' downward if the maximum-fraction of limiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gain. 4. IRM The IRM system consists of 6 chambers., 3 in each of the reactor ) protection system logic channels. The IRM is a 5-decade instrument l which covers the range of power level 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 trip set-For ting of 120 divisions is active in each range of the IRM. example, if the instrument were on range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accom-modate the increase in power level, the scram trip setting is also ranged up. The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that the 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. l l i

1. 1-19 Amendnent flo. 59

DAEC-1 l that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase. To protect the main condenser against over-pressure, a loss of condenser vacuum initiates automatic closure of the main steam isolation valves. 1 G. H. and I. Reactor Low Water Level Setpoint for Initiation of HPCI l and RCIC, Closing Main Steam Isolation Valves, and Starting LPCI and Core Spray Pumps l These systems maintain adequate coolant inventory and provide core l cooling with the objective of preventing excessive clad tempera-l tures. The design of these systems to adequately perform the intended function is based on the specified low level scram set-16 point and initiation setpoints. Transient analyses demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.

2.1 REFERENCES

1. " Generic Reload Fuel Application," NEDE-24011-P-A* or NEDO-24011-A.

  • Approved revision number at time analyses are performed.

Amendment No. 59

l l l l 1 1 l i DELETED i } i I t l l i 1 1.1-24 Amendment No. 59

s LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability: Applicability: Applies to the instrumentation Applies to the surveillance of the and associated devices which instrumentation and associated devices initiate a reactor scram. which initiate reactor scram. Objective: Objective: To assure the operability of the To specify the type and frequency of reactor protection system. surveillance -to be applied to the protection instrumentation. Specification: Specification: The setpoints, minimum number of trip systems, and minimum number A.1 Instrumentation systems shall be of instrument channels that must functionally tested and calibrated as be operable for each position of indicated in Tables 4.1-1 and 4.1-2 the reactor mode switch shall be respectively. as given in Table 3.1-1. The designed system response times 2.

Response

time measurements (from from the opening of the sensor actuation of sensor contacts er trip contact up to and including the point to de-energization of scram opening of the trip actuator solenoid relay) are not part of the contacts shall not exceed 50 normal instrument calibration. The milli-seconds. measurement of response time will be performed once per operating cycle. B. Daily during reactor power operation, the MFLPD and the FRP shall be checked and the APRM SCRAM and APRM Rod Block settings given by equations in Specification 2.1. A.1 and 2.1.B shall J be calculated if the MFLPD exceeds the 1 FRP. l l;. l 3

  • 1~1 Amendment No. 59 l

l

o LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS C. When it is dete rmined that a channel has failed in the unsafe condition, the other RPS channels that mon-itor the same variable shall be functionally tested immediately before the trip system contain-ing the failure is tripped. The trip system containing the unsafe failure may be placed in the untripped con-j dition during the period in which sur-veillance testing is being performed on the other RPS channels. The trip system may be in the untripped position for no more than eight hours per functional trip period for this testing. G. I ai l 3.1-2 {

TABLE 3.1-1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Modes in Which Number of Minfaus No. 2, Function Must be Instrument g of Operable Operable Channels ![ Instrument Provided g fer Trip Trip Level Refuel Startup Run By Design Action (1) Channels System (1) Trip function Setting (6) e X X X 1 Mode Switch A ES 1 Mode Switch in (4 Sections) Shutdown X X X 2 Instrument A 1 Manual Scram Channels 2 IRM High Flux <120/125 of Fuel X X (5) 6 Instrument A Channels Scale X X (5) 6 Instrument A l~ 2 IRM Inoperative Channels 2 'APRM High Flux (.66W+54) (FRP/MFLPD) X 6 Instrument A or 8 Channels (11) (12) 2 APRM Inoperative (10) X X X 6 Instrument A or 8 Channels 2 APRM Downscale > 5 Indicated on Scale (9) 6 Instrument A or B-Channels 2 APRM tiigh Flux in $ 15% Power X X 6 Instrument A Channels Startup 2 High Reactor Pressure $1035 psig X(8) X X 4 Instrument A Channels

