ML19208C501
| ML19208C501 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/06/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19208C502 | List: |
| References | |
| NUDOCS 7909260509 | |
| Download: ML19208C501 (24) | |
Text
m D
,{pu nua h UNITED STATES uq y '), m j
NUCLEAR REGULATORY COMM10SION
.- y WASHINGTON, D. C. 20555
%,...../
GEORGIA POWER C0"PANY OGLETHORPE ELECTRIC MEMBERSHIP CORPOPATIOil MUNICIPAL ELECTRIC ASSOCIATI0fl 0F GEORGIA CITY OF DALTON, GEORGIA DOCKET N0. 50-321 EDWIN I. HATCH NUCLEAR PLAf!T UilIT NO. 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 69 License No. DPR-57 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Georgia Power Company et al., (the licensee) dated March 22, 1979 as amended May 11 and 16, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the ccmmon defense and security or to the health and safety of the public, and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C. (2) of Facility Operating License No. OPR-57 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 69, are hereby incorporated in the license. The licensee shall operate the facility in accordance
/
with the Technical Specifications.
1012 213 97 7909260
-1
2 3.
This license amendment is effective as of the date of its issuance.
FOR THE fiUCLEAR REGULATORY COMT!ISSI0ft
~bd4 u,&
Yhomas(A Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: August 6, 1979 1012 214
ATTACHMENT TO LICENSE AMENDMENT N0. 69 FACILITY CPERATING LICENSE N0. DPR-57 DOCKET NO. 50-321 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
REMOVE INSERT iX IX 1.1-1 1.1-1 1.1-2 1.1-2 1.1-12 1.1-12 1.1-13 1.1-13 Fig.
1.1-1 Fig.
1.1-1 1.2-3 1.2-3 1.2-5 1.2-5 3.1-4 3.1-4 3.1-7 3.1-7 3.1-11
- 3.1-11 3.1-12 3.1-12 3.2-27 3.2-27 3.2-28
- 3.2-28 3.2-29
- 3.2-29 3.2-30 3.2-30 3.6-20 3.6-20 3.11-1 3.11-1 3.11-2 3.11-2 3.11-4 3.11-4 Fig.
3.11-2 (Sheet 1)
(deleted)
Fig.
3.11-2 (Sheet 2)
(deleted) 5.0-1 5.0-1
- Overleaf 1012 215
l I
LIST OF FIGURES t
Ficure Title 1.1-1 Core Thermal Power Safety Limit Versus Core Flow Rate 2.1-1 Reactor Vessel Water Levels 4.1-1 Graphical Aid for the Selection of an Adequate Interval Between Tests 4.2-1 System Unavailability 3.4-1 Sodium Pentaborate Solution Volume Versus Concentration Requirements 3.4-2 Sodium Pentaborate Solution Temperature Versus Concentration Requirements 3.5-1 Change in Charpy V Transition Temperature Versus Neutron Exposure 3.6-2 Minimum Temperature for Inservice Hydrostatic and Leak Tests 3.6-3 Minimum Temperature for Mechanical Heatup or Cooldown Following Nuclear Shutdown 3.6-4 Minimum Temperature for Core Operation (Criticality) 3.11 1 (Sheet 1) Limiting Value for APLHGR (Fuel Type 3) 3.11-1 (Sheet 2) Limiting Value for APLHGR (Fuel Types 1 and 2)
I 3.11-2 deleted l
l 3.11-3 K Factor f
6.2.1-1 Offsite Organization 6.2.2-1 Unit Organization 1012 216 kencment No jHI, 69 ix
SAFETt LIMITS LIMITING 3A?ETY SYSTEM SETTINGS 1.1 FUEL CLACDING INTEGRITY 2.1 FUEL CLACDING INTEGRITY Acolicability Acolicability The Safety Limits established to pre-The Limiting Safety System Settings serve the fuel cladding integrity apply apply to trip settings of the instru-to those variables which monitor the ments and devices wnicn are provided to fuel thermal behavior, prevent the fuel cladding integrity Safety Limits from being exceeded.
Objective Objective The objective of the Safety Limits is The objective of the Limiting Safety 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.
Scecifications Soecifications A. Reactor Pressure > 800 esia and Core A. Trio Settinos Flow > 10% of Rated The limiting safety system trip set-The existence of a minimum critical tings shall be as specified below:
power ratio (MCPR) less than 1.07 shall constitute violation of the
- 1. Neutron Flux Trio Settinos fuel cladding integrity safety limit.
