ML20010J235
| ML20010J235 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/15/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20010J236 | List: |
| References | |
| NUDOCS 8109300008 | |
| Download: ML20010J235 (79) | |
Text
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UNITED STATES
'k NUCLEAR REGULATORY COMMISSION El ),q g-WASHINGTO N, D. C. 20$55 f
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p TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No.76 License No. DPR-33 1.
The Nuclear Regulatory Commission (_the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated April 29, 1981, as supplemented by letters dated June 12, 1981 and July 13, 1981 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 tne 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 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 amended by changes to the Technical Speciti-cations as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility License No. OPR-33 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, l
as revised through Amendment No. 76, are hereby incorporated l
in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
B10Y30000s st0915-DR ADOCK 05000259 PDR
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'l 3.1 This: license. amendment is effective as of the date of its issuance.
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. FOR THE NUCLEAR-REGULATORY COMMISSION
.-m i
homas
_ Ippolito, Chief
. 0perating-Reactors Branch #2 Division of Licensing i'
Attachment:
-Changes to.the, Technic 11
-Speci fications Cate of Issuance:, September '.5,1981 I
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[7 ATTACHMENT TO LICENSE AMENDMENT NO. -'76
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y FACILITY OPERATING LICENSE NO. OPR ~
DOCKET NO. 50-259 G
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' LRevise Appendix A as follows:
. l., Remove the following pages and replace with identically numbered-pages-
- Jiii 'iv 73/74 J171 172 vii/viii' 79/M 172a/T72b -
U/Ti T71/122 181/182-
' T6
.12 TH -
218/TfT TY/20 130 220/72T 2T/22 132-226/2R
~H/24 144 235a 2T/26 146' F
27/30 158 2T6/251
'31/ E l_59_/'iTOE 252/23T y
Q48 T6T/
261 263
/267' 268 23T/270 330/T3T 2.
The ' underlined pages are.those t,eing changed; marginal lines on these pages indicate the revised area. The overleaf page is provided-for convenience.
3.
Add the, following new pages:
160a
.169a:
251a
-261a
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Page No.
Section 180 C.
Coolant Leakage.................
181
- 0. LSafety and Relief Valves............
s 181 E..
Jet Pumps................
182 F.
Recirculation Pump Operation 182 G.
Structural Integrity..............
H. ~ ShockSuppressors-(Snubbers)...........
185 227 3.7/4.7-Containment Systems................
227 A.
Primary Containment 236 B.
Standby Gas Trea'-- ' System......
240 C.
Primary Containment Isolation Valves......
242 E.
Control Room Emergency Ventilation.......
244 F.
Primary Containment Purge System'...
246
'G.
Containment Atmosphere Dilution System (CAD)..
248 H.
Containment Atmosphere Monitoring (CAM) System H2 and 02 Analyzer..............
249 3.8/4.8 Radioactive Materials 281 A.
Liquid E f fluents................
281
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B.
Ai rbo rne E f fl uents...............
282 C.
Mechanical Vacuum Pump 286 0.
Miscellaneous Radioactive Materials Sources 286 3.9/4.9 Auxiliary Electrical System 292 A.
Auxiliary Electrical Equipment.
292 B.
Operation with inoperable Equipment 295 C.
Operation in Cold Shutdown...........
298 3.10/4.10 Core Alterations.
302 A.
Refueling Interlocks..............
302 iii Amendment No. 76
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Section Page No.
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. Core Monitoring 305 C.-
Spent Fuel Pool Water.............
305 L
- 0. -Reactor Building Crane...-..........
307 e
F.
E.
Spent Fuel Cask 307 F.
Spent Fuel Cask Handling-Refueling Floor....
308 3.11/4.11-Fire Protection Systems 315 A.
High Pressure Fire Protection System.......
135 B..' C02 Fi re Pro tec ti on Sys tem...........
31 9
.C.
Fire Detectors
'320 D.-
Roving Fire Watch 321 E.
Fire Protection Systems Inspection.......
32?
5.0 Major. Design Features 330 5.1 Site Features.................
330
- 5. 2 E Re ac to r....................
330 5.3 Reactor Yessel 330 5.4 Containment..................
330
~5.5 Fuel Storage 330 5.6 Seismic Design 331 6.0 Administrative Controls 332 6.1 Organization 332 6.2 Review and Audit 332 6.3 Procedures 338 6.4 Actions to be Taken in the Event of a Reportable Occurrence in Plant Operation 346 6.5 Actionsto be Taken in the Event a Safety Limit is Exceeded...............