1 TABLE 3.1-1 (Continued) ~ REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT sF i g i Modes in Which Number of @ Minimus No. function Must be Instrument of Operable Operable Channels if Instrument Provided Channels g; for Trip Trip Level Refuel Startup Run By Design Action (1) System (1) Trip function Setting (6), 2 High Drywell Pressure 12.0 psig X(7) X(8) X 4 Instrument A Channels 2 Reactor Low Water >+12" Indicated X X X 4 Instrument A Level Level Channels 2 High Water Level 5 60 Gallons X(2) X X 4 Instrument A Channels in Scram Discharge u, Volume 2 Main Steam Line 53 x Normal Rated X X X 4 Instrument A High Radiation Power Background

  • Channels 4

Main Steam Line 5 10% Valve Closure X X X (13) 8 Instrument A or C Isolation Valve (3)(13) (3)(13) Channels Closure 2 Turbine Control Within 30 milliseconds X(4) 4 Instrument A or D Valve fast of the start of Control Channels Closure (Loss of Valve fast Closure Control Oil Pressure) 1 4 Turbine Stop $10% Valve X(4) 8 Instrument A or D valve Closure Closure Channels 2 first Stage Bypass below 192 psig X X X 4 Instrument A or D Pressure Permissive at shell Channels " Alarm setting 11.5 X Normal Rated Power Background

DAEC-1 7. Not reqi.4 red to be operable when primary containment integrity is not required. 8. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel. 9. The APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not high. 10. To be considered operable, APRM's A, B, C and 0 must have at least 9 LPRM inputs while APRM's E and F must have at. least 13 LRPM inputs. Additionally each APRM must have at least 2 LPRM inputs per level. 11. W is the recirculation loop flow in percent of rated. 12. See Subsection 2.1.A.1. 13. The design permits closure of any two lines without a scram being initiated. 14. Deleted. 3.1-7 l Anendnent No. 59

DAEC-1 NOTES FOR TABLE 4.1-1 1.' Initially once every month, the compilation of instrument failure rate data say include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of DAEC. The failure rate data must be reviewed and approved by the NRC prior to any change in tha once-a-month frequency. 2. A description of the three groups is included in the Bases of this Specification. 3. Functional tests are not required on the part of the system that is not required to be operable or are tripped. If tests are missed on parts not required to be operable or are tripped, then they shall be performed prior to returning the system to an operable status. 4. This instrumentation is exempted from the instrument channel test definition. This instrument channel functional test will consist i of injecting a simulated electrical signal into the measurement channels. 3.1-10 Anendment No. 59 l

1 DAEC-1 i i the calibration is being performed, a zero flow signal will be sent to half of the APRM's resulting in a half scram and rod block condition. Thus, if the calibration were performed during operation, flux shaping would not be possible. Based on experience at other generating stations, drift of instru-ments, such as those in the Flow Biasing Network, is not-significant and therefore, to avoid spurious scrams, a cali-bration frequency of each refueling outage is established. Group 3 devices are active only during a given portion of the operational cycle. For example, the IRM is active during i startup and inactive during full-power operation. Thus, the only test that is meaningful is the one performed just prior t i to shutdown or startup; i.e., the tests that are performed just prior to use of the instrument. Calibration frequency of the instrument channel is divided into two groups. These are as follows: 1. Passive type indicating devices that can be compared with like units on a continuous basis. 2. Vacuum tube or semi-conductor devices and detectors that l l drif t or lose sensitivity. I 3.1-27 l l i /

DAEC-1. 1 Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate. For those devices which employ amplifiers, etc.,. drift specifications call for drift to be less than 0.4%/ month; i.e., in the period of a month a maximum drift of 0.4% could occur, thus providing for adequate margin. For the APRM system, drift of electronic apparatus is not the only consideration in determining a calibration frequency. Change in power distribution and loss of chamber sensitivity dictate a calibra-tion every seven days. Calibration on this frequency assures plant operation at or below thermal limits. A comparison of Tables 4.1-1 and 4.1-2 indicates that two instrument channels have not been included in the latter table. These are: mode switch in shutdown and manual scram. All of the devices or sensors associated with these scram functions are simple on-off switches and, hence, calibration during operation is not applicable. 2. The peak heat flux is checked once per day to determine if the APRM scram requires adjustment. This will normally be done by checking the LPRM readings. Only a small nur.ber of control rods are moved daily and thus the power distribution is ( ~ Amendment No. 59 g