- a. IRM Hich Hich Flux Scram Trio Setti no B. Core Thermal Power Limit (Reacter Tne IRM flux scram trip setting Pressure < 800 osia) shall be < 120/12S of full scale.
When tne reactor pressure is < 800
- b. APRM Flux Scram Trio Settina psia or core flow is less thaii10% of (Refuel or Start & Hot Stancbv rated, the core themal power shall Mcae) not exceed 25". of rated themal power.
Wnen the Mode Switch is in the REFUEL or START & HOT STANCBY position, the APRM flux scram C. Power Transient trip setting shall be < 15/125 of full scale (i.e., ~< 155 of rated
~
To ensure that the Safety Limit estab-thermal power).
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 Settino (Run MoceJ excected scram signal. The Safety Limit shall be assumed to be exceeded (1) Flow Referenced Neutron Flux when scram is accomolished by a means Scram Trip Setting other than the expected scram signal.
When the Mode Switch is in the RUN position the APRM flow referenced flux scram trip setting shall be:
Amendment No. 27, 38, 42, 32, 69 1012 217 1.1-1
SAFEiY LIMIT 5 LIMITING 5AFET( SYSTEM 5tiTINGS 1.1.D Reactor Water Level (Hot or Cold 2.1.A.1.C APRM Flux Scram Trio Snutcown Conci:1on Settincs tRun toce) (Continued)
Whenever the reactor is in tne Hot 5 1 0.66 W + 54".
or Cold Shutdown Condition with irradiated fuel in the reactor vessel, where:
the water level shall be > 378 inches above vessel invert when fuel is S = Setting in percent of seated in the core.
rated themal power (2436 MWt)
W = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate. equals 34.2 x 100 lb/hr)
In the e,ent of operation with a maximv. total peaking factor (MTP ) greater than the design vriue, the setting shall be modified as follows:
1 (0.66 W + 54%) h S
where:
MTPF = The value of the existing maximum total peaking factor A = 2.60 for 7x7 fuel 2.42 for 8x8 fuel 2.48 for 8x8R fuel For no combination of loco. recir-culation ficw rate and core thermal ocwer shall the APRM flux scram trip setting be allowed to exceed 117*. of rated thermal power.
_..eillance recuirements for MTPF are given in Specification 4.1.3.
(2)
Fixed High Neutron Flux Scram Trip Setting When the Mode Switch is in the RUN position, the APRM fixed high flux scram trip setting shall be:
S 1 120% Power 1012 218 Amendment No. 27, /2, 32, 53, 69 i.1-2
BASES FOR LIMITIriG SAFETY SYSTEM SETTIriGS 2.1. A. l.a. IRM Flux Scram Trio Setting (Continued) tism was taken in tnis analysis by assuming that the IRM channel closer to the withdrawn rod is bypassed. The results of this analysis show tnat the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above 1.07.
Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continues withdrawal of control rods in sequence and provides backup protection for the APRM.
For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accomodate anticipated maneuvers asso-ciated with power plant startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns. are constrained to be unifom by operating procedures backed up by the rod worth minimizer and the Rod Sequence Control System. Worth of individual rods is very low in a unifem rod pattern. Thus, of all possible sources of reactivity input, unifom control rod withdrawal is the most probable cause of sig-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 flux is in near equilibrium with the fission rate.
In an assumed unifom rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 825 psig.
- c. APRM Flux Scram Trio Settinos (Run Model The APRM Flux scram trip in the run mode consists of a flow referenced scram setpoint and a fixed high neutron flux scram setooint. The APRM flow referenced neutron flux signal is passed through a filtering 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 ccre during transient or steady-state conditions. This prevents spurious scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses. Examoles 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 and turbine control valve testing.
These flux soikes represent no hazard to the fuel since they are only of a few seconds duration and less than 120*. of rated thermal power.
Amendment No. 27, 38, f2, j52, 69 1.1-12 1012 219
2.1.A.1:c.
APRM Flux Scram Trio Settines (Pun Mode) (Continued)
The APRM ficw referenced scram trip setting at full recirculation flow is adjustable up to 117". of rated power. This reduced flow referenced trip setpoint will result in an ea5 er scram during s1 w thermal 14 transients, such as the loss of 80 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 neutron flux signal 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 tne reactor if credit is not taken for the flow referenced scram.
The flow referenced scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of MTPF and reactor core thermal power. The scram setting is adusted in accordance with the formula in Specification 2.1. A.l.c., when the maximum total peaking factor is greater than 2.60 for 7x7 fuel, 2.42 for 8x8 fuel and 2.48 for 8x8R fuel.
