Text

Proposed Tech Specs,Allowing single-loop Operation & Jet Pump Surveillance
	ML20206K007
Person / Time
Site:	Hatch
Issue date:	06/20/1986
From:	; GEORGIA POWER CO.
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML20206J962	List: ML20206J957; ML20206K007;
References
	0533C, 533C, TAC-61900, TAC-61901, NUDOCS 8606270333
	Download: ML20206K007 (70)
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Georgia Powerkh ENCLOSURE 3 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 EDWIN I. HATCH NUCLEAR PLANT UNITS 1, 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

SIWGLE-LOOP OPERATION AND dET PUMP SURVEILLANCE PAGE CHANGE INSTRUCTIONS The proposed revisions to the Plant Hatch Unit 1 Technical Specifications (Appendix A to Operating Licenses DPR-7) would be incorporated as follows:

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Figure 3.11-1(Sheet 6) Figure 3.11-1(Sheet 6) l 8606270333 860620

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GeorgiaPowerkh ENCLOSURE 3 (Continued)

NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 EDWIN I. HATCH NUCLEAR PLANT UNITS 1, 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

SINGLE-LOOP OPERATION AND DEI PUMP SURVEILLANCE PAGE CHANGE INSTRUCTIONS The proposed revisions to the Plant Hatch Unit 2 Technical Specifications (Appendix A to Operating License NPF-5) would be incorporated as follows:

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Section Section .P_gge LINITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6. PRIMARY SYSTEM BO'JNOARY 4.6. PRIMARY SYSTEM 800NDARY 3.6-1 A. Reactor Coolant Heatup A. Reactor Coolant Heatup and Cooldown 3.6-1 and Cooldown

8. Reactor Vessel Temperature 8. Reactor Vessel Temperature and Pressure 3.6-1 and Pressure C. Reactor Vessel Head Stud C.

Reactor Vessel Head Stud 3.6-2 Tensioning Tensioning

0. Idle Recirculation Loop D.

Startup Idle Recirculation LoGP 3.6-2 Startup E. Recirculation Pump Start E. Recirculation Puno Start 3.6-3 F. Reactor Coolant Chemistry F. Reactor Coolant Chemistry 3.6-4 G. Reactor Coolant Leakage G. Reactor Coolant Leakage 3.6-7 H. Safety and Relief Valves H. Safety and Relief Valves 3.6-9

1. Jet Pumps 1. Jet Pumps 3.6-9b J. Recirculation System J. Recirculation System 3.6-9c K. Structural Integrity K. Structural Integrity 3.6-10 L. Snubbers L. Snubbers 3.6-104 3.7. CONTAINMENT SYSTEMS 4.7. CONTAINMfMT SYSTEMS 3.7-1 A. Primary Containment A. Primary Containment 3.1-1

8. Standby Gas Treatment System 8. Standby Gas Treatment System 3.7-10 C. Secondary Containment C. Secor:dary Containment 3.7-12 D. Primary containment D. Primary Containment Isolation Valves 3.7-13 Isolation Yalves 3.8. RA010 ACTIVE MATERIALS 4.8. RADIDACTIVE MATERIALS 3.8-1 A. Miscellaneous Radioactive A.

Materials Sources Miscellaneous Radioactive 3.8-1 Materials Sources 3.9. AUXILIARY ELECTRICAL SYSTEMS 4.9. AUXILIARY ELECTRICAL SYSTEMS 3.9-1 A. Requirements for Reactor A. Auxiliary Electrical Startup 3.9-1 Systems Equipment HATCH - UNIT 1 iv Proposed TS/0028q/168

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l SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS l

j 1.1. FUEL CLA00!NG INTEGRITY 2.1. FUEL CLAODING INTEGRITY

+ Apolicability Acolicability The Safety Limits established to pre- The Limiting Safety System Settings i serve the fuel cladding integrity apply apply to trip settings of the instru-to those variables which monitor the ments and devices which are provided to fuel thermal behavior. prevent the fuel cladding integrity l Safety Limits from being exceeded.

Objective Objective The objective of the Safety Limits is The objective of the Limiting Safety i to establish limits below which the System Settings is to define the level

} integrity of the fuel cladding is- of the process variables at which auto-

] preserved. matic protective action is initiated

to prevent the fuel cladding integrity

Safety Limits from being exceeded.

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Specifications Specifications s

A. Reactor Pressure > 800 osia and Core A. Trio Settinas j Flow > 10% of Rated

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The limiting safety system trip set-i The existence cf a minimum critical tings shall be as specified below:

i power ratio (MCPR) less than 1.07 for two-loop operation or 1.08 for 1. Neutron Flux Trio Settinas single-loop operation shall constitute violation of the fuel j cladding integrity safety limit.

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a. IRM High High Flux Scram Trio

{ Settina

8. Core Thermal Power Limit (Reactor The IRM flux scram trip setting Pressure s 800 osia) shall be $ 120/125 of full scale.

3 When the reactor pressure is 1 800 b. APRM Flux Scram Trio Settine

psia or core flow is less than 10% of (Refuel or Start & Hot Standby i

rated, the core thermal power shall Mode) not exceed 25% of rated thermal power. When the Mode Switch is in the REFUEL or START & HOT STAN08Y

! position, the APRM flux scram

2. Power Transient trip setting shall be $ 15/125 of full scale (i.e., 1 15% of rated a To ensure that the Safety Limit estab- thermal power).-

i lished in Specification 1.1.A and

1.1.8 is not exceeded, each required c. APRM Flux Scram Trio scram shall be' initiated by its Settinos (Run Mode)

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expected scram signal. The Safety limit shall be assumed to be exceeded (1) Flow Referenced Simulated 4

when scram is accomplished by a means Thermal Power Monitor Scram

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i other than the expected scram signal. Trio Settina When the Mode Switch is in

! the RUN position the APRM flow referenced simulated thermal power scram trip setting shall be:

HATCH - UNIT 1 1.1 -1 Proposed TS/0019q/168

SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 1.1.D. Reactor Water Level (Hot or Cold 2.1.A.1.c.(1) Flow Referenced Simulated Shutdown Condition) Thermal Power Monitor Trio Settina (Run Mode) (Continued)

Whenever the reactor is in the Hot or Cold Shutdown Condition with S $ 0.58W + 62% - 0.58 AW irradiated fuel in the reactor vessel, (Not to exceed 117%)

the water level shall be > 378 inches above vessel invert when fuel is where:

seated in the core.

S = Setting in percent of rated thermal power (2436 MWt)

W = Total loop recirculation flow rate in percent of rated (rated loop recircu-lation flow rate equals 34.2 x 106 lb/hr)

AW = Maximum measured difference between two-loop and single-loop drive flow for the same core flow in percent of rated recirculation flow for single-loop operation. The value is zero for two-loop operation.

(2) Fixed APRM High High Flux Scram TriD Settina (Run Mode)

The APRM fixed flux scram trip setting shall not be allowed to exceed 120% of rated thermal power.

l HATCH - UNIT 1 1.1-2 Proposed TS/0019q/168 l

BASES FOR SAFETY LIMITS 1.1 FUEL CLADDING INTEGRITY A. Fuel Cladding Integrity Limit at Reactor Pressure >800 osia and Core Flow >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 uncer tainties in monitoring the core operating state and in the procedures 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 determined using the General Electric Thermal Analysis Basis, GETAB (1) , which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X)

Boiling Length ~(L), GEXL, correlation. This Safety Limit MCPR is increased by 0.01 for single-loop operation over the comparable two-loop value. (*)

The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.

The required inputs to the statistical model are the uncertainties listed in Table S.2-1 of Reference 2 and the nominal values of the core parameters listed in Table S.2-2 of Reference 2.

O HATCH - UNIT 1 1.1-6 Proposed TS/0019q/168

1.1.A. Reactor Pressure > 800 rsia and Core Flow > 10% of Rated (Cont'd)

The basis for the uncertainties in the core parameters is given in NE00-20340 8 and the basis for the uncertainty in the GEXL correlation is given in NE00-10958 2 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 Hatch Unit No.1 during any fuel cycle would not be as severe as the distribution used in

' the analysis. The method used to handle the uncertainty in the statistical analysis to determine the MCPR cladding integrity Safety Limit for single-loop operation is described in Reference 4.

8. Core Thermal Power Limit (Reactor Pressure 1 800 Dsia)

At pressures below 800 psia, the core elevation pressure drop (0 power.

O flow) is greater than 4.56 psi. At low powers and 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 powers and flows will always be greater than 4.56 psi. Analyses show that with a flow of 28x108 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 driving head will be greater than 28x108 lbs/hr. 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 is conservative.

C. Power Transient i

Plant safety analyses have shown that the scrams caused by exceeding any safety system setting will assure that the Safety Limit of 1.1.A or

1.1.8 will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thernal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.

HATCH - UNIT 1 1.1-7 Proposed TS/0019q/168

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D. Reactor Water level (Hot or Cold Shutdown Condition) i For the fuel in the core during periods when the reactor is shutdown, consi-deration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the fuel during this time, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety limit has been established at 378 inches above vessel invert to provide a point which can be monitored and also provide adequate margin.

E. References i

1. " General Electric BWR Thermal Analysis Basis (GETA8) Data, Correlation and Design Application," NEDO-10958-P-A and NED0-10958-A, January 1977.

2. " General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDE-240ll-P-A.

i 3. General Electric " Process Computer Performance Evaluation Accuracy",

NE00-20340, and Amendment 1. NED0-20340-1, dated June, 1974 and i December,1974, respectively.

4. "Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"

NED0-24205, August 1979.