DAEC-1 NOTES FOR TABLE 3.2-A 1. Whenever Primary Containment integrity is required by Subsection 3.7, there shall be two operable or tripped trip systems for each l function. 2. If the first colusin cannot be met for one of the trip systems, that trip system shall be tripped or the appropriate action listed below shall be taken. A. Initiate an orderly shutdown and have the reactor in Cold Shutdown Condition in 24 hours. B. Initiate an orderly load reduction and have Main Steam Lines isolated within eight hours. C. Isolate Shutdown Cooling. l 0. Isolate Reactor Water Cleanup System. 3. Instrument setpoint corresponds' to 170" above top of active fuel.* i l 4. Instrument setpoint corresponds to 119.5" above top of active fuel.* l " Top of the active fuel zone is defined to be 344.5 inches above vessel zero (see Bases 3.2). 3.2-6 Anendnent. No. 59 i 1

DAEC-1 4 5. Two required for each steam line. 6. These signals also start 58GTS and initiate secondary containment isolation. ? 7. Only required in Run Mode (interlocked with Mode Switch). 8. Deleted i l t 3.2-7 Amendment No. 58

nn-e TABLE 3.2-B INSTRUM.ENTATION T!! AT INITIATES OR CONTROLS Tile CORE AND CONTAINMENT COOLING SYSTEMS Minimum No. of Operable Instrument Number of Channels Per Instrument Channels Trip System (1) Trip Function Trip Level Setting Provided by Design Remarks 2 Reactor Low-Low > -38.5 in. indicated 4 HPCI & RCIC Initiates HPCI & RCIC Water Level level Inst. Channels 2 Reactor Low-Low-Low > -139.5 in. indicated 4 Core Spray & RHR

1. In conjunction with Water Level level (4)

Instrument Channels Low Reactor Pressure initiates Core Spray 4 ADS Instrument and LPCI Channels 2. In conjunction with O confirmatory low level High Drywell O N Pressure, 120 second [ [ time delay and LPCI or Core Spray pump interlock initiates Auto Blowdown (ADS) 3. Initiates starting of Diesel Generator 2 Reactor High Water i +53 in. indicated 2 Inst. Channels Trips HPCI and RCIC Level level turbines ~ l e .?%

DAEC-1 NOTES FOR TABLE 3.2-B 1. Whenever any CSCS subsystem is required by Subsection 3.5 to be operable, there shall be two operable trip systems. If the first column cannot be met for one of the trip systems, trip system shall be placed in the tripped condition or that the reactor shall be placed in the Cold Shutdown Condition within 24 hours. i 2. Close isolation valves in RCIC subsystem. 3. Close isolation valves in HPCI subsystem. 4. Instrument setpoint corresponds to 18.5" above the top of active fuel.

  • 5.

IIPCI has only one trip systen for these sensors. 6. The relay drop-out voltage will be measured once per operating cycle and the data examined for evidence of relay deterioration. 7. Four undervoltage relays with integral timers per 4 KV bus. The relay output contacts are connected to form a one-out-of-two-twice coinci-dent logic matrix. With one relay inoperable, operation may proceed pro-vided that the inoperable relay is placed in the tripped condition within one hour.

  • Top of active fuel zone is defined to be 344.5 inches above vessel zero (see Bases 3.2).

i Amendment No. 59 3.2-15 l i

TABLE 3.2-C fMinimumNo. g of Operable Number of g Instrument Provided by Design Action Instrument Channels ,Ch:nnels Per Trip Level Setting 3 Trip System Instrument 2 ? 6 Inst. Channels W E 2 APRM Upscale (Flow Biased) 1(0.66W + 42) (M PD) 2 APRM Upscale (Hot in Run 112 indicated on scale 6 Inst. Channels (1) Mode) 2 APRM Downscale 15 indicated on scale 6 Inst. Channels (1) 1 (7) Rod Block Monitor 1(0.66W + 39) ( LPD}( ) (Flow Blased) 1 (7) Rod Block Monitor 15 indicated on scale 2 Inst. Channels (1) Downscale w 2 IRM Downscale (3) 15/125 full scale 6 Inst. Channels (1) .L 2 IRH Detector not in (8) 6 Inst. Channels (1) Startup Position 2 IRM Upscale $108/125 6 Inst. Channels (1) 2 (5) SRM Detector not in (4) 4 Inst. Channels (1) Startup Position 2 (5) (6) SRM Upscale 110 counts /sec. 4 Inst. Channels (1) 5 l I i l-