Analyses of the limiting transients show that no scram adjustment is required to assure MCPR > 1.07 when the transient is initiated from the operating MCPR limit.
d.
APRM Rod Block Trio Settino Reactor power level may be varied by movin'g control rods or by varying the recirculation flow rate. The APRM system provides a control rod bicck to prevent rod withdrawal beyond a given point at constant recirculation flow rate, and thus to protect against the condition of a MCPR less than 1.07.
This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow' variable trip setting provides substantial margin from fuel damage, assuming a steacy-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit increases as the ficy decreases for the specified trip setting versus flow relation-ship; therefore, the worst case MCPR which would occur during a steady-state operation is at 108". of rated thermal power because of the APRM rod block trip setting.
The actual power distribution in the core is estab-lished by specified control rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum total peaking factor exceeds 2.60 for 7x7 fuel, 2.42 for 8x8 fuel and 2.48 for 8x8R fuel, thus preserving the APRM rod block safety margin.
2.
Reactor Water Lew Level Scram Trio Settine (LL1)
The trip setting for low level scram is above the bottom of the separator skirt.
This level is > la feet above the too of the active fuel. This level has been used in transient analyses d3aling with coolant inventory decrease.
The results repcrted in FSAR Section 14.3 show that a scram at this levei adecuately crotects the fuel and the pressure barrier.
The scram trip setting is a: proximately 33 inches below the normal acerating range anc is thus adecuate to avoid scuricus scrams.
l 1-t3 1012 220 Amendment No. 27, 34, A2, 32, 38, 69
4 R A T E D T H E R M A L FCi ER = 243 5.".*/,
78.5 X *Go LSS/HR.
RATED CCRE FLOW
=
120 l
APRM FLCW 3
BIAS $ CRAM
/~~~~~
/
/
/
/
/
100
/
/
/
-c NOMINAL EX?!OED
/
FLOW CONTROL LINE
/
-l
=
,e u.
so O
/
U
/
/
.' C a
f
~
s y
/
o
~
60
/
/
)
~
/
E 54 u
5 m
C 8
40 CORE THERMAL PCWER LIMIT 25 20 l
I NATURAL c:ROU:.AT:CN L'NE 0
20 40 60 80'
~~" 100 ~ ~
12 0 ~~ '
0 CORE FLOW RATE (% C? MATED)
FIGUR E f.1-1 CORE THERMAL PCWER SAFETY LIMIT VERSL'S CCRE FLCW RATE An;endment flo.
- 69 si
-4
"+90 fr e.; qpp., -
1012 221
BASES FOR SAFETY LIMITS 1.2 REACTOR COOLANT SYSTEM INTEGRITY The reactor coolant system integrity is an important carrier in the prevention of uncontrolled release of fission products.
It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.
A.
Reacter Vessel Steam Done Pressure
- 1. When Irradiated Fuel is in the Reactor The pressure Safety Limit of 1325 psig as measured by the reactor vessel steam dome pressure indicator is equivalent to 1375 psig at the lowest elevation of the reactor coolant system.
The 1375 psig value is derived from the design pressure of the reactor pressure vessel (1250 psig) and coolant system piping (suction piping:
1150 psig; discharge piping :
1350 psig).
The pressure Safety Limit was chosen as the lower pressure resulting frcm the pressure transients permitted by the applicable design codes:
ASME Boiler and Pressure Vessel Code,Section III for the pressure vessel and USASI 331.1 Code for the reactor coolant system piping.
The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10% over design pressure (110% x 1250 = 1375 psig), and the USASI Code permits pres-sure transients up to 20% over the design pressure (120% x 1150 = 1380 psig; 120% x 1350 = 1602 psig).
The pressure relief system (relief / safety valves) has been sized to meet the overpressure protection criteria of the ASME 3 oiler and Pressure Vessel Code,Section III, Nuclear Vessels.
The details of the overpressure protection analysis showing co=pliance with the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels is provided in the FSAR, Appendix M, Su= mary Technical Report of Reactor Vessel Overpressure Protection. To deter =iae the required steamficw capacity, a parametric study was performed assuming the plant was operating at the turbine generator design condition of 105 percent rated steam flow (10.6 x 106 pounds per hour) with a vessel dome pressure of 1020 psig, at a reactor thermal power of 2537 Mw, and the reactor experiences the worst pressuriza-tion transient. The analysis of the worst overpressure transient, a 3 second closure of all main steam line isolation valves neglecting the direct scram (valve position scram) results in a maximum vessel pressure (bottom) of less than 1375 psig if a neutron flux scram is assumed.
