ML20096H524
| ML20096H524 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/20/1992 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20096H528 | List: |
| References | |
| NUDOCS 9205270272 | |
| Download: ML20096H524 (31) | |
Text
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ffb UNITED STATES E
E NUCLEAR REGULATORY COMMISSION J '/
WASHINGTON, D.C. 20066 o
HEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA
.GITY OF DALTON. GEORGIA DOCKET NO. 50-321 IDWIN 1. HATCH NUCLEAR PLANT. UNIT 1 at!ENDMENT TO FACILITY OPERATING LICENSE Amendment No.180 License No. DPR-57 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amcndment to the Edwin I. Hatch Nuclear Plant, Unit 1 (the facility) Facility Operating License No. DPR-57 filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated October 14, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations as 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 set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common 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.
9205270;'72 920520 "DR ANCK 0500c; p
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, 1 2.
Accordingly, the license is hereby amended by page changes to the i
Technical Specifications as indicated. in the attachment to this license I
amendment, and paragraph 2.C.(2) of Facility Operating License No. OPR-57 is hereby amended to read as follows:
Technical Specifications 1
The Technical Specifications contained in Appendix A and B, as revised through Amendment No.180, are hereby incorporated in the i
license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
l FOR THE NUCLEAR REGULATORY COMMISSION
/'
f
'1.Q David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
May 20,1992 l
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UNITED GTATES E
i NUCLEAR REGULATORY COMMISSION
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WASHINGTUN, D.C. 20066
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GEORGIA P_0VER COMPANY QS_l1 THOR E 10WER CORPORAT10N tg K lPAL ELECTRIC AUTHORITY OF GEORGIA RTY OF DALTON. GEORGIA DOCKET NO. 50-351 EDWlf d. HATCH NUCLEAR PLANT. UNIT 2 3H BDMEN1 TO FACitITY OPERATING LICENSE
{
Amendment No.121 Licem No. NPF-5 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The apnlication for amendment to the Edwin I. Hatch Nuclear Plant, Unit 2 (the facility) Facility Operating License No. NPF-5 filed by q
the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated October 14, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 wichout endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance uf this amendment will not be inimical to the common 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment,andparagraph2.C.(2) lows:of Facility Operating License No. NPF-5 is hereby amended to read as foi l
Technical Specifications j
The Technical Specifications contained in Appendix A and B, as revised through Amendment No.121, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with
-the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall-be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY C0hn!SS10N 4$
t David B. Matthews, Director-Project Directorate II-3 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
May 20, 1992 li m
ATTACHMENT TO LICENSE AMENDHENT NO.180 FACIllTY OPERATING LICENSE NO. OPR-57 DOCKET NO. 50-321 88Q TO LICENSE AMENDMENT NO.121 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 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 areas of change.
Bemove Paaes insert Panes Unit 1-111 iii 1.1-12 1.1-12 3.3-la 3.3-la 3.3-5 3.3-5 3.3-6 3.3-6 3.3-7 3.3-7 3.3-15 3.3-15 3.3-16 3.3-16 3.3-17 3."-17 3.3-18 3.3-18 3.3-18a 3.3-19 3.3-19 Unit 2 IV IV IX IX B 2-9 8 2-9 3/4 1-9 3/4 1-9 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12 3/4 1-14 3/4 1-14 3/4 1-15 3/4 1-15 3/4 1-16 3/4 1-16 3/4 10-2 3/4 10-2 B 3/4 1-3 B 3/4 1-3 8 3/4 1-4 8 3/4 1-4 B 3/4 1-4a B 3/4 1-4a B 3/4 1-4b B 3/4 1-4b B 3/4 10-1 B 3/4 10-1 l
t l
Settin SettjM f.ut tlMITltG CONDITIONS FOR 00EPail0N SURVEltLANCE PEOUIREMENTS 3.3.
REACTIVITY CONTROL (CONT')
4.3.
REACTIVITY CONTROL (CONT')
G.
Rod Wortn Minimizer (RWM) l G.
Rod Worth Minimizer (RW*i) 3.3-5l H.
Shutdown Requirements 3.3-7 3.4.
STANDEY LIQUID CONTROL SYSTEM 4.4 STANDBY L10VID CONTROL SYSTEM 3.4-1 A.
Normal System Availability A.
Normal Operational Tests 3.4-1 B.
Coeratirg with Incoerat e E.
Surveillance with 3.4-2 i
Cor;cnents Incperable Comperents C.
Sodium Pentaborate Solution C.
Sodium Pentaborate 3.4-2 Solution D.
Shutdown Repairements 3.4-3 3.5.
CORE AND CONTAlfNENT COOLING 4.5.
CORE AND CONTAINMENT COOLING 3.5-1 SYSTEMS SYSTEMS A.
Core Spray (CS) System A.
Core Spray (CS) System 3.5-1 B.
B.
Residual Heat Removal 3.5-2 System (LPCI and Containment (RHR) System (LPCI and Cooling Mode)
Containment Cooling Mode)
C.