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I HATCH - UNIT I 1.1-8 Proposed TS/00194/168 i

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BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.1 FUEL CLADDING INTEGRITY The abnormel operational transients applicable to operation of the HNP-1 Unit have been analyzed throughout the spectrum of planned operating conditions up to the thermal power condition of 2537 MWt. The analyses were based upon plant operation in accordance with the operating map given in Figure 3-1 of Ref. 3. l In addition, 2436 MWt is the licensed maximum power level of HNP-1, and this represents the maximum steady-state power which shall not knowingly be exceeded.

Transient analyses perforned for each reload are given in Reference 1. Models and model conservatism are also described in this reference. As discussed in Reference 2, the core-wide transient analyses for single-loop operation are conservatively bounded by two-loop analyses. The flow dependent rod block and scram setpoint equations are adjusted for one-pump operation.

Steady-state operation without forced recirculation will not be permitted, except during startup testing. The analysis to support operation at various f

HATCH - UNIT 1 1.1-10 Proposed TS/0019q/168

BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.1 FUEL CLADDING INTEGRITY (Continued) power and flow relationships has considered operation with either one or two recirculation pumps, References 1 and 2.

A. Trio Settinas The bases for individual trip settings are discussed in the following para-g raphs.

1. Neutron Flux Trio Settinas

a. IRM Flux Scram TriD Setting

! The IRM system consists of 8 chambers, 4 in each of the reactor protec-tion system logic channels. The IRM is a 5-decade instruvent which cov-ers 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 setting of 120 divisions is active in each range of the IRM. For example, if the instrument were on range 1, the scram setting would be a 120 divisions for that range; likewise, if the instrument were on range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up.

The most significant sources of reactivity change during the power in-crease are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the phys-ical limitation of withdrawing control rods, that heat flux is in equi-

librium with the neutron flux and an IRM scram would result in a reac-tor shutdown well before any Safety Limit is exceeded.

i

! In order to ensure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents i

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. Quarter rod den-sity is illustrated in Figure 7.5-8 of the FSAR. Additional conserva-1, 1

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HATCH - UNIT 1 l I-Il Proposed TS/0019q/168

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l i l BASES FOR LIMITING SAFETY SYSTEM SETTINGS i

j 2.1.A.1.a. IRM Flux Scram Trio Settine (Continued) j tism was taken in this analysis by assuming that the IRM channel closest l to the withdrawn rod is bypassed. The results of this analysis show that l the reactor is scramed and peak power limited to one percent of rated I power, thus maintaining MCPR above the fuel cladding integrity Safety Limit, y Based on the above analysis, the IRM provides protection against local control l rod withdrawal errors and continues withdrawal of control rods in sequence l and provides backup protection for the APRM.

i i b. APRM Flux Scram Trio Settino (Refuel or Start & Hot Standby Mode)

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 l thermal margin between the setpoint and the safety limit, 25 percent of l rated. The margin is adequate to accomodate anticipated maneuvers asso-7 ciated with power plant startup. Effects of increasing pressure at zero a or low void content are minor, cold water from sources available during i

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 and i

the Rod Sequence Control System. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of sig- i nificant 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 l 4

flux is in near equilibrium with the fission rate. In an assumed uniform i i rod withdrawal approach to the scram level, the rate of power rise is no {

q more than 5 percent of rated power per minute, and the APRM system would !

j be more than adequate to assure a scram before the power could exceed j

, the safety limit. The 15 percent APRM scram remains active until the j mode switch is placed in the RUN position. This switch occurs when i

reactor pressure is greater than 825 psig.

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, c. APRM Flux Scram Trio Settinas (Run Mode)

)

4 The APRM Flux scram trips in the run mode consist of the flow referenced simulated t'iermal power monitor scram setpoint and a fixed high-high j neutron flux scram setpoint. In the simulated thermal power monitor, ,

J the APRM flow referenced neutron flux signal is passed through a filter- ;

l ing network with a time constant which is representative of the fuel dy- '

namics. This provides a flow referenced signal that approximates the ,

average heat flux or thermal power that is developed in the core during I l transient or steady-state conditions. This prevents spurious scrams, which have an adverse effect on reactor safety because of the resulting i

thermal stresses. Examples of events which can result.in momentary neutron flux spikes are momentary flow changes in the recirculation system flow, and small pressure disturbances during turbine stop valve l and turbine control valve testing. These flux spikes represent no j

hazard to the fuel since they are only of a few seconds duration and less than 120% of rated thermal power. The flow independent portion of this scram setpoint must be adjusted downward during single-loop opera-tion to account for lower core flow with respect to two-loop operation with the same drive flow.

HATCH - UNIT 1 1.1-12 Proposed TS/0019q/168 l

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BASES FOR LIMITING SAFETY SYSTEM SETTINGS l

2.1.A.l.c. APRM Flux Scram Trip Settinos (Run Mode) (Continued) l l

The APRM flow referenced simulated thermal power monitor scram trip setting at full recirculation flow is adjustable up to 117% of rated power for two-recirculation loop and single-recirculation loop operations. l This reduced flow referenced trip setpoint will result in an earlier scram during slow thermal transients, such as the loss of 100*F feedwater heating event, than would result with the 120% fixed high neutron flux scram trip. The lower flow referenced scram setpoint therefore decreases the severity (ACPR) of a slow thermal transient and allows lower Operating Limits if such a transient is the limiting abnormal operational transient during a certain exposure interval in the cycle.

The APRM fixed high-high neutron flux scram trip, adjustable up to 120%

of rated power for two-recirculation loop and single-recirculation loop operations, does not incorporate the time constant, but responds directly to instantaneous neutron flux. This scram setpoint scrams the reactor during fast power increase transients if credit is not taken for a direct (position) scram, and also serves to scram the reactor if credit is not taken for the flow referenced scram.

2.

Reactor Vessel Water low level Scram Trio Settino (Level 3) 1he trip setting for low level scram is above the bottom of the separator skirt, Figure 2.1-1. This level is > 14 feet above the top of the active fuel. l This level has been used in transiert analyses dealing with coolant inventory decrease. The results reported in FSAR Section 14.3 show that a scram at this level adequately protects the fuel and the pressure barrier. The designated scram trip setting is at least 22 inches below the bottom of the normal operating range and is thus adequate to avoid spurious scrams.

.a HATCH - UNIT I 1.1-13 Proposed TS/0019q/168

i BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.1.A.3. Turbine Stoo Valve Closure Scram Trio Settinas f The turbine stop valve closure scram trip anticipates the pressure, neutron j flux and heat flux increase that could result from rapid closure of the 2 turbine stop valves. With a scram trip setting of < 10 percent of valve _

i closure from full open, the resultant increase in surface heat flux is

{ limited such that MCPR remains above the-fuel'. cladding integrity Safety j Limit during the worst case transient that assumes the turbine bypass is closed. This scram is bypassed when turbine steam flow is below that I corresponding to 30% of rated thermal power, as measured by turbine first j stage pressure.

! 4. Turbine Control Valve Fast Closure Scram Trio Settina j

This 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 turbine bypass. The Reactor Protection System initiates a scram when fast closure of the control valves is initiated by the fast acting solenoid valves. This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is sensed i i

by pressure switches whose contacts torn the one-out-of-two-twice logic input to the reactor protection system. This trip setting, a nominally 1

50% greater closure time and a dif ferent valve characteristic from that of the turbine stop valve, combine to produce transients very similar and no i more severe than for the stop valve. This scram is bypassed when turbine i steam flow is below that corresponding to 30% of rated thermal power, as

] measured by turbine first stage pressure.

5. Main Steam Line Isolation Valve Closure Scram Trio Settina i

The main steam line isolation valve closure scram occurs within 10% of

) valve movement from the fully open position and thus anticipates the l neutron flux and pressure scrams which remain as available backup pro-i tection. This scram function is bypassed automatically when the Mode l

Switch is not in the RUN position.

6. Main Steam Isolation Valve Closure on Low Pressure i

' The low pressure isolation of the main steam lines at 825 psig was provided

! to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel, which might result from a pressure regulator i failure causing inadvertent opening of the control andh r bypass valves.

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U HATCH - UNIT I 1,1 14 Proposed TS/0019q/168 h

BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.1.C. References

1. " General Electric Standard Application for Reactor Fuel (Supplement for United States)." NEDE-24011-P-A.

2. "Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"

NE00-24205, August 1979.

3. " Average Power Range Monitor, Rod Block Monitor and Technical Specifications Improvement (ARTS) program for Edwin I. Hatch Nuclear Plant, Units 1 and 2,"

NEDC-30474-P, December 1983.

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k HATCH - UNIT 1 1.1-17 Proposed TS/0019q/168

Table 3.1-1 (Cont'd) l at 3= Scram Ope ra b le

Q Number Source of Scram Trip Signal Channels Scram Trip Setting Source or Scram Signal is Required at (a) Required Per to be Operable Except as Indicated

, Trip System Below c:

(b)

I di 5 H igh D rywe l l Pressure 2 5 1.92 psig Not required to be operable when

-1 primary containment integrity is

,, not required.

6 Reactor Vessel Water Level 2 2 10.0 inches (Low) (Level 3) 7 Scram Discharge Volume Permissible to bypass (initiates High High Level control rod block) in order to

a. Float Switches 2 5 71 gallons reset RPS when the Mode Switch is

b. in the REFUEL or SHUTDOWN position, Therma l Level Sensors 2 5 71 gallons 8 . APRM Flow Referenced Simulated 2 S S 0.58W+62% - 0.58 AW See Specification 2.1.A.1.c(1) for The rma l Power Monitor (Not to exceed 117%) dorinitions of W and 4W'.