In addicion, l the same event was analyzed to determine the nu=ber of installed valves which would limit pressure to below the code limit.
The results 'of this analysis show that.the elaven installed relief / safety valves were adequate even if assuming the backap neutron flux scram.
l Turbine trip from high power without bypass is the most severe transient resulting di.ectly in a nuclear system pressure increase, assuming direct scram.
This event is presented in Reference 5.
The analysis shows that the peak pressure in the bottem of the vessel is li=ited to 1180 psig.
Peak steam line pressure is 1149 psig, showing adequate protection for this ab-ner=al operational transient.
knendment No. 1, 42, 52, 69 1.2-3
EASES FOR SAFETY LIMITS 1.2.B.
References 1.
ASME Boiler and Pressure Vessel Code Section III.
2.
USASI Piping Code, Section B31.1.
3.
FSAR Section 4.2, Reactor Vessel and Appurtenances Mechanical Design.
4.
FSAR Section 14.3, Analysis of Abnornal Operation Transients.
5.
General Electric Boiling Water Reactor Supplenental Reload Licensing Submittal for the Edwin I. Hatch Nuclear Plant Unit 1 Reload 3, NEDo-24175, January, 1979.
r 1012!
i?i?3 Amendment No.,42, 32, 69 1.2-5
Table 3.1-1 (Cont'd)
Scram Operable ilumber Source of Scram Trip Signal Channels Scram Trip Setting Source of Scram Signal is Required (a)
Required Per to be Operable Except as Indicated Trip S stem Below S
liigh Drywell Pressure 2
1 2 psig tiot required to be operable when primary contaisunent integrity is not required. May be bypassed when necessary during purging for containment inerting or deinerting.
6 Reactor Water Low Level 2
-> 12.5 inches (LL1) (flarrow Range) 7 Scram Discharge Volume liigh 2
1 71 gallons Permissible to bypass (initiates liigh Level control rod block) in order to reset RPS when the Mode Switch is in the REFUEL or SilUTDOWil position.
{'
8 APRM Flow Referenced 2
3 < 0.66We54%
tieutron Flux (ft t to exceed 117%)
Tech Spec 2.1.A.I.c Fixed liigh fleutron 2
S 1 120% Power Flux Tech Spec 2.1.A.I.c b
Inoperative 2
flot Applicable An APRM is inoperable if there are less than two LPRM inputs N
per level or there are less than 11 LPRM inputs to the N
APRM channel N
4 Amendment flo. #, 69
Table 4.1-1 Reactor Protection System (RPS) Instrumentation functional Test, functional Test Minimum Frequency, and Calibration Minimum frequency Scram Instrument Functional Test instrument Calibration flumber Source of Scram Trip Signal Group Minimum Frequency Minimum Frequency j a }_
_(li (c) 1 Mode Switch in ShuTOOWil A
Once/ Operating Cycle flot Applicable 2
Hanual Scram A
Every 3 months flot Applicable 3
IRM liigh liigh Flux C
Once/ Week during refueling Once/ Week and witiiin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of Startup (e)
Inoperative C
Once/ week during refueling Once/ Week and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of Startup(e) 4 liigh Reactor Pressure n
Once/ Month (f)
Every 3 months S
liigh Drywell Pressure A
Once/ Month (f)
Every 3 months 6
Reactor Water Low Level (LL1)
A Once/ Month (f) (9)
Every 3 months 7
Scram Discharge Volume liigh liigh A
Every 3 months (h)
Level 11 APRM Fixed liigh Flux B
Once/ Week (e)
Twice/ Week l
]
Once/ Week (e)
Twice/ Week
{
Downscale B
Once/ Week (e)
Twice/ Week Flow Reference B
Once/ Week (f)
Once/ Operating Cycle N
N Um 15% Flux C
Within 24 llours of Startup (e) Once/ Week Amendment flo.
69
BASES FOR LIMITING CONDITIONS FOR OPERATION 3.1.A.2. Manual Scram The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all codes of reactor operation.
- 3. IRM The bases for the IRM High High Flux Scram Trip Setting are discussed in the bases for Specification 2.1.A.l.a.
Each protection trip system has one more IRM channel than is necessary to meet the mini =um number re-quired. This allows the bypassing of one IRM channel per protection trip system for saintenance, testing or calibration.