RHR Service Water System C.
RHR Service Water System 3.5-5 D.
High Pressure Coolant Injection D.
High Pressure Coolant in-3.5-6 (HPCl) System
.jection (HPCI) System E,
Reactor Core Isolation i.coling E.
Reactor Core Isolation 3.5-7 (RCIC) System Cooling (RCIC) System F.
Automatic Depressurization F.
Automatic Depressurization 3.5-9 System (ADS)
System (ADS)
G.
Minimum Core and Contain ent G.
Surveillance of Core 3.bl0 Cooling Systems Availability and Contatnment Cooling Systems H.
Maintenance of Filled Discharge H.
Maintenance of filled 3.5-10 Pires Discharge Pipes 1.
Minimut Fiver Flow Minimum River Flow 3.5-11 J.
Plant Service Water System J.
Plant Service L ter System 3.5-12 K.
Engineered Safety Features K.
Engineered Safety 3.5-13 Compart'ent Cooling features Compartment Cooling Amendment No. 180 HATCH - UNIT 1 tii
m e
EASES FOR t!MITIN0 SAFfiY SYS1fM SETTINGS 2.1.A.).a' 10'4 F1 u r Scram Trie Setting (Continued) tism was taken-in this analysis by assuming-that the IRM channel closest to the withdrawn rod is bypassed. S e results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power..thus maintaining MCPR above the fuel cladding integrity Saf a limit.
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, 1
b.
ADPH flui 9am Trie Settina (Refuel or Start & Hot Standby Model 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 iimit, 25 percent of rated. The margin is adequate to.a commodate anticipated maneuvers asso-ciated with power plant startup. Ef fects of' increasing pressure at zero or icw voisi content are einer, cold water from sources available during startup is nct much colder than that alreacy 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 minimi2er.
l Worth of-individual rods is very low in a uniform rod pattern. Thus, of all possible sources of-reactivity input, uniform control red withdrawal is the most probable cause of significant power rise, Because the flux distribution assottated with uniform red withdrawals does not involve high local peaks, and M cause 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 a%umed uniform rod withdrawal approach to the scram level, the rate of m wer rise is no more than 5 percent of rated power per minute, and.ce 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 untG the mode switch is placed in the RUN position. This switch occurs wnen reactor pressure is greater than 825 psig.
c.
APRM Fluy Scram Trie Settiros (Dun Mode)
The APRM Flux scram trips in the run mode consist of the flow referenced simulated thermal power monitor scram setcoint and a fixed high-high neutron flux scram settoint. In the s1rulated thermal power monitor, the APRM flow referenced neutron flux signal is passed through a filter-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 transient or steady-state conditions, This prevents spurious scrats, which have an adverse effect on reactor safety because of the resulting thermal stresses. Examples of events which can result in momentary neutron flux spikes are momentary-flow changes in the recirculation system ficw, and small pressure disturbarces curing turbine stop valve and turbtee control valve testing. These flux spikes represent no hazard to the fuel since they are only of a few seconds duration and I
less than 120% of rated thermal power. The flow independent portion l
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.
d HATCH - UNil 1 1.1-12 N
. ~. -
LIM 111N",CO?OITINJ_QROPERal104 SURVElttAM E FE0J1FEMEN15 3.3.
FEa*TIVITY CONTPOL B.
Increrable Centrcl Reds (Cent'd) 1.
No Movement by Centrol Rod Drive tip Pressure (Cont'd)
'~
If a partially or fully withdrawn control rod drive cannot be moved with drive or scram pressure, the reacter shall be brought to the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and shall not be started unless (1) investigation tas cemenstrated trat the cause cf the f ailure is not a f ailed control rod drive mechanism collet housing, and (2) adequate shutdown margin has been denenstrated as required by Specification 4.3.A.
l If investigation demonstrates that the cause of control roc drive failure is a cracked collet housing or if that possibility cannot be eliminated, the reactor shall not be started until.he affected control rod drive has been replaced or repaired.
HATCH - UNIT 1 3.3-la
/Jnendment No.180 l
l
LIMITl% C0'01110N5 FOR OPEPET10N SU;VilttANCE PE0VIREMEN15 3.3.F.
Oreratica with a tiattica Coatrol 4.3.F.
Operation with a limitina Centrol Fod Pattern (for Rod Withdrawal Rod Pattern (for Red Withdrawal Errer. RWE)
Errer. FWE)
A Limiting Rod Pattern for RWE During operation when a Limiting exists when the MCPR is less Control Rod Pattern for RWE exists than the value provided in the and only one RBM channel is Core Operating Limits Report.
operable, an instrument functional test of the RBM shall be performed During operation with a limiting prior to withdrawal of the control Control Rod Pattern for RWE and rod (s). A Limiting Rod Pattern for when core thermal power is 2 30%,
RWE is defined by specification either:
3.3.F.
1.
Both red blo:k c'oniter PEu,)
j channels shall be C;ERAELE, or 2.