DAEC-1 l NOTES FOR TABLE 3.2-C 1. For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function. The SRM and IRM blocks need not be operable in "Run" mode, and the APRM (except for APRM Upscale (Not in Run Mode)] and RBM rod blocks need not be operable in "Startup" mode. If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately I and daily thereafter; if.this condition lasts lor,ger than seVen days, the system shall be tripped. If the first column cannot be met for both trip systems, the systems shall be tripped. 2. W is the recirculation loop flow in percent of design. Trip level setting is in percent of rated power (1593 MWt). Refer to Limiting Safety System Settings for variation with MFLP0 & FRP (applicable only when MFLP0 exceeds FRP). 3. IRM downscale is bypassed when it is on its lowest range. 4. This function is bypassed when the count rate is >100 cps. l l 3.2-17 Amendment No. 59 i (

DAEC-1 3.2 BASES In additicn to reactor protection instru' mentation which ini-tiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation func-tion, initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the Specifications are: 1. To assure the effectiveness of the protective instru-mentation when required even during periods when portions of such systems are out of service for maintenance. 2. To prescribe the trip settings required to assure adequate performance. When necessary, one channel may be made i.ioperable for brief intervals to conduct required fur.ctional tests and calibrations. Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances 3.2-35 t 9

DAEC-1 explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The setpoints of other instru-mentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations. Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2-A which senses the conditions for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required. The instrumentation which initiates prima,ry system isolation is connected in a dual bus arrangement. Many of the reactor water level trip settings are defined or described in terms of " inches above the top of the active fuel." Because the reload fuel is longer than the initial core, the core is now composed of fuel bundles of differing lengths and the term " top of the active fuel" no longer has a definite meaning. Since the basis of all safety analyses i is the absolute level (inches above vessel zero) of the trip settings, the " top of the active fuel" has been defined to be 344.5 inches above vessel zero. This definition is the same as that given by the FSAR for the initial core and maintains the consistency between the various level i definitions given in the FSAR and the technical specifications. Anendnent No. 59 3.2-36 i s

DAEC-1 The low water level instrumentation set to trip at 170" above the top of the active fuel closes all isolation valves except those in Groups 1, 6, 7 and 9 (see notes to Table 3.7-3 for isolation valve groups). Details of vilve grouping and required closing times are given in Specification 3.7. For valves which isolate at this level this trip setting is I i 3.2-364 Amendment No. 59 i i

DAEC-1 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 safety limit. The RBM rod block function provides local protectio'n of the core; i.e., the provention of boiling transition in a local region of the core, for a single rod withdrawat error from a limiting control rod pattern. The IRM rod block function provic -local as well as gross core protection. 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 5 indicated on scale for APRM's and S/125 full l scale for IRM's. i The flow comparator and scram discharge volume high level components have only one logic channel and are not required for safety. The 1 flow comparator must be bypassed when operating with one recirculation water pump. 3.2-42 Aaendnent No. 59

OAEC-1 The value of "R", in units of %ak/k, is the amount by which the core reactivity, in the most reactive condition at any time in the subsequent operating cycle, is calculated to be greater than at the time of the demonstration. "R", therefore, is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of "R" must be positive or zero and must be determined for each fuel cycle. In determining the " analytically strongest" rod, it is assumed that every fuel assembly of the same type has identical material properties. In the actual core, however, the contrul cell material properties vary within allowed manufacturing tolerances, and the strongest rod is determined by a combination of the control call geometry and local Km. Therefore, an additional margin is included in the shutdown margin te t to account for the fact that the " analytically stronges)" r,od is 1 l l l 3.3-9 l Amendment No. 59 l 1 L

J 1 3.3 and

4.3 REFERENCES

1) NEDO 24087-3, 78NED265, class 1_, June 1978 " General Electric Boiling. Water Reactor Reload 3 (Cycle 4) 1,1 censing Amendment for Duane Arnold Energy Center, Supplement 3: Application of Measured Scram Times". I l i r i Amendment No. 59 3,3, 19 l l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 2. The temperature of the 2. Temperature: Check and liquid control solution record at least once per shall be maintained above day. l the curve shown in Figure 3.4-2. This includes the piping between the standby liquid control, tank and the suction inlet to the pumps. 3. Concentration: Check and record at least once per month. Also check con-centration anytime water or boron is added to the i solution or solution j temperature is below the temperature required in Figure 3.4-2. D. If Specification 3.4. A through C cannot be met, the reactor shall be placed in a Cold Shutdown i Condition with all operable control rods fully inserted l" within 24 hours. [ i 0 4 l i l 3.4-3 l

DAEC-1

3. 4 BASES Standby Liquid Contro1 System 1.