Tech Spec 2.1.A.1.c(1)

Fixed High High Neutron 2 S 5 120% Power Flux Tech Spec 2.1. A.1.c(2 )

I Inope ra t ive 2 Not Applicable An APRM is inoperable if there are La

less than two LPRM inputs per level T' cr there are less than 11 LPHM 4 inputs to the APRM channel.

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Proposed TS/0020q/168

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Table 3.2-7 (Continued) 4._ y

-(

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Regis t red Ope ra b le e Rer., Trip Channois '

c ' No . ~ , Condition per Trip t z (a) Inst rument Nomencla ture System Trip Setting .9ema rk s '

3. APRM Downscale 2(b)(e) 23/125 or rull scefe Not required while performing low N

power physics , test at atmosg:hcric

- ,, pressure duriss or af ter refueling a t powe r leve n s no t to exceed 5 MWt.

12% riux 2(b)(e) 512/125 or full'scate 'This functiosi is bypassed when the

V4de Switch is~placed in the RUN Position.

Upscale 2(b)ie) , 50.58 W + 50% - 0.58 AW See s,,ectrication 2.1.A.I.O(1) for l y derir.itions of W and AW. Trip l levet setting is in percent or ra ted power. Not reqtsi red wrei te performing low power physics tests at atmospheric pressure during or

" arter refueling r.t power levels not to exceed 5 MWt..

14 IlqM inopera tive 1(e)( r) s

,, i y Not applicable Inoperative trip produced by switch W ,

not i n ope ra te, c e rcui t boa rds set m, ' - ' in circuit, rails to null,.less a " ' than required numboe or LPRM enputs for rod selected.

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-Downscale 1(e)(r) 298/125 4 or rull scale 7

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Proposed TS/00204/168

BASES FOR llMITING CONDITIONS FOR OPERATION 3.2.F.5. Core Sorav Pump Discharge Flow A differential pressure transmitter is provided downstream of each core spray pump to indicate the condition of each pump. To protect the pumps from over-heating at low flow rates a minimum flow bypass line, which routes water from the pump discharge to the suppression chamber, is provided. A single motor-operated valve controls the condition of each bypass line. The minimum flow bypass valve automatically opens upon sensing low flow in the discharge line.

The valve automatically closes whenever the flow is above the low flow setting.

6. Core Sorav toqic Power Failure Monitor The Core Spray Logic Power Failure Monitor monitors the availability of power to the logic system. In the event of loss of availability of power to the logic system, an alarm is annunciated in the control room.

G. Neutron Monitorino Instrumentation Which initiates Control Rod Blocks (Table 3.2-7)

These control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to the fuel cladding integrity Safety Limit. The trip logic for this function is I 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 instrumentation to assure that the single failure criteria is met.

1. SRM

a. Inoperative

. This rod block assures that no control rod is withdrawn during low neutron flux level operations unless proper neutron monitoring capa-bility is available, in that all SRM channels are in service or properly bypassed,

b. Not Fully inserted Any source range monitor not fully inserted into the core when the SRM count rate level is below the retract permit level will cause a rod block. This assures that no control rod is withdrawn unless all SRM detectors are properly inserted when they must be relied upon to pro-vide the operator with a knowledge of the neutron flux level,

c. Downscale This rod block assures that no control rod is withdrawn unless the SRM count rate is above the minimum prescribed for low neutron flux level monitoring.

HATCH - UNIT 1 3.2-63 Proposed TS/0019q/168

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.G.l.d. UDscale This rod block assures that no control is withdrawn unless the SRM detectors are properly retracted during reactor startup. This setting is selected at the upper end of the range over which the SRM is designed to detect and measure neutron flux.

2. IRM The trip logic for this function is 1 out of 8; any trip on one of the eight IRM's will result in a rod block. The IRM rod block function provides local as well as gross core protection.

a. Inoperative This rod block assures that no control rod is withdrawn unless the IRM's are in service,

b. Not Fully Inserted (Refuel and Start & Hot Standby Mode)

This rod block assures that no control rod is withdrawn during low neutron flux level operations unless proper neutron monitoring capability is available in that all IRM detectors are properly located,

c. Downscale A downscale indication of 5 5/125 full scale on an IRM is an indication that 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 trip is set at > 5/125 full scale. This rod block trip is bypassed when the IRM l is on the range 1.

d. High Flux If the IRM channels are in the worst condition of allowed bypass, the scaling arrangement is such that for unbypassed IRM channels a rod block signal.is generated before the detected neutron flux has increased by

, more than a factor of 10.

3. APRM The trip logic for this function is 1 out of 6; any trip on one of the six APRM's will result in a rod block. The APRM rod block function 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 fuel cladding integrity I Safety Limit under normal operating conditions.

a. Inoperative I' This rod block assures that no control rod is withdrawn unless the APRM's are in service.

HATCH - UNIT 1 3.2-64 Proposed TS/0019q/168

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS

b. With the relief valve function and/or the low low set function of more than one of the above required reactor coolant system relief / safety valves inoperable, be in at least HOT SHUTOOWN with-in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.6.I. Jet Pumps 4.6.I. Jet Pumos All jet pumps corresponding to the Each of the jet pumps shall be operating loop (s) shall be operable demonstrated operable prior to during STARTUP and RUN modes by thermal power exceeding 25% of meeting at least une of the following rated power; following recircu-rcquirements: lation pump restarts; following any unexpected or unexplained

1. For any specific core flow change in core flow, jet pump loop condition, each individual jet pump flow, recirculation pump flow, or flow shall not differ by more than core plate dif ferential pressure; 10% of the average loop jet pump and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by flow f rom the normal range

of recording jet pump loop flows, average loop jet pump flows recirculation pump flows, recircu-experienced for these flow lation pump speeds, and individual conditions, or jet pump flows (D/P); and verifying

2. For any specific core flow that neither of the following conditions occur:

condition, each individual jet pump diffuser to lower plenum 1. The recirculation pump flow /

differential pressure (D/P) shall speed ratio deviates more than not differ by more than 20% of 5% f rom the normal range,* or the average loop D/P from the normal range

of average loop jet 2. The jet pump loop flow / speed pump D/Ps experienced for these ratio deviates more than 5%

flow conditions. f rom the normal range.*

With one or more jet pumps exceeding If any required jet pump fails to the above requirements, evaluate the reason for the deviation, and be in meet either or both of the above 4.6.I.1 or 4.6.I.2 Surveillance HOT SHUTOOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in the Requirements, review the jet pump circumstance that one or more jet operability as defined in the-LCO pumps tre verified to be inoperable. Section 3.6.1 and in BASES Section 3.6.I.

Normal expected operating range based on data obtained f rom operating experience.

HATCH - UNIT 1 3.6-9b Proposed TS/0025q/196

l l

l LIMITING CONDITIONS FOR OPERATION SURVE[LLANCE RE0VIREMENTS 3.6.J. Recirculation System 4.6.J. Recirculation System

1. Core thermal power shall not exceed 1. Recirculation pump speeds shall be l 1% of rated thermal power without recorded at least once per day.

forced recirculation.

2. With only one recirculation loop

2. Whenever the reactor is in the in operation, verify that the STARTUP or RUN modes, at least one reactor operating conditions are -

recirculation loop shall be in outside the Operation Not Allowed operation. Region in Figure 3.6-5:

3. The requirements applicable to (a) At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , single-loop operation as identified in Sections 1.1.A. 2.1.A. 3.1.1, (b) Whenever thermal power has 3.2.G. 3.11. A, and 3.11.C shall be been changed by at least 5% of in effect following the removal of rated thermal power and steady-one recirculation loop from service, state conditions have been or the unit shall be placed in the reached.

HOT SHUTOOWN condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4. With only one recirculation loop in operation and the unit in the Operation Not Allowed Region, specified in Figure 3.6-5, initiate action within 15 minutes to place the unit in the Operation Allowed Region, identified in Figure 3.6-5, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Otherwise, place the reactor in the HOT SHUTOOWN condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

5. Following one pump operation the discharge valve of the low speed pump may not be opened unless the speed of the faster pump is less than 50% of its rated speed. ~

HATCH - UNIT 1 3.6-9c Proposed TS/0025q/196

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BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.6.H. Relief / Safety Valves (Continued)

Experience in relief / safety valve operation shows that a testing of 50 per-cent of the valves per year is adequate to detect failure or deteriorations.

The relief / safety valves are benchtested every second operating cycle to ensure that their set points are within the tolerance given in Specification 2.2.A. The relief / safety valves are tested in place at low reactor pressure once per operating cycle to establish that they will open and pass steam.

The requirements established above apply when the nuclear system can be pres-surized above ambient conditions. These requirements are applicable at nu-clear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed. However, these transients are much less severe in terms of pres;ure, than those starting at rated condi-tions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.

The low low set (LLS) system lowers the opening and closing setpoints on four preselected relief / safety valves. The LLS system lowers the setpoints af ter any relief / safety valve has. opened at its normal steam pilot setpoint when a concurrent high reactor vessel steam dome pressure scram signal is present.

The purpose of the LLS is to mitigate the induced high frequency loads on the containment and thrust loads on the SRV discharge line. The LLS system incr, eases the amount of reactor depressurization during a relief / safety valve blowdown because the lowered LLS setpoints keep the four selected LLS relief / safety valves open for a longer time. The high reactor vessel steam dome pressyre signal for the LLS logic is provided by the exclusive analog trip channels. The purpose of installing special dedicated steam dome pres-sure channels is to maintain separation from the RPS high pressure scram functions.

I. Jet PumDs Failure of a jet pump nozzle assembly hold down mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown fol-lowing the design basis double-ended line break. Therefore, if a failure occurred, repairs must be made.