If only one REM channel is OPERABLE, control rod with-dra.al shall be blocked within 24 r.ours, or 3.
If neither RSM channel is OPER-ASLE, control rod withdraaal shall be blocked.
l l
G.
Pod Worth Mini-izer (RWH)
G.
Ped Worth Minimiter (PLH) 1.
OreraH14tv l
1.
Oterability l
Wherever the reactor is in the a.
The RkN shall be demon-Start & Hot Standby
- or Run Mode strated OPERABLE in the below 10% rated thermal power, Start and Hot Standby the RhM shall be OPERABLE.
Mode prior to withdrawal of control rods for the a.
With the Rh9 inoperable purpose of making the before the first 12 control reactor critical and in rods are withdrawn on a the Run Mode when the Ek9 startup, one startup per is initiated during control calendar year eay be per-rod insertion wnen redating formed provided that IhERMAL POWER by:
control rod r.ovement and compliance with the pre-(1) Verifying proper scribed BPWS control rod annunciation of the pattern are verified by a selection error of at second licensed operator least one control rod or Qualified memoer of the which violates the pre-plant technical staff, scribed withdrawal sequence loaded into b.
Wit
>e Pk9 inocerable the RhH, and after :he first 12 control rods have been fully with-(2) Verifying the rod block drava on a startup, opera-function of the RhM by tion may continue provided attempting to move a that control rod movement centrol rod that and compliance with the violates the prescribed withdrawal sequence loaded into the RWE
- Er.try tr*.o the Start ard H t Stard y M Me and withdraaal cf selected centrol rods is perritted for the purpose of determining the OPERAE!LITY of the PWP prior to withdrawal of control rods fcr tre purpose of bringing the rea:ter to criticality, HAICH - UNIT 1 3.3-5 Amendment No. 180
L 4
tlHilN; COWITIONS FOR UDECATION SURVEILLAN:E #!0VICEMEN15 3.3.G.l.b.
prescribed BFWS control 4.3.G,1.b.
The RWH shall be demonstrated rod pattern are verified OPERABLE after a sequence of by a second licensed rod moves has been loaded into operater or Qualified the hH by verifying that member of the plant sequence confo m to BPWS.
technical staff.
c.
With RhH inoperable on a shutdown, shutdown may continue,provided control rod movement and com-pliance with the prescribed BPWS control rod pattern are verified by a second licensed operator or 0;alified re-ter of tne plant technical staff.
l
'HAICH - UNII 1 3.3-6
_ Athendrnent No.180 4
4
-(
LIMllltd C0'01110N5 FOR OPEWATION SURVElttANCE R!0VIRf MEN 15 3.3.G.2.
Srectal Test Encertions-4,3.G.2.
Soecial Test Excertions The BPWS rod pattern If the RWM or individual requirements of $pecification' rods in the RWM are bypassed,
-3.3.G.) may be suspenced while a second licenseJ operator or in Startup and Hot Standby-and qualified member of the plant-
' Run Modes with thermal power technical staff shall verify less than 10% of rated to that movement of control rods allow performance of SHUT-is in compilance with the DOWN MARGIN demonstrations, approved centrol rod moves control rod scram time for the specified test, testing, control rod friction testing, or startup testing, provided the RWM is bypassed or individual rods in the FJM are bypassec and con-formance to the approved control rod movement for the specified test is t
verified by a second-licensed operator or qualified member of the plant technical staff.
I-l.'
l H.
-Shutdown Pecuire=ects if'$pecifications 3.3, A through 3.3.G are not met, an orderly shutdown shall be initiated and the reacter placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
' HATCH - UNii 1 3.3-7
/VDeDdm00t IK). 180 l-
-. - -... -.. - ~
e FASES 500 tlMITifG COND111045 F0P Orf rifl0N t.ND SURyf tttt.N:E s!OWPEMENM 3,3,F.' Oceratien with a timitiro Control Rod Pattern (fer Rod' Withdrawal Errer. PWE)
Surveillance Requirements:
A limiting control rod pattern for RWE is a rattern which, due to unrestricted withdrawal of any. single control rod, could result in violation of the MCFR
- Safety Limit, Specif teation 3.3.F. defines a limiting control rod pattern for RWE, During use of such patterns when both RBM channels are not operable, i
it is judged that testing of the RBM system prior to withdrawal of control rods to assure its operability will assure that improper withdrawal does not occur, Reference NEDC-30474-P (Ref, 17) for more inforrration, G,
Pod Werth MiniMrer (PM
- l. DreraHlity Limiting Conditions for Operation:
.The RW) ret t.*icts withdrawals and.1nsertions of control rods to prespecifiea sequences that comply v:ith BPW5. All patterns associated with these seguences have the characteristics that, assuming the worst single deviation from the pattern, the drop of any Control rod from the fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in any pellet average enthalpy in excess of 280 calories per gram. An enthalpy of 280 calories per gram is well telow the level at which rapid fuel dispersal could occur (i.e., 425 calories per gram). Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Reference Section 3.6.5.4, 3.6.6, 7.14.5.3, and 14,4,2, and Appendix P of the FSAR, and NEp0-2123),
l The rfRC requires the RWM to be highly reliable to minimize the need to depend on a second licensed operator or cualified member of the plant technical staff to verify compliance with EPW5 below 10% RTP. To accomplish this, RWM must be OPERABLE during the first 12 red withdrawals during startup. The NRC is willing to allow one startup per calendar year without the RWM to avoid delays that r:ay occasionally occur. Below 10% RTP with the RWM ir. operable, all control red rnovements and ccmoliance with the prescribed control rod patterns must be verified by a seconc licensed cperator or qualified member of the plant technical' staff.