- a. High High Flux The IRM system provices protection against excessive power levels and short reactor periods in the snurec and intermediate power ranges. Th e requirement that the IRM's be inserted in the core until the APRM's read 3/125 of full scale er greater assures that there is proper c erlap in the neutron monitoring syste=s and thus, that adequate coverage is pro-vided for all ranges of reactor operation.
A source range monitor (SRM) system is also provided to supply addi-tional neutron level information during start-up but has no scram function (Section 7.5.4 FSAR). Thus, the IRM and APRM systems are required in the Refuel and Start & Hot Standby modes.
In the power range, the APRM system provides the required protection (Section 7.5.7 FSAR). Thus, the IRM System is not required when the APRM's are on scale and the Mode Switch is in the RUN position.
- b. Inoperative When an IRM channel becomes unable to perform its normal monitoring function, the condition is recognized and an inoperative trip results. This trip is given the same logic significance as the upscale trip; thus the faulty channel immediately fails safe by contributing to a potential scram condition.
- 4. High Reactor Pressure High pressure within the nuclear system poses a direct threat of rupture to the nuclear system process barrier. A nuclear system pressure increase while the reactor is operating compresses the steam voids and results in a positive reactivity insertion causing increased core heat generation that could lead to fuel failure and system over-pressurization. A scram counteracts a pressure increase by quickly reducing the core fission heat generation.
The nuclear system high pressure scram setting is chosen slightly above the reactor vessel maximum normal operation pressure to permit normal operation without spurious scrams yet provide a wide margin to the =axi=um allowable nuclear system pressure. The location of the pressure measurement, as compared to the location of highest nuclear system pressure during transients, was also considered in the selection of the high pressure scram setting. The nuclear system high pressure scram works in conjunction with the pressure relief system in preventing nuclear system pressure from exceeding the =axi=um allowable pressure. This s me cuclear system high pressure scram setting also protects 3.1-11
3.1.A.4 Hich Peactor Pressure (Continued) the core from exceeding thermal hydraulic limits as a result of pressure increases for some events that occur when the reactor is operating at less than rated power and flow.
5.
Hich Orywell Pressure Pressure switch instrumentation for the drywell is provided to detect a loss of coolant accident and initiate the core standby cooling equipment.
A high drywell pressure scram is provided at the same setting (12 psig) as the core standby cooling systems initiation to minimize the energy which must be acconnodated during a loss of coolant accident.
The instru-mentation is a backup to the reactor vessel water level instrumentation.
6.
Reactor Water low Level (LLl)
The bases for the Reactor Water Low Level Scram Trip Setting (LL1) are
' discussed in the bases for Specification 2.1. A.2.
T.
Scram Discharoe Volume Hich High Level The control rod drive scram system is designed so that all of the water
~ which is discharged from the reactor by a scram can be accommodated in the discharge piping. A part of this piping is an instrument volume which is the low point in the piping.
No credit was taken for this volume in the. design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation the discharge volume is empty; however, should the discharge volume fill with water, the water discharged to the piping frcm the reactor could not be acccamodated which would result in a slow scram time or partial or no control rod insertion. To preclude this occurrence, level switches have been provided in the instrument volume which scram the reactor when the volume of water reaches 71 gallons. As indicated above, there is suffi-cient volume in the piping to accommodate the scram withcut imoairment of the scram times or amount of insertion of the control rods.
This function shuts the reactor dcwn while sufficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not able to perfonn its function adequately.
8.
APRM Three APPJi instrument channels are provided for each protection trip sys-tem.
APRM's A and E operate contacts in one trip logic and APRM's C and E operate contacts in the other trip logic. APRM's 3, D and F are arranged similarly in the other protection trip system. Each protection trip sys-tem has one more APRM than is necessary to meet the minimum number re-quired per channel. This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration.
~
a.
Flow Referenced and Fixed Hion Neutron Flux The bases for the APRM Flow Referenced and Fixed High Neutron Flux Scram Trip Settings are discussed in the bases for Specification 2.1. A.l.c.
Amendment No. 69
.!! li
!!I!
Table 4.2-2 Check, Functional Test, and Calibration Minimum Frequency for Instrumentation Which Initiates or Controls llPCI Ref.
Instrument Check Instrument Functional Test Instrument Calibration No.