HATCH - UNIT 1 3.6-21 Proposed TS/0025q/196

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0ijlREMENTS 3.6.I. Jet Pumps (Continued)

A nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle riser system failure.

One of the acceptable procedures for jet pump surveillance, identified in NUREG/CR-3025, Reference 2, was adopted for Hatch Unit 1. The surveillance is performed to verify that neither of the following conditions occur:

(a) The Recirculation Pump Flow / Speed Ratio deviates by more than 5% from the normal range, or (b) The Jet Pump Loop Flow / Speed Ratio deviates by more than 5% from the normal range.

If either criterion is failed, then the procedure calls for comparing either the individual jet pump flow or individual jet pump diffuser to lower plenum differential pressures to the criteria of the Limiting Conditions for Operation (LCO). If the LCO criteria are not satisfied and pump speed is less than 60% of rated, it may be necessary to increase pump speed to above 60% of rated and repeat the measurements before declaring a jet pump inoperable. In this case, it is recommended that close monitoring and increased recirculation pumo speed should be perfonned only if the criteria are exceeded by an amount to be determined from previous plant operating experience.

3.6.J. Recirculation System Operation with a single reactor coolant system recirculation pump is allowed, provided that adjustments to the flow referenced scram and APRM rod block setpoints, MCPR cladding integrity Safety Limit, MCPR Operating Limit, and MAPLHGR limit are made. An evaluation of the performance of the ECCS with single-loop operation has been performed and detennined to be acceptable, Reference 4. Based on this Reference, a MAPLHGR factor of 0.75 is applied to the specifications, Figure 3.11-1 (Sheet 6). To account for ,

increased uncertainties in the total core flow and TIP readings when '

operating with a single recirculation loop, a 0.01 increase is applied to the MCPR cladding integrity Safety Limit and MCPR Operating Limit over the comparable two-loop values. The flow referenced simulated thermal power -

scram and rod block setpoints for single-recirculation-loop operation is reduced by the amount of maw, where m is the flow reference slope for the rod block monitor and AW is the largest difference between two-loop and single-loop effective drive flow when the active loop indicated flow is the same. This adjustment is necessary to preserve the original relationship between the rod block and actual effective drive flow.

When restarting an idle pump, the discharge valve of the idle loop is required to remain closed until the speed of the faster pump is below 50% of its rated speed to provide assurance that when going from one- to two-loop operations, excessive vibration of the jet pump risers will not occur.

The possibility of experiencing limit cycle oscillations during single-loop operation is precluded by restricting the core flow to greater than or equal to 45% of rated core thermal power when core power is greater than the 80%

rod line. This requirement is based on General Electric's recommendations contained in SIL 380, Revision 1, which defines the region where the limit cycle oscillations are more likely to occur.

HATCH - UNIT 1 3.6-22 Proposed TS/0025q/196

BASES 3/4.6.L. SNUBBERS (Continued)

The service lif e of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc...). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.

References:

(1) Report, H. R. Erickson, Bergen Paterson to K. R. Goller, NRC, October 7,1974.

Subject:

Hydraulic Shock Sway Arrestors.

(2) NUREG/CR - 3052, " Closeout of IE Bulletin 80-07: BWR Jet Pump Assembly Failure," Published November 1984.

(3) " General Electric BWR Licensing Report: Average Power Range Honitor, Rod Block Honitor, and Technical Specifications Improvement (ARTS)

Program for Edwin I. Hatch Nuclear Plant Units 1 and 2,"

NEDC-30474-P, December 1983.

(4) "Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"

NEDO-24205, August 1979.

HATCH - UNIT 1 3.6-32 Proposed TS/00254/196

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limit, which is a function of average rated thermal power.

planar exposure and fuel type, is the appropriate value from Figure 3.11-1, sheets 1 through 5, multiplied by the '

smaller of the two MAPFAC factors de-termined from Figure 3.11-1, sheets 6 and 7. For single-loop operation, the MAPFAC p is a constant value of 0.75 when power is greater than 52%

of rated thernal power. For power less than 52% of rated thermal power, the MAPFAC p is the same as the comparable two-loop value (Figure -

3.11.1, sheet 6) . If at any time during operation it is determined by normal surveillance that the limiting value for APLHGR is being l exceeded, action shall be initiated i within 15 minutes to restore operation to within the prescribed limits.