Above 10% RTP the RWM is not required to t>e OPERABLE nor is it required to be loaded with a secuence of rod moves that conforms to BPW5.
i' v.
I L
-HATCH - t%!T 1 3.3-15 Amendment no. 180 l
I
, _. ~ - -. -
a FaSES F09 tlMITING C00lT10NS F0D OE!G110N A'O SL9VEILL ANCE F10UlFEMEN15 3,3.0,1 Operability ~ (Continued) ln performing the function described above, the RhH is not required l
to impose any restrictions at core power levels in excess of 10%
of rated. Material in the cited references shows that it is impossible
' to reach 280 calories per gram in the event of a control rod drop occur-ring at power greater than lot, regardless of the rod pattern. This is true for all. normal and abnormal patterns including th'ose which maximize the individual control rod worth.
At. cower levels belew 10*. of rated, abnormal control rod patterns could produce rod worths high enough to be of concern relative to the 280 cal-i orie per gram rod drop limit. In this range, the RWM constrains the control l
rod sequences and patterns to those which involve cely acceotable rod
' ;rths.
ine RWM providEs automatic : gerviston to assure that out of sequence control rods will not be withdrawn or inserted; i.e.. It limits operator deviations-from planned withdrawal sequences. It serves as a backup to l
procedural control of control rod sequences, which limit the maximum reactivity worth of control rods, in the event that the RWM is out of service, when required, a second licensed operator or qualified member of the plant technical staff can manually fulfill the control rod 4
pattern conformance functions of this system.
The function of the Rh9 makes it uarecessary to specify a license limit l
on rod worth to preclude unaaceptable consequences in the event of a control rod drop. At low powers,' below 10%,.this device forces l
adherence to acceptable rod patterns. Above 10% of rated power, the consequences of a rod drop event without Rhh are acceptable. Power level for automatic cutout of the RhH function is sensed by feedwater and steam flow.
Surveillance Requirements:
functional testing of the Rei prior to the start of control rod withdrawal at startup and prior to attaining 10% of rated thermal power during rod insertion while shutting cown will ensure reliable cperation.
l 2.
Seecial Test Excertions
-In order to perform the tests required in the Techr.ical Specifications. it is necessary to bypass the BPWS restraints on control rod movement. The additional surveillance requirements ensure the specifications on beat l
generation rates and shutdown margin requirements are not exceeded during the
'~
period when these tests are being performed, and individual rod worths do not exceed the values assumed in the safety analysis.
F _ - Shutdown Pecuire*ent s.
Should circ.nstances be such that the Limiting Conditions for Operation as stated in Specifications 3.3.A. through 3.3.C. cannot be rnet, an orderly shutdown shall be_ initiated and the reactor placed in
- the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l l
HATCH - UNIT 1 3.3-15 Amendment No. 180 l.
L
L q
1 Fa5ES F00 LIMITIN3 CON 31110' 5 F00 OPEcal104 AND SUPVElttANCE CEDUlcEMENTS
- 1.. Scram Discharae Volute vent and Drain Valves
'The scram discharge volume-vent and drain valves are. required to be
-OPERABLE so _that the scram discharge volume will be availeble when
-i needed to accept discharge water from the control rods daring a reactor scram and will isolate the reactor coolant system from the containment when required.
J.
Feferences 1.
FSAR Section 3.4. Reactivity Control Mechanical Design.
2.
FSAR Section 3.5.2, Safety Design Bases.
2.
EIAA Stettet 2.i.". -af t t, [+.al Lat ic.n.
4 FSAR 5ection 3.5 Control Red Drive Housing Supports.
5.
FSAR Section 14.4.3 Loss-of-Coolant Accident.
6.
FSAR Section 14.4.2. Control Rod Drop Accident.
a 7.
C. J Paone, " Banked Position Withdrawal Sequence,"
NED0-21231. January 1977 B.
FSAR Section 3.6.5.4, Control Rod Worth.
- 9. FSAR Section 3.6.E Naclear Evaluations.
3 t
HATCH - UNIT 1 3.3 17 Amendment No. 180
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e s'
I L
NTCH.- Ur?!T I 3 3-18 Amendment No. 180
Bul$ FOR LIMlflM CON 01110NS FOR OPICAllON AND SURVilLLANCE RIOUIREMENTS 3.3.J.