The purpose of the liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shut-down condition assuming that none of the withdrawn control rods can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of,bo,ron that produces a concentra-tion of 600 ppe of boron in the reactor core in less than 96 minutes. The 600 ppm concentration in the reactdr core is required to bring the reactor from full power to a subcritical condition, considering the hot to cold reactivity difference, xenon poisoning, analytical b,iases and uncertainties, etc. The time requirement for inserting the boron solution was selected to override the rate of reactivity insertion caused by cooldown of the reactor following the xenon poison peak. The minimum limitaticn on the relief valve setting is intended to prevent the recycling of liquid control solution via the lifting of a relief valve at too low a pressure. The upper limit on the relief valve settings provides system protection from overpressure. Anendment No. 59 3.4-4

o DAEC-1 the direct scram (valve position scram) results in a peak vessel pressure less than the code allowable overpressure limit of 1375 psig if a flux scram is assumed. 1 The analyses of the plant' isolation transients are evaluated in each I reload analyses. These analyses show that the six relief valves assure margin below the setting of the safety valves such that the safety valves would not be expected to open during any anticipated normal transient. These analyses verify that peak system pressure is limited to greater than a 125 psi margin to the allowed vessel overpressure of 1375 psig. Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to t 9 3.6-24 Amendment No. 59 L

1 OAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

3. 12 CORE THERMAL LIMITS 4.12 CORE THERMAL LIMITS Applicability Applicability The Limiting Conditions for The Surveillance Requirements i

Operation associated with the apply to the parameters which fuel rods apply to those monitor the fuel rod operating parameters which monitor the conditions, fuel rod operating conditions. Objective Objective The Objective of the Limiting The Objective of the Surveil-Conditions for Operation is to lance Requirements is to assure the performance of the specify the type and frequency i fuel rods. of surveillance to be applied to the fuel rods. Specifications doecifications A. Maximum Average Planar Linear A. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Heat Generation Rate (MAPLHGR) During reactor power operation, The MAPLNGR for each type of the actual MAPLHGR for each type fuel as a function of average of fuel as a function of average planar exposure shall be planar exposure shall not exceed determined daily during the limiting value shown in Figs. reactor operation at > 25% l 3.12-2, -3, -4, -5, -6, and 7. rated thermal power. If at any time during reactor power operation it is determined by normal surveillance that the limiting value for MAPLHGR (LAPLHGR) is being exceeded, i action shall then be initiated within 15 minutes to restore i operation to within the pre-scribed limits. If the MAPLHGR (LAPLHGR) is not returned to within the prescribed limits within two hours, the reactor shall be brought to the cold shutdown condition within 36 hours. Surveillance and corresponding action shall continue until the prescribed limits are again being met. Amendment No. 59

o DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT B. Linear Heat Generation Rate B. Linear Heat Generation Rate (LHGR) (LHGR) 1. During reactor power opera-The LHGR as a function of core tion, the linear heat genera-height shall be checked daily tion rate (LHGR) of any rod during reactor operation at l in any 7x7 fuel assembly at > 25% rated thermal power. any axial location shall not exceed the maximum allowable LHGR as calculated by the following equation: LHGR < LHGR (1 - {(AP/Pj,x(L/LT)}] eax d LHGR = Design LHGR = 18.5 KW/ft d (7x7 array) j (aP/P,,, = Maximum power spiking penalty = 0.026 LT = Total core length - 12_ feet L = Axial position above bottom of core. 2. During reactor power operation the linear heat generation rate (LHGR) of any rod in any 8x8 fuel assembly shall not exceed 13.4 KW/ft. If at any time during reactor power operation it is deter-mined by normal surveillance that the limiting value for LHGR is being exceeded, action shall then be ini-tiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within two hours, the reactor shall be brought to the cold shutdown condition within 36 hours. Surveil-lance and corresponding action shall continue until the prescribed limits are [ again being met. 3.12-2 Amendment No. 59 i