Instrument Minimum Freqiiency Minimum Freqiiency Minimum Frequency (a)
(b)
(c) 1 Reactor Water Level Once/ day (d)
Every 3 months (Yarway) 2 Drywell Pressure None (d)
Every 3 months 3
IIPCI Turbine Overspeed None N/A Once/ operating cycle 4
IIPCI Turbine Exhaust None (d)
Every 3 months c..
Pressure s',
5 IIPCL Pump Sut.clon None (d)
Every 3 months Pressure 6
Reactor Water Level Once/ day (d)
Every 3 months (Narrow Range) 7 IIPCI System Flow None (d)
Every 3 months (Flow Switch)
CD 8
IIPCI Equipment Room Hone (d)
Every 3 months Temperature N
9 deleted g
Nco 10 llPCI Steam Line Pressure None (d)
Every 3 months Amendnent flo. 69
Table 4. 2-2 (Cont 'd)
Ref.
Instrument Check Instrument Functional Test Instrument Calibration No.
Instrument 111nimum Frequency Minimum Frequency Minimum Frequency (a)
(b)
(c) 11 IIPCI Steam Line None (d)
Every 3 months AP (Flow) 12 IIPCI Turbine Exhaust None (d)
Every 3 months Diaphragm Pressure 13 Suppression Chamber Area None (d)
Every 3 months Air Temperature 14 Suppression Chamber Area None (d)
Every 3 months Differential Air Temperature 15 Condensate Storage None (d)
Every 3 months u)
Tank Level 16 Suppression Chamber None (d)
Every 3 months Water Level 17 IIPCI Logic Power None once/ operating cycle None Failure Motor Notes for Table 4.2-2 The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be CD a.
N established between items in Table 4.2-2 and items in Table 3.2-2.
N Nc l f!
ii!$il
9!!
- ii
'iiiii;
-S Notes for Table 4.2-2 (Cont 'd) b.
Instrument functional tests are not required when the instruments are not required to be operable or are trig.,ed.
Ilowever, if functional tests are missed, they shall be performed prior to returning the instrument to an operable status, c.
Calibrations are not required when the instrinnents are not required to be operable, lloweve r, if calibrations are missed, they shall be performed prior to returning the instrument to an operable
- status, d.
Initially once per month or according to Figure 4.1-1 with an interval of not less than one month nor more than three months. The compilation of instrument failure rate data may include data obtained i
from other BWR's for which the same design instrument operates in an environment similar to that i,a
'd o f IINP-1.
The failure rate data must be reviewed and approved by the AEC prior to any change in the' once-a-month frequency.
Ingic system functional tests and simulated automatic actuation shall be performed once each operating cycle for the following:
1.
IIPCI Subsystem 3.
Diesel Cenerator Initiation C~{
2.
IIPCI Subsys tem Auto Isolation 4.
Area Cooling for Engineered I')
Safeguard Systems N.)
1he logic system functional tests shall include a calibration of time relays and timers necessary for
(,4 CD proper functioning of the trip systems.
Table 4.2-3 Check, Functional Tes t, and Calibration Minimum Frequency for Instrumentation Whirli Initiates or Controls RCIC Ref.
Instrument Check Instrument Functional Test Instrument Calibration tio.
Instrument Minimum Frequency Minimum Frequency Minimum Frequency (a)
(b)
(c) 1 Reactor Water Ievel Once/ day (d)
Every 3 months (Yarway) 2 RCIC Turbine Overspeed Electrical /
None N/A Once/ operating cycle Mechanical None N/A Once/ operating cycle 3
RCIC Turbine Exhaust None (d)
Every 3 months p'
Pressure 4
RCIC Pump Suction None (d)
Every 3 months Pressure 5
Reactor Water level Once/ day (d)
Every 3 months (Narrow Range) 6 RCIC System Flow None (d)
Every 3 months (Flow Switch) o 7
RCIC Equipment Room None (d)
Every 3 months Temperature N
8 deleted Nu 9
RCIC Steam Line Pressure None (d)
Every 3 months
~
Amendment No. 69 h
C.
BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE kE03IREMENTS 3.6.G.
Reactor Coolant Leakage (Continued) would grow rapidly. However, the establishment of allowable unidentified leakage greater than that given in Specification 3.6.G on the basis of the data presently available would be premature because of uncertainties asso-ciated with the data. For leakage of the order of 5 gpm, as specified in Specification 3.6.G, the experimental and analytical data suggest a reason-able margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation (Reference FSAR, Question 10.4.2).
Leakage less than the magnitude specified can be detected reasonably in a manner of a few hours utilizing the available leakage detection sche =e, and if the origin cannot be determined in a reasonably short time the plant shall be shutdown to allow further investigation and corrective action.