References (Continued)
- 10. FSAR Section 7.14.5.3, Rod Worth Minimizer function l
i
- 11. FSAR Section 3.6.4.1. Control Rods
{
- 12. FSAR Question 3.6.7, Amendment 24
- 13. ' Average Power Range Monitor, Red Block Monitor and Technical Specification improvement (ARTS) Program for Edwin 1. Hatch Nuclear Plant, Units 1 and 2,*
NEDC-304'i4-P, Decemt.er 1983.
HATCH - UNIT 1 3.3-19 Amendment No. 180
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION.
PAGE
-3/4.0--APPLICABILITY...............................................
3/4 0-1 3/4.1-REACTIVITY CONTROL SYSTEMS 3/4.1.1 IH'JTDOWN M 00lN....
3/t 11 3/4.1..
RIA; IVITY Ah;MA.!!L...........................
3/4 1-2
-3/4.1.3 CONTROL RODS Control Rod Operability..................................
3/4 1-3 Control Rod Maximum Scram Insertion Times................
3/4 1-5 Control Rod Average Scram Insertion Times................
3/4 1-6 four Control Rod Group Scram Insertion Times.............
3/4 1-7 Control Rod-Scram Accumulators...........................
3/4'l-8 Control Rod Drive-Coupling...............................
3/4 1-9 Control Rod Position Indication....................
3/4 1-11 Control' Rod Drive Housing Support.................
3/4 1-13 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer................
3/4 1-14 I
Rod. Block Monitor........
.................... =3/4 1-17 3/ k._1,5 STANDBY LIQUID = CONTROL SYSTEM............................
3/4 1-18 l
t 3/4.2 POWER DISTRIBUTION LIMITS 4
3/4.2.1-AVERAGE PLANAR LINEAR HEAT GENERATION RATE...............
3/4 2-1 3/4 2.2 APRM SETP0lNTS...........................................
3/4 2-5 i.
3/4.2.3 MINIMUM CRITICAL POWER RAT10.............................
3/4 2-6
- 3/4.2.4 LINEAR-HEAT GENERATION RATE..............................
3/4 2-8 HATCH-UNIT 2 IV.
Amendment No.
121 1
e
. INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION-PAK 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH 3/4 9-1 3!: i.I "h3TR'.9?Nitii N 4
3/4 5-3 3/4.3.3 CONTROL RCE POSITION 3/4 9 5 3/4.9.4 DECAY-TIME-3/4 9-6 3/4.9.5 SECONDARY CONTAINMENT Refueling-Floor 3/4 9-7 Secondary Containment Automatic Isolation Dampers 3/4 9-B Standby Gas Treatment System 3/4 9-10 3/4.9.6 COMMUNICATIONS 3/4 9-11 3/4.9. 7 CRANE AND HOIST OPERABILITY 3/4 9-12 3/4.9.8 CRANE TRAVEL - SPENT FUEL STORAGE POOL 3/4 9-13 3/4.9.9 WATER LEVEL - REACTOR VESSEL 3/4 9-14 3/4.9.10 WATER LEVEL - SPENT FUEL STORAGE POOL 3/4 9-15 3/4.9.11 CONTROL ROD REMOVAL Single Control-Rod Removal 3/4 9-16 Hultiple Control Rod Removal 3/4-9 18 3/4.9.12 REACTOR COOLANT CIRCULATION 3/4 9-20
-3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY 3/4 10-1 3/4.10.2 ROD WORTH MINIMIZER 3/4 10-2 3/4.13.3 SHUTDOWN MARGIN DEMONSTRATIONS 3/4 10-3 3/4.10.4 RECIRCULATION LOOPS 3/4 10-4 HATCH-UNIT 2 IX Amendment No. 121
E 2.2 LIMITING SAFETY SYSTEM SETTINGS-BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Setpoints specified in Tacle 2.2.1-1-are the values at which the reactor trips are set for each e m -eter.
The Trin Setpoints have been selected to enture that the rereter
- +
rc t: r c d ir* O ster et cre,entt. #r;- exced ug ti 1 U tt,
L-it.
m 1;r mtr t :": si lu: ce u vat we
- 1rw On *. t :
within its specified Allowable Value, is acceptaoie on the basis tnat eacn Allowabli Value is equal to or less than the drift allowance assumed for each trip in_the safety analyses.
1.
Intermediate Rance Monitor. Neutron Flux The IRM 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, tM trip setpoint is also ranged up.
The IRM instruments provide for overlap with both the APRM and SRM systems.
The most significant source of reactivity chenges during the power increase 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 '/.5 of the FSAR. The most severe case involves an initial condition in which the reactor is just subtritical, 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%
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.
Averaae Power Ranae 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 petterns are constrained by the RWM, l
1 HATCH - UNIT 2 B 2-9 Amendment No. 121
REACTIVITY CONTROL SYSTEMS CONTROL ROD-ORIVE COUPLING tIMITING CONDITION FOR OPERATION 3.1,3.6' All control rods shall be coupled to their drive mechanisms.