DAEC-1 l l 1 3.12 BASES: CORE THERMAL LIMITS l A. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR Part 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 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 i 20*F rela-time 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 Part 50, Appendix K limit. The calculational procedure used to establish the MAPLHGRs is based on a loss-of-coolant accident analysis. The analysis was perfonned using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR Part 50. l 3.12-4 Amendnent No. 59

O DAEC-1 5. 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 and that the fuel cladding 1% plastic diametral strain linear heat generation rate is not e3ceeded during any abnormal operat-ing transient if fuel pellet densification is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of Reference 3 and in References 4 ano 9, and assumes a linearly increasing variation in axial gaps between core bottom and top, and assures with a 95% confidence, that no more than one fuel rod exceeds the design linear heat genera-tion rate due to power triking. The LHGR as a function of core height shall be checked daily during reactor operation at > 25% power to determine if fuel burnup, 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. C. Minimum Critical Power Ratio (MCPR) 1. Operating Limit MCPR The required operating limit MCPR's at steady state operat-ing conditions as specified in Specification 3.12.C are 3.12-5 Amendment No. 59 l

DAEC-1 derived from the established fusi cladding integrity Safety Limit MCPR value, and an analysis of abnormal operational transients (1) For any abnormal operating transient analysis evaluation with the it.itial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip settings given in Specification 2.1. \\ i l l 1 3.12-Sa Anendment No. 59

o:C-1 To assure that the fuel cladeling integrity Safety Limit is not, exceeded during any anticipated abnormal operational transient, the most limiting trar.siente have been analyzed to determine which result in the largest reduction in . critical power ratio (CPR). The type of transients evalu-ated were loss of flow, increase in pressure and power,. positive reactivity insertion, and coolant temperature decrease. The limiting transient, which determines the required steady state MCPR limit, is the transient which yields the largest aCPR. The minimum operating limit MCPR of Spec'ification 3.12.C bounds the sum of the safety limit MCPR and the largest ACPR. i \\ l I I 3.12-6 Amendnent No. 59

DAEC-1 OELETED l l 1 l I l l l l 3 ' U~ ' Anendment No. 59

o DAEC - 1 TABLE 3.12-2 MCPR LIMITS Fuel Type 7x7 1.25 8x8 1.24 8 x 8R 1.26 t = l l l I .1 l l Amendment No. 59 3.12-9a l

e OAEC-1 l 3.12 REFERENCES 1. Quane Arnold Energy Center Loss-of-Coolant Accident Analysis Report, NE00-21082-02-1A, Class I, July 1977, Appendix A. 2. " Generic Reload Fuel Application," NEDE-24011-P-A**. 3. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel," Supplements 6, 7 and 8, NE0M-19735, August 1973. 4. Supplement 1 to Technical Reports on Densifications of General Electric Reactor Fuels, December 14, 1973 (AEC Regulatory Staff). 5. Communication: V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974. 6. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NE00-10802). 7. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR Part 50, Appendix K, NEDE-20566 (Draft), August 1974. 8. Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, NE00-24087, 77 NED 359, Class 1, December 1977. 9. Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, Supplement 2: Revised Fuel Loading Accident Analysis, NE00-24087-2. 10. Boiling Water Reactor Reload-3 Licensing Amendment for Quane Arnold Energy Center, Supplement 5: Revised Operating Limits for Loss of Feedwater Heating, NE00-24987-5. 4 "" Approved revision number at time reload fuel analyses are performed. t ~ Amendment No. 59 I

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. t.;r--'-{7 , L. p 7 p. .1 {' -..-p,.j.} L+. [. F6 t t i 11 L li + - - - * - - t [ - t --_I t c.t'l 1 i1 I i f~t _ _t..y.; t 10 0 5,000 10,000 15,000 20,000 25,000 30,000 Planar Average Exposure (P.,iD/T) JJ When core flow is equal to or less than 707. of raced, the MAPLHGR shall not aceed 957 of the limiti:ug values shown. DUANZ ARNOID ENERCT COiTER ICWA EIICTRIC LIGHT AND POVE.R CCMPANT l< TECHNICAL SPECIFICATIONS L'I ITING A'ERAGE PIANAR LDfEAR HEAT GENERATICN RATE AS A FUNCTION CF PLANAR AVERACE EXPGSURI / FUEL TTFE: P8DP289 FICURI 3.12-7 I i 3. 12-18 i Amendment No. 59 l 1 )

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