346 6.6 Station Operating Records........
346 1
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t.15) 0F T A&LES (Cont'd)
Table
. Title Page No.
4.2 F Minimus Test and Calibration Frequency for Surveillance Instrumentation 105 4.2.G Surveillance nequirerents.for Control Room Isolation instrumentation 106 4.2.H Minimun.lest and Cal!biation Frequency for Flood Protection Instruinentation 107 Seismic Monituring Instrument Surveillance 108 l4.2.J.
- 3. $. I MAPLER u AversAe, Planar Exposure 171,172,172-a 3.6.H Shock Suppressors (Snutbers) 190 4.6.A Reactor Coolanc System Inservice Inspeccien 209 Schedule 3C A Primary Containmenc Isolacion Valves 250 3.7.5.
Testable Penetrations with Double 0-Ring Seals........
256 3.7.C Testable Penetrations with 1estable Gellows....
257 3.7.0 Primary Containment Testable Isolation valves...
250 3.7.E Suppression Chamber influent Lines Stop-Check Globe Valve Leakage Rates............
263 3.7.F Check Valves on Suppression Chamber influent Lines 263 3.7.H Testable Electrical Penetrations 265 4.8.A Radioactive Liquid Waste Sampling and Analysis 287 4.8.3 Radioactive Gaseous Waste Sampling and Analysis..
283 3.11. A Fire Protection hyste, Hydraulic Recuirewnts...
324 6.3.A Pmtection Factors for Respirators 343 6.8.A Minimum Shi f t Crew Requirenents..........
360 vil Amendment No. 76
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L15T0FtLLUSTRAT103 g
T1 t i e.
Pac tio.
3 2.1.1 APM Flow Reference Scram and Da Rod Stoc'r Settings....................
13 2.1 2 APRM. Flow Blas Scram Ys Reactor Core Flow 25 4.1 1-Graphic Aid in the Selection of an Adequate Interval Between Tests 49 4.2 1 System Unavailability...............
119 3.4 1 Sodium Pentaborate Solution Volt.me Concentrstica Requirements 133 3.4 2
. Sodium Pentaborate Solutten Te perature Requirements 137 l3.5.K-1 MCPR Liraits 172b 352
.Kt Tactor.............
173 3.6 1 Hint:num Tee perature *F Above Change in Transient T ee.pe ra t u r e....................
188 3.5 2 Change in Charpy V Transt tlen Ter.Serature vs.
Neutron Exposure 109 6.1 1 TVA Office of Power Organization for Operatico of Nuclear Pewar Plants.............
351 6.1 2 Functional Organization.........
362 6.2 1 Review and Audit Function.............
353 6.3 1 In Plant Fire Program Organization 364 vill Amendment No. 76
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% ' T9 ES AFETY ' LIMIT ~
LIMITING SAFETY SYSTEM SETTING kr a.
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2.1 FUEL CLADDING INTEGRITY ih ic FU!L CLADDING INTEGRITY en JAl'Yh
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W For no combination of loop s
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recirculation flow rate ans te thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of. rated thermal power.
(Note:
These settings assume operation within the basic thermal hy'draulic design criteria. Inese criteria are LHGR<13.4 kw/ft for.8x3, 8x8R, and pbx 3R fuel, MCPR limits of
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Spec 3 5.k.
If it is d-cernined that s
either of these desig.: criteria is i
s
- being. violated during operation,-
action shall be initiated within 15 minutes to restore operation within prescribed. limits. Surveillance requirements for APRM scram setpoint are given in specification 4.1.3.
)
2.
APRM--When the reactor mode switch is in the STARTU? POSITICN, the APRM scras shall be set at less than or equal to 15% of rated power.
3 IRM--Ine IRM scram shall be set at 3
less than er equal to 120/125 of full 23 scale.
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B.: Core Thermal Power Limit 3.
APRM Rod 31ock Trio Setting
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(Reactor Pressure 6800 osia)
The APRM Rod block trip setting hf WhenLthe' reactor pressure is; shall be:
Yg
.less than or equal to 800 psia, a
Amendment No. 76 0
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q3s SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING I
I J.1' FUEL CLADDING INTEGR1TY~
2.1 FUEL CLADDING INTEGRITY be or core coolant flow is less SRB f (0.66W + 42%)
' than.105,of rated, 'the core P
thermal power shall not ex--
where:
coed.823 MWt (aocut 25% of rated thermal powar).