!JH 11 U.J C V:
COND!CNE 1. 2 and ",
- T1%.
a.
In CONDITION 1 or 2, with one control rod not coupled to its
.i associated drive mechanism, the provisions of Specification 3.0.4 are not applicable, and operation may continue provided; 4
1.
If permitted by the RWM, the control rod drive mechanism l
is inserted to accomplisk oupling, and recoupling is verified by demonstrati t the control rod will not go to the overtravel posi
, or 2.
If recoupling is not secomplished on the first attempt or if not permitted by the RWM, the control rod is declared l
inoperable-and fully inserted, and -the requirements of Specification 3.1.3.1 are satisfied, b.
In CONDITION 5*, with a withorawn control rod not coupled to its associated drive mechanism, within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
1.
Insert the control rod to accomplish recoupling, and verify recoupling by demonstrating that the control rod will not go to the overtravel position, or 2.
If recoupling is not accomplished, fully insert the control rod and either electrically disarm the control rod or close the withdraw isolation valve.
3.
The provisions of Specification 3.0.3 are not applicable.
l i
- At least each withdrawn control rod.
Not applicable to control rods removed per Specification 3.9.11.1 or 3.9 11.2.
^
HATCH - UNIT 2 3/4 1-9 Amendment No. 121
..D
- REACTIVITY CONTROL SYSTEMS CONTROL ROD p0SITION INDICATION LIMITING CONDITION FOR OPERATION 3.1.3.7 All control rod reed switch position indicators shall be OPERABLE.
!!?TR'd 'Iv:
C0r!'!?NS 1. 2 ar.d In CONDITION 1 or 2, with one or more control rod reed switch a.
position indicaters inoperable, the provisions of Specification l
3.0.4 are not applicable, and operation may continue provided that within 1 hour:
1.
The position of the control rod is determined by an alternate method, or 2.
The control rod is moved to a position with an OPERABLE reed switch position indicator, or 3.
The control rod is declared inoperable and the requirements.of Specification 3.1.3.1 are satisfied; Otherwise,~be in at least HOT SHUT 00WN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In C0t'91 TION 5*, with a withdrawn control rod reed switch posi-tion indicator inoperable, move the control rod to a position with an OPERABLE reed switch position indicato-or fully insert the control rod.
The provisions of Specification 3.0.3 are not applicable.
'i
- At-least each withdrawn control rod.
Not applicable to control rods t
removed per Specification 3.9.11.1 or 3.9.11.2.
HATCH - UNIT 2 3/4 1-11 Amendment-No._121
~.
~
.. -. ~.. - ~.
-... ~
REACTIVITY CONTROL SYSTEMS
- SURVElltANCE RE0VIREMENTS 4.1.3.L1 The control. rod reed switch position indicators shall be determined OPERABLE by_ verifying:
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, that the position of the. control a.
red is indicated, t.
Th:1 i'.
ci:ned ccrtr:1 rcd ::siti:-
,2 rg.n c, i r o t he
_ PovctM u ;r.e ter. trol rod nner. performins 5urvelilai;ce Requirement-4.1.3.1, and That-the control rod reed switch position indicator corresponds c.
to the control rod position indicated by the " full-out" reed switches when performing Surveillance Requirement 4.1.3.6.b.
t P
HATCH - UNIT 2 3/4 1-12 Anendment No.121
1, -.
9EACTIVITY CONTROL SYSTEMS-3/4.1.4' CONTROL R0D PROGRAM CONTROLS ROD WORTH MINIM 12ER-LIMITING CONDITION FOR OPERATION 3.1.4.1 The Rod Worth Minimizer (RWM) shall be ODEPABLE.
h;I' m 'i';
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i.miP. is iet s it,an :CL of RAliJ THi % L-F0niR.
ACTION:
-a.
With the RWM_ inoperable before the first 12 control rods are withdrawn on a startup, one startup per calendar year may be performed provided control
-rod movement-and compliance with the prescribed BPWS control rod pattern are verified _by a second-licensed operator or qualified member of the plant technical staff, b.
With the RWM inoperable after the first 12 control rods have been fully withdrawn.on a startup, operation may continue provided that control rod movement and compliance with the prescr" sed BPWS control rod pattern are
! verified-by a second licensed operator or qualified member of the plant technical staff,_
With RWM inoperable on a shutdown, shutdown may
'nue provided control c.
rod movement'and compliance with the prescribed t ontrol rod pattern are-verified by a second licensed operator or qual;-
1 member of the plant: technical staff.
L L
1:
2 l
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior _to withdrawal of control rods for the purpose of bringing the
. reactor to criticality.
i HATCH - UNIT 2 3/4 1 14 Amendment No. 121
-6 REACTIVITY CONTROL SYSTEMS 3 /4II 4 CONTROL ROD PROGRAM (QNTROLE ROD WORTH MIN 1MIZER
\\
t jVRVEILLANCE REOUIREMENTS 1
4.1.4.1 -The RWM shall be demonstrated ODERABLE:
a.