If the APLHGR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal

power within the next four (4) hours.
If the limiting condition for operation is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated l thermal power is not required.
HATCH - UNIT 1 3.11-1 Proposed TS/0019q/168 4
LIMITING _ CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
8. Linear Heat Generation Rate (LHGR) 8. Linear Heat Generation Rate (LHGR)
During power operation, the LHGR as The LHGR as function of core a function of core height shall not height shall be checked daily dur-exceed the limiting value shown in ing reactor operation at 2 25% .
Figure 3.11-2 for 7 x 7 fuel or the rated thermal power.
limiting value of 13.4 kw/ft for 8 x 8/
8 x 8R fuel. If at any time during i
HATCH - UNIT 1 3.11 -l a Proposed TS/0019q/168 e
i 4
LIMITING CONDITIONS FOR OPERAT10N SURVEllLANCE REOUIREMENTS 4
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1 HATCH - UNIT 1 3.ll-lb Proposed TS/0019q/168
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LIMITING CONDITIONS FOR OPERATION SURVEILLAN((_,REOUIREMENTS 3.11.B. Linear Heat Generation Rate (LHGR)
(Continued) operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiited within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power within the next four (4) hours. If the limiting condition for operation is restored prior to expiration of the specified time interval, then further progression to less than 25%
of rated thermal power is not required.
C. Minimum Critical Power Ratio (MCPR) 4.11.C.l. Minimum Critical Power Ratio (MCPR)
MCPR shall be determined to be The mininum for two-loop criticalshall operation power ratio (MCPR) equal be equal l to or greater than the to or greater than the operating applicable limit, daily during limit MCPR (0LMCPR), which is a reactor power operation at 2 25%
function of scram time, core rated thernal power and following power, and core flow. For 25% < any change in power level or dis-power < 30%, the OLMCPR is given in tribution that would cause opera-Figure 3.11.7. For power 2 30%, tion with a limiting control rod the OLMCPR is the greater of either: pattern as described in the bases for Specification 3.3.F.
1. The applicable limit determined f rom Figure 3.11.3, or 4.11.C.2. Minimum Critical Power Ratio Limit
2. The aaplicable limit from either Figures 3.11.4, 3.11.5, The MCPR limit at rated flow and or 3.11.6, multiplied by the ~
rated power shall be determined for Kp factor determined f rom each fuel type, 8X8R, P8X8R, 7X7 Figure 3.11.7, where: from figures 3.11.4, 3.11.5, and 3.11.6 respectively using:
x = 0 or 'Tave TB' , whichever is a. v=1.0 prior to initial scram
. TA ~B . greater time measurements for the cycle, performed in accordance tA = 0.90 sec (Specifications 3.3.C.2.a. with specifications 4.3.C.2.a.
scram time limit to 20% insertion f rom fully withdrawn) or
- I t/m b. r as defined in specification TB = 0.710+1.05 N1 (0.053)[Ref.7]l 3.11.C.
n I Nj The determination of the limit
- i=1 -
must be completed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram i time surveillance test required by specification 4.3.C.2.
HATCH - UNIT 1 3.11-2 Proposed TS/00194/168
LIMITING CONDITIONS FOR OPERATI_0N_ SURVEILLANCE RE0VIREMENTS 3.11.C. Minimum Critical Power Ratio (MCPR)
"n I Njtj tave = i=1 n
I Nj
. i=1 .
n = number of surveillance tests performed to date in cycle Nj = number of active control rods measured in the ith surveillance test tj = Average scram time to 20% in-sertion from fully withdrawn of all rods measured in the ith surveillance test, and, N1 = total number of active rods measured in 4.3.C.2.a.
For single-loop operation, the MCPR limit is increased by 0.01 over the comparable two-loop value.
If at any time during operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power -
within the next four(4) hours. If the Limiting Condition for Operation is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated thernal power is not required.
D. Reportino Reauirements If any of the limiting values iden-tified in Specifications 3.11.A.,
B., or C. are exceeded, a Reportable Occurrence report shall be submitted.
If the corrective action is taken, as described, a thirty-day written report will meet the requirements of this specification.
HATCH - UNIT 1 3.11-2a Proposed TS/0019q/168
BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0UIREMENTS 3.11. FUEL RODS A. Averaae Planar Linear Heat Generation Rate (APLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K, even considering the postulated effects of fuel pellet densification.
The peak cladding temperature following a postulated loss-of-coolant acci-dent 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 second-arily on the rod to rod power distribution within an assembly. Since ex-pected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than ! 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures conform to 10 CFR 50.46. The limiting value for APLHGR at rated conditions is shown in Figures 3.11.1. sheets 1 thru 5.
A flow dependent correction factor incorporated in to Figure 3.11-1 (sheet 7) is applied to the rated conditions APLHGR to assure that the 2200*F PCT limit is complied with during LOCA initiated from less than rated core flow. In addition, other power and flow dependent corrections given in Figure 3.11-1 (sheets 6 and 7) are applied to the rated conditions APLHGR limits to assure that the fuel thermal-nechanical design criteria are met during abnormal transients initiated from off-rated conditions for two-loop and single-loop operations, References 2 and 8. For single-loop operation, a 0.75 multiplica-tion factor to APLHGR limits for all fuel bundle types conservatively bounds that required by Reference 2.
The calculational procedure used to establish the APLHGR shown in Figures 3.11.1, sheets 1 thru 5, is based on a loss-of-coolant accident analysis.
The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.
A list of the significant plant input parameters to the loss-of-coolant' accident analysis is presented in Table 3-1 of NED0-24086(3). Further discussion of the APLHGR bases is found in NEDC-30474-p(*).
For sin ARTS (*)gle-loop MAPLHGRS operation (SLO),
will define the the mostConditon Limiting restrictive forofOperation.
the SLO and i
)
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HATCH - UNIT 1 3.11-3 Proposed TS/0Ci9q/168
I l
BASES FOR LIMITING CONUITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.11. B . Linear Heat Generation Rate (LHGR)
This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification is postulated. The power spike penalty specified for 7 x 7 fuel is based or. the analysis presented in Section 3.2.1 of Reference 4 and References 5 and 6, 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 generation rate due to power spiking. 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 ratio of peak LHGR to core average LHGR would have to be greater than 9.6, which is precluded by a considerable margin when employing any permissible control rod pattern.
C. hinimum Critical Power Ratio (MCPR)
The required operating limit MCPR as specified in Specification 3.ll.C. is derived from the established fuel cladding integrity Safety Limit MCPR and an analysis of abnormal operational transients presented in References 1, 2, and 8.
Various transient events will reduce the MCPR below the operating MCPR.
To assure that the fuel cladding integrity safety limit is not violated during anticipated abnormal operational transients, the most limiting transients have been analyzed to determine which one results in the largest reduction in critical power ratio (a MCPR). Addition of the largest a MCPR to the safety limit MCPR gives the minimum operating limit nCPR to avoid violation of the safety limit should the most limiting transient occur.
The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
l I
i.
HATCH - UNIT 1 3.11-4 Proposed TS/0019q/168
BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS 3.ll.C. Minimum Critical Power Ratio (MCPR) (Continued)
The purpose of the MCPR f , and the Kp of Figures 3.11.3 and 3.11.7, respectively, is to define operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power, the required MCPR is the larger value of the MCPRf and MCPRp at the existing core flow and power state. The MCPRys are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.
The MCPR f s were calculated such that for the maximum core flow rate and the corres-ponding THERMAL POWER along the 105% of rated steam 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 MCPRs were calculated at different points along the 105% of rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPRf .
The core power dependent MCPR operating limit MCPR p is the power rated flow MCPR operating limit multiplied by the Kp factor given in Figure 3.11.7.
The Kp s are established to protect the core from transients other than core flow increases, including the localized event such as rod withdrawal error. The Kps were determined based upon the most limiting transient at the given core power level. (For further information on MCPR operating limits for off-rated conditions, reference NEDC-30474-P.(e))
When operating with a single-recirculation pump, the MCPR Safety and Operating Limits are increased by an amount of 0.01 over the comparable values for two-recirculation pump operation.(2)
HATCH - UNIT 1 3.11-4a Proposed TS/0019q/168
BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS 3.11.E. References
1. " General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDE-240ll-P-A.
2. "Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"
NE00-24205, August 1979.
3. " Loss-of-Coolant Analysis for Edwin I. Hatch Nuclear Plant Unit 1,"
NEDO-24086, December 1977.
4. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel", Supplements 6, 7 and 8, NEDM-10735, August,1973.
5. Supplement 1 to Technical Report on Densification of General Electric ,
Reactor Fuels, December 16,1974 (USA Regulatory Staf f).
6. Communication: V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification", Docket 50-321, March 27, 1974.
7. Letter from R. H. Buchholz (G. E.) to P. S. Check (NRC), " Response to NRC request for information on ODYN computer model". September 5,1980.
E. " Average Power Range Monitor, Rod Block Monitor and Technical Specification l Improvement (ARTS) Program for Edwin I. Hatch Nuclear Plant, Units 1 and 2,"
NEDC-30474-P, December 1983.
HATCH - UNIT 1 3.11-6 Proposed TS/0019q/168
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INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
'SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or Low F10w) . . . . . . . . . . . . . . . . . . . . . . 2-1 THERMAL POWER (High Pressure and High Flow) . . . . . . . . . . . . . . . . . . . 2-1 Reactor Coolant System Pressure............................... 2-1 Reactor Vessel Water Leve1.................................... 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints. . . . . . . . . . . 2-3 BASES 2.1 SAFETY LIMITS
-lLTHERMAL POWER ( Low Pres sure or Low F1ow) . . . . . . . . . . . . . . . . . . . . . .B 2-1 THERMAL POWER (High Pressure and High Flow) . . . . . . . . . . . . . . . . . . . B 2-2
~
Reactor Coolant System Pressure............................... B 2-8 !
Reactor Vessel Water Leve1.................................... B 2-8 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints........... B 2-9
2.3 REFERENCES
.................................................... B 2-13 HATCH-UNIT 2 III Proposed TS/0026q/169 l 1
INDEX BASES SECTION PAGE INSTRUMENTATION (Continued)
Remote Shutdown Monitoring B 3/4 3-3 Instrumentation Post-Accident Monitoring B 3/4 3-4 Instrumentation Source Range Monitors B 3/4 3-4 Traversing Incore Probe System B 3/4 3-4 Chlorine Detectors B 3/4 3-4 Fire Detection Instrumentation B 3/4 3-4 Radioactive Liquid Effluent .
Instrumentation B 3/4 3-5 Radioactive Gaseous Effluent Instrumentation B 3/4 3-5 3/4.3.7 TURBINE OVERSPEED PROTECTION SYSTEM B 3/4 3-5 3/4.3.8 DEGRADED STATION VOLTAGE PROTECTION INSTRUMENTATION B 3/4 3-Sa -
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM B 3/4 4-1 Jet Pumps B 3/4 4-1 Idle Recirculation Loop Startup B 3/4 4-la 3/4.4.2 SAFETY / RELIEF VALVES B 3/4 4-la Low-Low Set Systems B 3/4 4-lb 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems B 3/4 4-2 Operational Leakage B 3/4 4-2 3/4.4.4 CHEMISTRY B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS B 3/4 4-4 HATCH-UNIT 2 XI Proposed TS/0026q/168