The total leakage rate consists of all leakage, identified and unidentified which flows to the drywell floor drain and equipment drain sump.
The capacity of the drywell floor sump pumps is 100 gpm and the capacity of the drywell equipment nap pumps is also 100 gpm.
Removal of 25 gpm f rom either of these sumps can be accomplished with considerable margin.
H.
nelief/Safetv Val' is, The pressure re3'.ef system (relief / safety valves) has been sized to meet the overpressure protection criteria of the ASME Boiler and Pressure Vessel Code,Section III, Nuclecr Vessels.
The details of the overpressure protection analysis showing compliance with ASEE,Section XII is provided in the FSAR, Appendix M, Summary Technical Report of Reactor Vessel Overpressure Protection.
To determine the required steamflow capacity, a parametric study was perfor=ed assuming the plant was operating at the turbine-generator design condition of 105 percent rated 6
steam flow (10.6 x 10 pounds per hour) with a vessel dome pressure of 1020 psig, at a reactor thermal power of 2537 Mw, and the reactor experiences the worst pressurization transient.
The reanalysis for Reload-3 (NEDO-24175) of the worst overpressure transient, a 3 second closure of all main steam line isolation valves neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1232 psig if a reutron flux scram is assumed.
Turbine trip from high power without bypass is the most severe transient resulting directly in a nuclear syste= pressure increase, assuming direct scram.
This event is presented in NEDD-24175. The analysis shows that the peak pressure in the bottem of the vessel is limited to 1180 psig. Peak steam line pressure is 1149 psig, showing adequate protection for this worst ab-normal operational transient.
Amendment No. I, A2, 32, 69 m
3.6-20
BASES FOR LIMITING CONDITIONS F05 OPERATION AND SURVEILLANCE REQUIREMENTS 3.11 FUEL RODS 4.11 FUEL RODS Applicability Aoplicability The Limiting Conditions for Operation The Surveillance Requirements apply associated with the fuel rods apply to to the parameters which monitor the those parameters which monitor the fuel rod operating conditions.
fuel rod operating conditions.
Objective Obiective The Objective of the Limiting Condi-The Objective of the Surveillance tions for Operation is to assure the Requirements is to specify the type perfor=ance of the fuel rods.
and frequency of surveillance to be applied to the fuel rods.
Specifications Specifications AO.Aversee Planar Linear Heat Genera-A. Average Planar Linear Heat Genera-tion Rate (APLHGR) tion Kate (APLHGR)
During power operation, the APLEGR The AFLHGR for each type of fuel as for each type of fuel as a function a function of average planar of average planar exposure shall exposure shall be determined daily not exceed the limiting value shown during reactor operation at 2,25%
in Figure 3.11-1, sheets 1 and 2.
rated thermal power.
If at any time during operation it is determined by normal surveillance that the limiting value for APLEGR is being exceeded, action shall be initiated within 15 minutes to re-store operation to within the pre-scribed limits.
If the APLHGR is not returned to within the pre-scribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power with-in 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.
.B. Linear Heat Generation Rate (LHGR)
B. Linear Heat Generation Rate (LHGR)
During power operation, the LEGR as The LHGR as function of core a function of core height shall not height shall be checked daily dur-exceed the limiting value shcwn in ing reactor operation at 3, 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 SR fuel.
If at any time during operation it is determined by normal
, surveillance that the limiting value for LHGR is being exceeded, action shall 1012' i'3 be initiated within 15 mihutes to restore operation to within the prescribed limits.
If the Amendment No. 3J, 32, 69 3.11-1
BASES FOR LIMITING CONDITIONS FOR OPERATION AND SUKVELLI.ANCh KtUU1KtMbNIS 3.ll.B.
Linear Heat Generation Rate (LHCR)
(Continued)
LHGR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated ther.a1 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 re-quired.
C.
Minimum Critical Power Ratio (MCPR) 4.ll.C Minimum Critical Pcuer Ratio (MCPR)
The MCPR limit is specified through-MCPR shall be determined daily out the cycle. From B0C4 to E0C4-during reactor power operation at 2000 MWD /t the MCPR limit is 1.26 1 25% rated thermal power and for 7 x 7, i.24 for 8 x 8, and 1.21 following any change in power for 8 x 8R fuels.* During power level or distribution that would operation, MCPR shall be as abor cause operation with a li=iting at rated power and flow.