!r CO*C:~:% 2 r er to e t'drN21 0' ;0 r, t r : 1 r:d: ':- int n rcrse of making the reactor critical, and in CONDITION 1 wnen the RWM is initiated during control rod insertion when reducing THERMAL POWER, by,.
1.
Verifying proper annunciation of the telection error of at least one out-of-sequence control rod, and 2.
Verifying the rod block function of the RWM by moving an out-of-sequence control rod.
~b.
By verifying the sequence of rad moves loaded into the RWM conforms to BPWS following the loading of that sequence.
F t
4 HATCH - UNIT 2 3/4 1-15 Amendment No.
121
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i 1
I 1
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ggtg. Utill 2 g4g fpendm00t No. W
[
JPECIAL TEST EXCEPTIONS 3/4.10.2 ROD WORTH Mit41til7ER l
LIMIT 1tiG CONDITION FOR OPERATION 3,10 2 The BPWS rod pattern requirements of Specification 3.1.4.1 may be suspended while in Conditions 1 and 2 with THERMAL POWER LESS THAN 10% of RATED to allow performance of SHUTDOWN MARGIN demonstrations, control rod scram time testing, control rod friction testing, or startup testing, provided
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i SUDVElltANCE RE0VIPEt4ENTS h
4,10.2 If the RWM or individual rods in the RWM are bypassed, verify preposed movement of control rods is in compliance with the approv. ' control rod rnoves for the specified test.
i HATCH - UNIT 2 3/4 10-2 Amendment No. 121 L
PEACTIVITYsCONTROL SYSTEMS BASES CONTRO1 RODS fContjnued)
I than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure.
Operability of the accumulator ensures that there is a means available to insert the cetrol l
rc a: <.v. e m
-. ; u. i n e r n. r.
..p t :;ruatu
- = f Da ru m t m rol red ;cupitt,; intyrity is re;;nre. to e:..ure complianct with the analysis of the rod drop tecident in the FSAR.
The overtravel position feature provides the only positive means of determining that a rod is properly coupled, and therefore, this check must be performed prior to achieving criticality-after each refueling.
The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod
-position indication system must be OPERABLE.
The conti'o1 rod housing support restricts the octward movement of a control rod to less than 3 inches in the event of a housing failure.
The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not
-contribute to any damage to the primary coolant system.
The support is not required when there is no pressure to act as a driv N torce to rapidly f
ej::ct a drive housing, The required surveillance intervals are adequate to determine'that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
1/4.1.4 CONTROL p00 PP0GRMi CONTROLS Control rod _ withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to cause the peak _ fuel enthalpy for any postulated control rod accident to exceed 280 cal /gm. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal.
When THERMAL POWER is 2: 10% of RA1ED THERMAL POWER, there is no possible rod l
worth which, if dropped at the design rate of the velocity limiter, L
could _ result-in a peak enthalpy of 280 cal /gm.
Thus, requiring the RWM to be OPEP.ABLE below 10% of RATED THERMAL POWER provides adequate control.
HATCH - UNIT 2 B 3/4 1-3 Amendment No. 121 I-
4 a
REACTIVITY CONT _ROL SYSTEMS BASES CONTROL RQDS PROGRAM CQUTROLS (Continued)
The RWM previo.5 automatic supervision to assure that out-of-sequence l
rods will not be wi*,hdrawn or inserted.
The analysis of the rod drop accident is presented in Section 15.1.38 of the fSAR. ead the technicues :# the analyr4s are e-etent-d in e tonical i*<a.,
i rs::rt.
4 The NRC requires the RWM De highly reliauie to minimize the neeo to depend on a second licensed operator or qualified member of the plant technical staff to verify compliance with BPWS below 10% R1P.
To accompitsh this, RWM must be operable during the first 12 rod withdrawals during startup.
The NRC is willing to allow one startup per calendar year without RWii in order to avoid delays that may occasionally occur.
Below 10% RTP with the RWM inoperable, all control red movements and compliance with the prescribed control rod patterns must be verified by a second licensed operator or qualified member of the plant technical staff.
Above 10% of RTP, the RWM is not required to be OPERABLE nor is it required to be loaded with a saquence of rod moves that conforms to BFWS.
The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation.
The-RBM is only required to be OPERABLE when the Limiting Condition defined in Specification 3.1.4.3 exists.
Two channels are provided.. Tripping one of the channels will block erroneous rod with-drawal soon enough to prevent fuel damage.
This system backs up the written sequence usea by the operator for withdrawal of control rods, further dis-cussion of the RBM system-and power dependent setpoints may be'found in NEDC-30474-P (Ref. 4).
+
HATCH - UNIT 2 B 3/4 1-4 Amendment No. 121
a t
PEACTIVITY EONTROL SYSTEMS BASES 3 / 4.1.uS STANDBY L10VID CONTROL SYSTEM The standby liquid control (SLC) system provides a backup reactivity control capability to the control rod scram system.