r INDEX BASES SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES B 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY B 3/4 4-6 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM B 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM B 3/4 5-1 3/4.5.3 LOW PRESSURE CORE COOLING SYSTEMS Core Spray System B 3/4 5-2 Low Pressure Coolant Injection System B 3/4 5-3 3/4.5.4 SUPPRESSION CHAMBER B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT INTEGRITY Primary Containment Integrity B 3/4 6-1 Primary Containment Leakage B 3/4 6-1 Primary Containment Air Lock B 3/4 6-1 MSIV Leakage Control System B 3/4 6-2 Primary Containment Structural Integrity B 3/4 6-2 Primary Containment Internal Pressure B 3/4 6-2 Drywell Average Air Temperature B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS B 3/4.6-3 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES B 3/4 6-4 3/4.6.4 VACUUM RELIEF B 3/4 6-5 3/4.6.5 SECONDARY CONTAINMENT B~3/4 6-5 3/4.6.6 CONTAINMENT ATMOSPHERE CONTROL B 3/4 6-5 HATCH-UNIT 2 XII Proposed TS/0026q/168
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS l'
THERMAL POWER (Low Pressure or Low Flow) -
2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
APPLICABILITY: CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
THERMAL POWER (High Pressure and High Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 for two-loop recirculation or 1.08 for single-loop recirculation operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.
APPLICABILITY: CONDITIONS 1 AND 2.
ACTION:
With MCPR less than 1.07 for two-loop recirculation or 1.08 for-single-loop recirculation operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
APPLICABILITY: CONDITIONS 1, 2, 3 and 4.
ACTION:
With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTOOWN with reactor coolant system pressure s 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . l l
HATCH-UNIT 2 2-1 Proposed TS/0029q/168 l
E TABLE 2 2.1-1 k
j$ REACTOR PROTECTION SYSTEM INSTRUMENTATION SFTPOINTS '
N FUNCTIONAL UNIT TRIP SETPolNT ALLOWABLE VALUES c= 1 Intermediate Range Monitor, Neutron Flux-High 5 120/125 divisions 5 120/125 divisions ge (2C51-K601 A,B,C,D,E F.G,H) or full scale or full scale
~i
2. Average Power Range Monitor; h3 (2C51-K605 A,B,C,0,E,F)
a. Neutron Flux-Upscale, 15% 5 15/125 divisions s 20/125 divisions or full scale or full scale
b. Flow Referenced Simulated Thermal Powe r-Up sca l e 5 (0.58 W + 59% -0.58 AW)** 5 (0.58 W + 62% -0.58 AW)**
with a maximum with a maximum l 5 113.5% of RATED 5 115.5% of RATED THERMAL POWER
c. Fixed Neutron Flux-Upscale, 118% 5 118% of RATED THERMAL POWER 5 120% or RATED THERMAL POWER THERMAL POWER
3. Reactor Vessel Steam Dome Pressure - High 5 1054 psig (2821-N678 A,B,C,D) 5 1054 psig l 4. Reactor Vessel Water Level - Low (Level 3) 2 8.5 inches above (2B21-N680 A,B,C,D) instrument zero* 2 8.5 inches above Instrument zero*
5. Main Steam Line isolation valve - Closure s 10% closed (NA) 5 10% closed ro
j. 6. Main Steam Line Radiation - High 5 3 x rull power (2D11-K603A,B,C,D) s 3 x rull power backg round backg round
7. Drywell Pressure - High 5 1.85 psig (2C71-N650A,B,C,D) 5 1.85 psig
See Bases Figure B 3/4 3-1.
W = total loop recirculation flow rate in percent or rated.
MLB/hr. Rated loop reci rculation flow is equal to 34.2 1
AW = Maximum measured dirrerence between two-loop and single-loop drive flow for the same core flow in percent or rated recirculation flow for single-loop operation. The value is zero for two-loop operation.
Proposed TS/0029q/168 ,
2.1 SAFETY LIMITS BASES 2.0 The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated tran-sients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07 for two-loop operation and 1.08 for single-loop operation. These limits represent a conservative margin relative to the conditions required to maintain fuel .
cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some cerrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel. cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.
The evaluations which justify normal operation, abnormal transient, and accident analyses for two-loop operation are discussed in detail in Reference
3. Evaluation for single-loop operation demonstrates that two-loop transient and accident analyses are more limiting than single loop, Reference 4.
2.1.1 THERMAL POWER (Low Pressure or Low Flow)
The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10' lbs/hr, ;
bundle pressure drop is nearly independent of bundle power and has a l value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head i will be greater than 28 x 10' lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembTy critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of
, RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
HATCH - UNIT 2 B 2-1 Proposed TS/0027q/168
SAFETY LIMITS BASES (Continued) 2.1.2 THERMAL POWER (High Pressure and High Flow)
The fuel cladding 1 agrity Safety Limit is set such that no fuel damage is calculated to ccur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable dur-ing 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 procedures 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 CPR 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 determined using the General Electric Thermal Analysis Basis, GETAB(", which is a statistical model that combines all of the uncertainties in operating parameters and the pro-cedures used to calculate critical power. The probability of the occur-rence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L), GEXL correlation.
The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.
1 HATCH - UNIT 2 8 2-2 Proposed TS/0027q/168 !
I SAFETY LIMITS BASES (Cpntinued) 2.1.2 THERMAL POWER (High Pressure and High Flow) (Continued) l The required input to the statistical model are the uncertainties listed in Bases Table B 2.1.2-1, the nominal values of the core para-meters listed in Bases Table B 2.1.2-2.
The bases for the uncertainties in the core parameters are given in NE00-20340'2', and the basis for the uncertainty in the GEXL correlation is given in NED0-10958'" . 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 Hatch - Unit 2 during any fuel cycle would not be as severe as the dis-tribution used in the analysis. The method used to handle the uncertainty in the statistical analysis to determine the MCPR cladding integrity Safety Limit for single-loop operation is based on Reference 3, as described in Reference 4. The ; ssure Safety Limits are arbitrarily selected to be the lowest transient ov aressures allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
HATCH - UNIT 2 8 2-3 Proposed TS/0027q/168
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2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS
~
1 The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each pa rameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant ~ system are prevented from exceeding their Safety Limits. Operation with a trip set less conservative than its Trip Setpoint, but within its specified Allowable Value, is acceptable on the basis that each !
Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
1. Intermediate Range Monitor, Neutron Flux The IPM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus, as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap w;th both the APRM and SRM systems.
The most significant source of reactivity changes during the power increase i
are due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed, Section 7.5 of the FSAR. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM's are not yet on scale. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 1%
1 of RATED THERMAL POWER, thus maintaining MCPR above the fuel cladding integrity"~ ~
Safety Limit. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM. -
2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15/125 divisions of full scale neutron flux provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and 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 by the RSCS and RWM.
HATCH - UNIT 2 B 2-9 Proposed TS/0027q/168
4 I
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Average Power Range Monitor (Continued) '
Of all the possible sources of reactivity input, uniform control rod with-drawal is the most probable cause of significant power increase. 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 amount, the rate of power rise is very slow. Gen-erally the heat flux is in near equilibrium with the fission rate. In an i
assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown beforg.the power could exceed the Safety Limit. The 15% neutron flux trip remains active until the mode switch is placed in the Run position.
The APRM flux scram trip in the Run mode consists of a flow referenced simulated thermal power scram setpoint and a fixed neutron flux scram set-point.
The APRM flow referenced neutron flux signal is passed through a filtering network with a time constant which is representative of the fuel
dynamics. This provides a flow referenced signal that approximates the average heat flux or thermal power that is developed in the core during transient or steady-state conditions. The flow independent portion of this scram setpoint must be adjusted downward during single-loop operation to account for lower core flow with respect to two-loop operation with the same
! drive flow.
The APRM flow referenced simulated thermal power scram trip setting for two-loop and single-loop operation is adjustable up to 113.5% of RATED THERMAL POWER. This reduced flow referenced trip setpoint will result in an earlier scram during slow thermal transients, such as the loss of 100*F feedwater heating event, than would result with the 118% fixed neutron flux j
scram trip. The lower flow referenced scram setpoint therefore decreases the 4 severity, ACPR, of a slow thermal transient and allows lower operating limits _
if such a transient is the limiting abnormal operational transient during a certain exposure interval in the fuel cycle.
The APRM fixed neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux. This scram setpoint 4
scrams the reactor during fast power increase transients if credit is not taken for a direct (position) scram, and also serves to scram the reactor if credit is not taken for the flow referenced simulated thermal power 4
scram.
The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of
) unnecessary shutdown.
1
. l 1
HATCH-UNIT 2 B 2-10 Proposed TS/0027q/168
LIMITING SAFETY SYSTEM SETTING BASES (Continued) __._
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Turbine Control Valve Fast Closure', Trip Oil Pressure-Low (Continued) pressure switches whose contacts form the one out-of-two-twice logic input to the Reactor Protection System. This trip setting, a nominally 50% greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve. No significant change in MCPR occurs. Relevant transient analyses are discussed in Section 15 of the Final Safety Analysis Report. This scram is bypassed when turbine steam flow is below that corresponding to 30% of RATED THERMAL POWER, as measured by turbine first stage pressure.
11. Reactor Mode Switch In Shutdown Position The reactor mode switch Shutdown position trip is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.
12. Manual Scram The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
2.3 REFERENCES
1. " General Electric BWR Thermal Analysis Bases (GETAB) Data, Correlation and Design Application," NEDE-10958-P-A and NE00-10958-A, January 1977.
2. General Electric " Process Computer Performance Evaluation Accuracy," _
NED0-20340 and Amendment 1, NED0-20340-1, June 1974 and December 1974, respectively.
3. " General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDO-24011-P-A.
4. "Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"
NED0-24205, August 1979.
1 HATCH-UNIT 2 B 2-13 Proposed TS/0027q/168
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 ALL AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall be equal to or less than the applicable APLHGR limit, which is a function of fuel type and AVERAGE PLANAR EXPOSURE. The APLHGR limit is given by the applicable rated power, rated-flow limit taken from Figures 3.2.1-1 through 3.2.1-9, multiplied by the smaller of either:
a. The factor given by Figure 3.2.1-10, or
b. The factor given by Figure 3.2.1-11.
APPLICABILITY: CONDITION 1, when THERMAL POWER 2 25% of RATED THERMAL POWER.
ACTION:
With an APLHGR exceeding the applicable limits, initiate corrective action within l 15 minutes and continue corrective action so that the APLHGR meets 3.2.1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the applicable limit:
l
a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,
b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been -
established, and
c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.
HATCH - UNIT 2 3/4 2-1 Proposed TS/0021q/168
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POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 ALL MINIMUM CRITICAL POWER RATIOS (MCPRs) for two loop operation, l shall be equal to or greater than the MCPR operating limit (0LMCPR), which is a function of average scram time, core flow, and core power. For 25%
s Power < 30%, the OLMCPR is given in Figure 3.2.3-5. For Power 2 30%,
the OLMCPR is the greater of either:
a. The applicable limit determined from Figure 3.2.3-4, or
b. The appropriate K given by Figure 3.2.3-5, multiplied by the appropriate limit from Figure 3.2.3-1, 3.2.3-2, or 3.2.3-3, where:
t = 0 or
ave *B , whichever is greater,
. *A ~
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t A = 1.096 sec (Specification 3.1.3.3 scram time limit to notch 36),
t g = 0.834 + 1.65 N
3 W (0.059),
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=
Ei=1 "i' i t,y, n
EN g
.1 = 1 .
n= number of surveillance tests performed to date in cycle, Ng = number of active control rods measured in the i th surveillance test, t g = average scram time to notch 36 of all rods measured in the i th surveillance test, and Ny = total number of active rods measured in 4.1.3.2.a.
For single-loop operation, the MCPR limit is increased by 0.01 over the comparable two-loop value.
APPLICABILITY: CONDITION 1, when THERMAL POWER 2 25% RATED THERMAL POWER HATCH - UNIT 2 3/4 2-6 Proposed TS/00219/168
3/4.2.3 MINIMUM CRITICAL POWER RATIO (CONTINUED)
SURVEILLANCE REQUIREMENTS ACTION:
With MCPR less than the applicable limit determined from Specification '
3.2.3.a, or 3.2.3.b for two-loop or single-loop operation, initiate corrective l action within 15 minutes and continue corrective action so that MCPR is equal to or greater than the applicable limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than or equal to 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
4.2.3 The MCPR limit at rated flow and rated power shall be determined for each type of fuel (8X82, P8X8R, and 7X7) from Figures 3.2.3-1, 3.2.3-2, and 3.2.3-3 using:
a. t = 1.0 prior to the initial scram time measurements for the cycle performed in accordance with Specification 4.1.3.2.a, or
b. t as defined in Specification 3.2.3; the determination of the limit must be completed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 4.1.3.2.
MCPR shall be determined to be equal to or greater than the applicable limit:
a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,
b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and
c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R0D PATTERN for MCPR.
l HATCH - UNIT 2 3/4 2-7 Proposed TS/0021q/168
l TABLE 3.3.5-2 gt CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPolNTS F1 DC 1RI* FUNCTION TRIP SETPolNT ALLOWABLE VALUE
1. APRM
-4
a. Flow Referenced Simulated 4
The rma l Power - Upsca le l no b. Inope ra t ive 5 NA (0.58 W + 50% - 0.58 AW) sNA(0.58 W + 50% - 0.58 AW) l-
c. Downscale 2 3/125 or rull scale 2 3/125 or full scale
, e
d. Neutron Flux - High, 12% s 12/125 or rull scale 5 12/125 or rull scale
2. ROD BLOCK MONITOR
a. Upscale
1) Low Trip Setpoint (LTSP) 5 115.1/125 or full scale 5 115.5/125 or full scale
2) Intermediate Trip Setpoint (ITSP) 5 109.3/125 or full scale 5 109.7/125 of full scale
3) High Trip Setpoint (HTSP) s 105.5/125 or full scale 5 105.9/125 or full scale
b. Inope ra t ive NA NA l c. Downscale 2 94/125 or rull scale 2 93/125 or ruli scale
d. Power Range Setpoint'* 3
1) Low Power Setpoint (LPSP) s 27% of RATED THERMAL POWER 5 29% of RATED THERMAL POWER
2) Intermediate Power Setpoint
.(IPSP) s 62% of RATED THERMAL POWER 5 64% of RATED THERMAL POWER
3) High Power Setpoint (HPSP) 5 82% or RATED THERMAL POWER 5 84% or RATED THERMAL POWER
e. RBM Bypass Time Delay 5 2.0 sec 5 2.0 sec W (tda) **8 5
W
3. SOURCE RANCE MONITORS jm a.
Detector not full in NA NA C) b. Upscale 5 1 x 10' cps s 1 x 10' cps
c. Inopera t ive NA NA
d. Downscale 2 3 cps 2 3 cps 1
4
+
Proposed TS/0021g/168 l
TABLE 3.3.5-2 (Continued) 3g 1
CONTROL ROD WITHDRAWAL BLOCM INSTRUMENTATION SETPolNTS 3$
E2 TRIP FUNCTION TRIP SETPOINT &LLOWABLE VALUE
' 4. INTERMEDIATE RANCE MONITORS C
3E a. Detector not full in NA Il b.
c.
Upscale Inope ra tive 5 108/125 or full scale NA 5 108/125 or rull scale S3 NA NA
d. Downscale 2 5/125 or full scale 2 5/125 or full scale
5. SCRAM DISCHARGE VOLUME
a. Water Level-High 5 36.2 gallons 5 36.2 gallons NOTES:
a. See Table 2.2.1-1 for definitions or W and AW.
b. There are three upscale trip levels. l ra nge. All RBM trips are automatically bypassed below the low power setpoint.Only oneTheisupscale applicable over LTSP is aapplied specirled operating core ther between the low power setpoint and the intermediate power setpoint. The upscale ITSS is applied between the intermediate power setpoint and the high power setpoint. The HTSP is applied above the high power setpoint, c.
Power ranges.range setpoints The power control signal to theenforcement or appropriate RBM is provided upscale trips over the proper core thermal power by the APRM.
La
); d.
RBM bypass time delay is set low enough to assure minimum rod movement while upscale trips are bypassed.
Y s
S Proposed TS/0021g/168
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LO0pS LIMITING CONDITION FOR OPERATION-3.4.1.1 At least one recirculation loop of the reactor coolant system shall be in operation with its recirculation pump operating and the associated pump discharge valves OPERABLE, and
a. With only one recirculation loop in operation, the Functional Units 2.b of Table 2.2.1-1 and 1.a of Table 3.3.5-2, the limits on APLHGR 1
in Section 3/4.2.1 and MCPR in Section 3/4.2.3 shall be in effect.
b. With only one recirculation loop in operation, the limit specified in Figure 3.4.1.1-1 shall be in effect.
APPLICABILITY: CONDITIONS 1* and 2*.
ACTION:
a. With no recirculation loops in operation, place the rector mode switch in the HOT SHUTOOWN position.
b. With requirements of Specification 3.4.1.1.a not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the removal of one recirculation loop from service, place the unit in the HOT SHUTDOWN condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in COLD SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
c. With only one recirculation loop in operation and the Unit in the Operation Not Allowed Region specified in Figure 3.4.1.1-1, initiate action within 15 minutes to place the Unit in the Operation Allowed Region in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Otherwise, place the i reactor in the HOT SHUTDOWN condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . .
SURVEILLANCE REQUIREMENTS 4.4.1.1 Each pump discharge valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel:
a. Each startup** prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER, and
b. During each COLD SHUTOOWN which exceeds 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .**
.
See Special Test Exception 3.10.4.
If not performed within the previous 31 days.
HATCH - UNIT 2 3/4 4-1 Proposed TS/0023q/169
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM
<ECIRCULATION LOOPS SURVEILLANCE REQUIREMENTS 4.4.1.2 With only one recirculation loop in operation, verify that the reactor operating conditions are outside the Operation Not Allowed Region in Figure 3.4.1.1-1:
a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b. Whenever THERMAL p0WER has been changed by at least 5% of RATED THERMAL POWER and steady state conditions have been reached.
~
HATCH - UNIT 2 3/4 4-la Proposed TS/0023q/169
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HATCH UNIT 2 3/4 4-lb
REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps corresponding to the operating loop (s) shall be OPERABLE.
APPLICABILITY: CONDITIONS 1 and 2.
ACTION:
a. All jet pumps corresponding to the operating loop (s) shall be OPERABLE with at least one of the following requirements:
1. For any specific flow condition, each individual jet pump flow shall not differ by more than 10% of the average loop jet pump flow from the normal range
of average loop jet pump flows experienced for those flow conditions, or
2. For any specific core flow condition, each individual jet pump diffuser to lower plenum differential pressure (D/P) shall not differ by more than 20% of the average loop D/P from the normal range
of average loop jet pump D/Ps experienced for those flow conditions.
b. With one or more jet pumps exceeding the above requirements, evaluate the reason for the deviation, and be in HOT SHUTOOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in the circumstances that one or more jet pumps are verified to be inoperable.
SURVEILLANCE REQUIREMENTS 4.4.1.2 Each of the above required jet pumps shall be demonstrated OPERABLE"'
prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER; following i
recirculation restarts; following any unexpected or unexplained change in core flow, jet pump loop flow, recirculation pump flow, or core plate differential pressure; and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by recording jet pump loop flows, recirculation pump flows, recirculation pump speeds, and i individual jet pump flows (D/P) and verifying that neither of the following I conditions occur:
Normal expected operating range is based on data obtained from operating experience.
HATCH - UNIT 2 3/4 4-2 Proposed TS/0023q/169
SURVEILLANCE REQUIREMENTS
a. The recirculation pump flow / speed ratio deviates more than 5% from the normal range,* or
b. The jet pump loop flow / speed ratio deviates more than 5% from the normal range." -
If any required jet pump fails to meet either or both of the above Surveillance Requirements, review the jet pump operability as defined in the ACTION statement for Section 3.4.1.2 and in BASES Section 3/4.4.1.2.
4 1
~
I l
1 Normal expected operating range is based on data obtained from operating i
. experience.
HATCH - UNIT 2 3/4 4-2a Proposed TS/0023q/169 1
l
1 i !
i l
1 -
l l
(This page is intentionally left blank.)
a 4
4 i .
I i,
'l d
1 1 -
HATCH - UNIT 2 3/4 4-2b Proposed TS/0023q/169
- . , -- - - - --m, ,, n. ., -- - ,,-.g . . _-- - - -,-- - . , , , ,, , =,, - - ,. - .,e- -- - - - -,; +- -
SPECIAL TEST EXCEPTION RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION b
3.10.4 The requirements of Specification 3.4.1.1 that a recirculation loop (s) be in operation may be suspended for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during the performance of the Startup Test Program and PHYSICS TESTS.
APPLICABILITY: CONDITIONS I and 2.
ACTION:
With the above specified time limit exceeded, actuate the manual scram.
SURVEILLANCE REQUIREMENTS 4.10.4 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at least once per hour during the Startup Test Program and PHYSICS TESTS.
HATCH - UNIT 2 3/4 10-4 Proposed TS/0022q/169
l l
i REACTIVITY CONTROL SYSTEMS
BASES A
3/4.1.3 CONTROL RODS The specifications of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained. (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) the potential effects of the rod drop accident are limited. The ACTION statements permit variations
, from the basic requirements, but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be 4 kept to a minimum. The requirements for the various scram time measurements l ensure that any indication of systematic problems with rod drives will be !
i a
investigated on a timely basis.
, Damage within the control rod drive mechanism could be a generic
problem; therefore, with a control rod imovable because of excessive friction or mechanical interference, operation of the reactor is limited to l a time period which is reasonable to determine the cause of the