If a:
control rod pattern as described any time during operation it is in the bases for Specification detemined by normal surveille ce 3.3.F.
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 thermal power is not required.
For core flows other than rated the MCPR shall be Kf times the MCPR value applicable above, where Kf is as shown in Figure 3.11-3.
- MCPR values for E0C4-2000 MWD /t to E0C4 will be determined after completion of the review of the D.
Reporting Recuirements hardware implementation of E0C Rec' culation Pump Trip feature.
If any of the limiting values iden-tified in Specifications 3.ll.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
}Q}2 4
report will meet the' requirements of this specification.
Amendment No. 57, $2, 69 3.11-2
BASES FOR LIMITNG CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS 3.ll.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 on the l
nnalysis 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 ther=al pcwer, the MTPF would have to be greater than 10 which is precluded by a consi-derable margin when employing any permissible control rod pattern.
C.
Minimum Critical Power Ratic (MCPR)
The required operating limit MCPR as specified in Specification 3.ll.C is derived from the established fuel cladding integrity Safety Limit MCPR of 1.07 and an analysis of abnormal operational transients presented in Reference 7.
Various transient events will reduce. the MCPR below th operating MCPR.
To assure that the fuel cladding integrity saf ety limit (MCPR of 1.07) 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 saf ety it:it MC?R gives the mini =u= operating limit MC?R to avoid violation of the safety limit should the, cost limiting transient occur.
The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
The evaluation of a given transient begins w_wh the system initial parameters shewn in Table 6-2 of Reference 9 that are input to a GE core dynamic behavior transient cc=puter program described in Ref erence 8.
Also, the void reactivity eas *ficients that were input to the transient calculational procedure are based on a new method of calculation ter=ed NEV which provides a better agreement between the calculated and plant instrument power distributions.
The outputs of this program along with the initial MCPR for= the input for further analyser of the ther= ally it=iting burdle with the single channel transient thermal hydraulic SCAT code described in Ref erence 1.
The principal result of this evaluation is the reduction in MC?R caused by the transient.
From 30C4 to EOC4, the most limiting transient for the 8 x 8R fuel is the loss of 1000F feedwater heating with a dCPR of 0.14.
The most limiting event through-out cycle 4 for 8 x 8 and 7 x 7 fuel is the Rod Withdrawal Error (KRE) with a ACPR of 0.17 for 8 x 8 and 0.19 for 7 x 7.
Therefore, the MCPR's specified in 3.ll.C are based on loss of 100 F feedwater heating and the Rod Withdrawal Error.
Amendment No. 33, 38, A2, A3, 32, 69 3.11-4 s
3.u em ux utdios etatunts A. Site Edwin I. Hatch Nuclear Plant Unit No. 1 is located on a site of about 2244 acres, which is ovncd by Georgia Power Company, on the south side of the Altamaha River in Appling County near Baxley, Georgia.
The Universal Transverse Mercator Coordinates of the center of the reactor building are:
Zone 17R LF 372,935.2m E and 3,533,765. 2m N.
B. Reactor Core
- 1. Fuel Assemblies The core shall consist of not more than 560 fuel assemblies of the licensed combination of 7 x 7 bundles which contain 49 fuel rods and 8 x 8 and 8 x SR fuel bundles which contain 62 or 63 fuel rods each.
l
- 2. Control Rods The reactor shall contain 137 crucifer =-shaped control rods.
The control material shall be boron carbide powder (B4C) compacted to apDroximately 70% of its theoretical density.
C. Reactor Vessel The reactor vessel is described in Table 4.2-2 of the FSAR.
The applicable design specifications shall be as listed in Table 4.2-1 of the FSAR.
D. Containment
- 1. Primary Containment r
The principal design parameters and characteristics of the primary con-tainment shall be as given in Table 5. 2-1 of the FSAR.
- 2. Secondary Containment The secondary containment shall be as described in Section 5.3.3.1 of the FSAR and the applicable codes shall be as given in Section 12.4.4 of the FSAR.
- 3. Primary Containment Penetrations Penetrations to the pri=ary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.
E. Fuel Storace
- 1. Soent Fuel All arrangements of fuel in the spent fuel storage racks shall be main-tained in a suberitical configuration having a keff not greater than 0.90 for normal conditions and a keff not greater than 0.95 for abnormal conditiens.
1012 236
-2. sw Fu The new fuel storage vault shall be such that the keff dry shall not be Srcater than 0.90 and the keff floeded shall not be greater than 0.95.
Amendment No. 69 5.0-1