The original design basis for the standby liquid control system is to provide a soluble boron cr rentrat'm to tr a ratt er ve"el ru"itivt to trire the reactne to a cid v t: w..
1:
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41-
% r. w t. t n
. : t. -
a'.'c ntisfy tr4 r r o<r':' t i L
F.'. U c la "TR R.C2 w rarW ('
U),
which requires that the system nave a ca. trol capacity equivalent to that for a system with an injection rate of B6 gpm of 13 weight percent unenriched sodium pentaborate, normalized to a 251 inch diameter reactor vessel.
To meet its original design basis, the SLC system was designed with a sodium pentaborate solution tank, redundant pumps, and redundant explosive injection valves.
The tank contains a sodium pentaborate solution of sufficient volume, concentration and B,0 enrichment to bring the reactor to a cold shutdown. The solution is injected into the reactor vessel using one of the redundant _ pumps.
The volume limits in figure 3.1.5-1 are calculated such that for a given concentration of sodium pentaborate, the tank contains a volume of solution adequate to bring the reactor to a cold shutdown, with margin.
These volume limits are based on gross volume and account for the unusable volume of soletion in the tank and suction lines.
To meet 10 CFR 50.62 Paragraph (c) (4), the system must have a reactivity cor; trol capacity equivalent to that of a system with an 86 gpm injection flow rate of 13 weight percent unenriched sodium pentsborate into a 251 inch diameter reactor vessel.
The term " equivalent geactivity control capacity" refers to the rate at which the boron isotope B is injected into the reactor core. - M e standby liquid control system meets this requirement byusggasodiumpentaboratesolutionenrichedwithahigherconcentrationof the B isotope. The minimum c ecentration limit of pentaborate solution is based on 60 atomic percent B,fg.2 percent sodium enriched boron in sodium pentaborate and a flow rate of 41,2 gpm. The method used to show equivalence with 10 CFR 50.62 is set forth in NEDE-31096-P (Ref. 5).
Limiting Conditions for Operation are established based on the redundancy within the system and the reliability of the control rod scram system. With the standby liquid control system inoperable, reactor operation for short periods of time is justified because of the reliability of the control rod scram system. With one redundant mmponent inoperable, reactor operation for longer periods of time is justified because the system could still fulfill its function.
HATCH - UNIT 2 B 3/4 1-4a Amendment No. 121
Qi G yDY K *ROL SYSTIM
- y. i <,. ~.n o
~
)
b!? Q LIOUID C0f1 TROL SYSTEM (Continued)
Surveillance requirements are established on a frequency that assures a high system reliability.
Thorough testing of the system each operating cycle assures that the system can be actuated from the control room and will develop the flow rate required.
Replacement of the explosive charges in the valves at regular intervals assures that there valves will nSt 'til due to deteriaration e'
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. :r.u ;. r rar.in t o a. s u r e ;,.7 :: n a t 41 P...
The sodium pentaborate solution is carefully monitored to assure its
~
reactivity control capability is maintained.
The enriched sodium pentaborate solution is made by mixing aranular, enriched sodium pentaborate with water.
actual B', tests on the granular sodium pentaborate are performd to verify the Isotopic enrichment, prior to mixing with water.
Once the enrichment is established, only the solution concentration, volume, and temperature must be monitored to insure that an adequate amount of reactivity control is r;
available.
Determining the solution concentration once per 31 days verifies that the solution has not been diluted with water.
Checking the volume once each day will guard against noticeable fluid losses or dilutions, and daily temperature checks will prevent sodium pentaborate precipitation.
1.
C. J. Paone, " Banked Position Withdrawal Sequence," f4ED0-21231, January 1977.
2.
Deleted.
3.
Deleted.
4.
" Average Power Range Monitor, Rod Block Monitor and Technical Specifi-cation Irmrovement (ARTS) Program for Edwin 1. Hatch fluclear Plant, Units 1 and 2," f4EDC-30474-P, December 1983.
5.
" Anticipated Transients without Scram, Response to t4RC ATWS Rule, 10 CFR 50.62", tiEDE-31096-P, December 1985.
4-HATCH - UtilT 2 B 3/4 1-4b Amendment No. 121
3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 PRIMARV CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is removed during the period when open vessel tests are being performed during low power PHYSICS TESTS.
- $.
- r:D E;E;3 "!N!"::Li i
in order to perform the tests required in the Technical Specifications, it is necessary to bypass the sequence restraints on control rod movement.
The additional surveillance requirements ensure that the Specifications on heat generation rates and shutdown margin requirements are not exceeded during the pericd when these tests are being performed.
3/4.10.3 SHUTD.QWN MAPGIN DEMONSTRATION!
Performance of shutdown margin demonr* rations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur.
These additional restrictions are specified in this LCO.
3/4.10.4 REClRCULATION LOOPS
-This special test exception permits reactor critica ity under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.
HATCH - UNIT 2 8 3/4 10-1 Amendment No. 121
..