inoperability and at the same time prevent operation with a large number of j inoperable control rods.
j Control rods that are inoperable for other reasons are permitted to be 4
taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.
The number of control rods permitted to be inoperable could be more than
' the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor 4
must be shutdown for investigation and resolution of the problem.
The control rod system is analyzed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the fuel cladding integrity MCPR Safety Limit during the limiting power transient analyzed in 4
Section 15 of the FSAR. This analysis shows that the negative reactivity 1
rates resulting from the scram with the average response of all the drives as l
given in the specifications provide the required protection and MCPR remains
' greater than the fuel cladding integrity MCPR Safety Limit. The occurrence of l scram times longer than ,those specified should be viewed as an indication of a
! systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of
time with a potentially serious problem.
Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram
)
I j HATCH - UNIT 2 8 3/4 1-2 Proposed TS/0024q/169 i
i
- , - - . - _ - - - - - - . - - - - . - _ _ , . - - - . - _ - - - - - . _ - - - . . - . - . - . -. .- - ~
4 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding i
temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in the Final Acceptance Criteria ;
(FAC) issued in June 1971 considering the postulated effects of fuel pellet !
densification. These specifications also assure that fuel design margins are maintained during abnormal transients. '
1
] 3/4.2.1 AVERA6E PLANAR LINEAR HEAT GENERATION RATE j
This specification assures that the peak cladding temperature following l the postulated design basis loss-of-coolant accident will not exceed the limit j specified in 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 i
the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming an LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.
This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor.
The Technical Specification APLHGR is this LHGR of the highest powered rod
. divided by its local peaking factor. The limiting value for APLHGR is shown 2
in the figures for in Technical Specification 3/4.2.1. For single-loop operation, Reference 1 requires a 0.75 multiplication factor to 8X8R and P8X8R bundles.
The calculational procedure used to establish the APLHGR shown in the 4
figures in Technical Specification 3/4.2.1 is based on a loss-of-coolant ,
accident analysis. Tne analysis was performed using General Electric (GE) '
calculational models which are consistent with th? requirements of Appendix K_ i to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.
A flow dependent correction factor incorporated into Figure 3.2.1-10 is i
applied to the rated conditions APLHR to assure that the 2200*F PCT limit is complied with during a LOCA initiated from less than rated core flow. In i i addition, other power and flow dependent corrections given in Figures 3.2.1-10 and 3.2.1-11 are applied to the rated conditions to assure that the fuel )
thermal-mechanical design criteria are preserved during abnormal transients i
initiated from off-rated conditions. l i
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in bases Table 8 3.2.1-1. Further discussion j of the APLHGR limits is given in Reference 2.
l l l
! HATCH - UNIT 2 8 3/4 2-1
Proposed TS/0024q/169 I " - ~ ~ " ~ ~ "
l l
POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS This section 't leted.
, 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.07 for two-loop operation and 1.08 for single-loop operation, and an analysis of abnormal operational transients as described in References 1 and 3. 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 time during the transient assuming instrument trip setting as given in Specification 2.2.1.
To assure that the fuel cladding integrity Safety Limits are not exceeded during any anticipated abnornal operational transient, the most limiting transients have been analyzed to determine which results in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
i HATCH - UNIT 2 B 3/4 2-3 Proposed TS/0024q/169
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)
The purpose of the MCPR , and the Kp of Figures 3.2.3-4 and 3.2.3-5, f
respectively is to define operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power, the required MCPR is the larger value of the MCPRf and MCPRp at the existing core flow and power state. The MCPRfs are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.
The MCPR f s were calculated such that for the maximum core flow rate and the corresponding THERMAL POWER along the 105% of rated steam flow control
line, the limiting bundle's relative power was adjusted until the MCPR was j slightly above the Safety Limit. Using this relative bundle power, the MCPRs j were calculated at different points along the 105% of rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPRf .
The core power dependent MCPR operating limit, MCPR p , is the rated power i
and rated flow MCPR operating limit multiplied by the Kp factor given in Figure 3.2.3-5.
The K p s are established to protect the core from transients other than core flow increases, including the localized event such as rod withdrawal error. The l Kp s were determined based upon the most limiting transient at tne given core power level. For further infonnation on MCPR operating limits for of f-rated conditions.
See Reference NEDC-30474-P.(a) l l
i t
)
i-I l
HATCH - UNIT 2 B 3/4 2-4 Proposed TS/0024q/169
- ,,-a, .,w--- , - - . - - - - , - - - , - - - - - -,ra--e,-,,,,-- ,-....-.m w.m.- -e--, ----m ,.- .- e,---
POWER DISTRIBUTION LIMITS BASES
References:
1. " General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDE-24011-P-A.
2. " Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Edwin I. Hatch Nuclear Plant, Units 1 and 2," NEDC-30474-P, December 1983.
3. "Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"
NEDO-24205, August 1979.
l HATCH - UNIT 2 8 3/4 2-6 Proposed TS/0024q/169
3/4.4 REACTOR COOLANT SYSTEM BASES ,
l 3/4.4.1.1 RECIRCULATION SYSTEM l
Operation with a reactor coolant recircula' tion loop inoperable is allowed, provided that adjustments to the flow referenced scram and APRM rod l block setpoints, MCPR cladding integrity Safety Limit MCPR Operating Limit, i and MAPLHGR limit are made. An evaluation of the performance of the ECCS with <
single-loop operation has been performed and determined to be acceptable, '
Reference 1. The maximum uncovered time results in a reduction factor to the l MAPLHGR limit of 0.75. To account for increased uncertainties in the total core flow and TIP readings when operating with a single recirculation pump, a 0.01 increase is applied to the MCPR cladding integrity Safety Limit and MCPR Operating Limit over the comparable two-loop values. The flow referenced simulated thermal power setpoint for single-loop operation is reduced by the amount of maw, where a is the flow reference slope for the rod block monitor and AW is the largest difference between two-loop and single-loop effective drive flow when the active loop indicated flow is the same. This adjustment is necessary to preserve the original relationship between the scram trip and actual drive flow.
The possibility of experiencing limit cycle oscillations during single-loop operation is precluded by restricting the core flow to greater than or equal to 45% of rated when core thermal power is greater than the 80%
rod line. This requirement is based on General Electric's recommendations ,
contained in SIL-380, Revision 1, which defines the region where the limit cycle oscillations are more likely to occur.
3/4.4.1.2 JET PUMPS l
An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design basis Loss-of-Coolant Accident by increasing the blowdown l area and eliminating the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable is -
necessary.
One of the acceptable procedures for jet pump failure surveillance identified in NUREG/CR-3052, Reference 2, was adopted for Hatch Unit 2. The surveillance is performed to verify that neither of the following conditions occur:
(a) The recirculation pump flow / speed ratio deviates by more than 5%
from the normal range, or (b) The jet pump loop flow / speed ratio deviates by more than 5% from the normal range. ,
I If either criterion is failed, then the procedure calls for comparing either the individual jet pump flows or the individual jet pump diffuser to ;
lower plenum differential pressures to the criteria of the Limiting Conditions for Operation (LCO). If the LCO is not satisfied and pump speed is less than l l HATCH - UNIT 2 83/44-1 Proposed TS/0024q/169 I
BASES 60% rated, it may be necessary to inc'Fease pump speed to above 60% of rated and to repeat the measurements before declaring a jet pump inoperable. In this case, it is recommended that close monitoring and increasing recirculation pump speed be performed only,1f the criteria are exceeded by an amount to be determined from previous plant operating experience.
3/4.4.1.3 IDLE RECIRCULATION LOOP STARTUP When restarting an idle pump, the discharge valve of the idle loop is required to remain closed until the speed of the faster pump is below 50% of its rated speed to provide assurance that when going from one- to two-loop operations, excessive vibration of the jet pump risers will not occur.
In order to prevent undue stress on the vessel nozzles and bottom head region the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop. Since the coolant in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference were greater than 145'F. The loop temperature must be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.
3/4.4.2.1 SAFETY / RELIEF VALVES The reactor coolant system safety valve function of the safety-relief valves operate to prevent the system from being pressurized above the Safety Limit of 1325 psig. The system is designed to meet the requirements of the ASME Boiler and Pressure Vessel Code,Section III, for the pressure vessel, and ANSI B31.1, 1975 Code, for the reactor coolant system piping.
The capacity of the safety-relief valves is based on the full MSIV closure
. transient with failed trip scram, position switches, as described in Supplement 5.A of the FSAR, Section 5.A.6.
l Demonstration of the safety-relief valve lift settings will occur -
only during shutdown and will be performed in accordance with the
provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
i 1
l l
i 1
I i
j HATCH - UNIT 2 8 3/4 4-la Proposed TS/0024q/169,
BASES 3/4.4.2.2 LOW-LOW SET SYSTEM The low-low set (LLS) system lowers the opening and closing setpoints on four preselected safety / relief valves (S/RVs). The LLS system lowers the setpoints 'after any S/RV has opened at its normal steam pilot setpoint when a concurrent high reactor vessel steam dome pressure scram signal is present. The purpose of the LLS is to mitigate the induced high frequency loads on the contain-ment and thrust loads on the SRV discharge line. The LLS system increases the amount of reactor depressurization during an S/RV blowdown because the lowered LLS setpoints keep the four selected LLS S/RVs open for a longer time. The high reactor vessel steam dome pressure signal for the LLS logic is provided by the exclusive analog trip channels. The purpose of installing special ded-l icated steam dome pressure channels is to maintain separation from the RPS high pressure scram functions.
l HATCH - UNIT 2 B 3/4 4-lb Proposed TS/00249 /169 l
l
BASES 3/4.
4.9 REFERENCES
1. "Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"
NEDO-24205, August 1979.
2. NUREG/CR-3052, " Closeout of IE BULLETIN 80-07: BWR Jet Pump Assembly Failure," Published November 1984.
HATCH - UNIT 2 8 3/4 4-7 Proposed TS/0024q/169 i l
_

v t e Letter Sequence Other
TAC:61900, (Open) TAC:61901, (Open)
Initiation Request, Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML20203H787, ML20206K007, ML20236D318, ML20236E256
MONTHYEAR1986-07-22 [Table View]

ML20206K007

Text

References:

Subject:

2.3 REFERENCES

2.3 REFERENCES

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References:

4.9 REFERENCES

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