Sg3 : Rod block setting is percent of rated thermal power (3293 M1t).
W
= Loop recirculation flow rate in' percent of rated (rated loop recirculation P
q flow rate equals 34.2 x 106 lb/hr) s
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C.
Whenever the reactor is C.
Scram & isolation-1533 in above in tLe shutdown conditien reactor low water sessel zero leve".
reactor vessel, the water level shall not be less D.
Sc.aan--turbine step 610 percent than 17.7 inches above the valve closure valve elesure.
top of the normal active fuel tene.
E.
Scram--turbine control valve Upen trip i
1.
7sst Closure of the fast acting solenoid valves.
4 2.
Loss of Centrol 2 550 psig oil pressure F.
Scram--low con-
- 2. 23 inches denser vacuum
-Hg Vacuum G.
Scram--main steam 610 percent line isolation valve closure H.
Main steam isolation 2325 psig valve closure--nuclear system icw pressure y
10' Amen' ment No. 76 d
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1.1' BASE 5_
necause the bnfling transition correlation is based on a IP ge quantity et 4 of s fuel full senle data there is a very high confidence that opersta assembly at the condition of HCPR = 1.07 vould not produce bO.11c4 tran.
.Thus, although it is not required to establish the sa.'ety 11=1 sition.
additional mar;in exists between the safety limit and the actual occurence of losa of cladding integrity.
However, if boiling trannitien vere to occur, clad perforation vould not C1sdding ter.peratures veuld increase to tpproxtactely be expected.
11000T vhich is below the perforation tenperature of the cicoding This has been verified by tests in the General Ilectric Test material.
Beactor (Cl*la) where fuel si ilar in desigs to Efi? operated above the critical heat flux for a significant period of ti=e (30 miautes) without clad perforat en.
If reactor pressure should ever exceed ih00 psia during nor:-1 power
. operating (the li=it of applicability of the boilir.6 trsasition corre-lation) it vould be assu=ed that the fuel claddir.g inte=rity Safety L1=it has been viciated.
At pressures bslev 803 psia, the core elevation pressu e dmp (0 p<ner.
O flow) is greater than b.56 pai.
At lov powers and flovt this pressure differential is maintained in the bypass region of the core. Since the pressure drcp in the bypa'ss regien is essentially all ele 7stien head, s be greater the core precsure d cp at lov powers and flev vill alve."b lbs/hr bundle thsn 4.5C pei. Analyses show thet with a flev of 2SX10 f1:v, bundle pressure drop is nestly independent of bundle power and has the bundle flov vith a h.56 psi driving head
-a value of 3 5 psi. Thus,3 Full scale ATLAS test data taken vill be greater the.n 29x10 lbs/hr.
at pressurco from ik.7 psia to 300 psia indicate that the fuel asce=bly With the desis:
critical power nt this flov is approxt:stely 3.35 NVt.
pes.ing facters this corresponds to a cere therr_si p:ver os core than 50%.. Thus, a core ther=al power limit of 25% for reactor pressures belov 600 psia is ccnservative.
1 for the fuel in the cere during periods when the reactor is shut dovn, ec=-
cideration must also be given to.vnter icvel requirements due to the effect of decoy hent.
If vster level should drop below the top of th* fuel du.-Ing tal: tine, the ability to remove decoy heat is reduced. This reduction in cooling entability eculd lead to elevated cladding te=peratures and elad perferation. Ar long as the fuel raesins covered with vnter, cufficient cecling is available to prevent fuel clad perforation.
Amendment flo. 76
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- '2.1 BASES:
LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL.
L
. CLADDING INTEGRITY
_The abnormal operational transients applicable to operation'of the Browns Ferry Nuclear Plant have been analyzed throughout the' spectrum of. planned operating j;
conditions up to the design thermal power condition of 3440 MWt..
The-analyses were based upon plant operation in
- 4.ccordance with the operating map given in Figure 3.7-1 of the FSAR.
In addition, 3293 MWt.is the licensed maximum power level'of Browns Ferry Nuclear Plant, and this-represents ths' maximum steady-state power which shall not i
p knowingly be exceeded.
L Conservatism is incorporated in the transient analyses in l
estimating 1the controlling factois, such as void react!vity coefficient, control rod' scram worth, scram i
delay time, peaking f actors, and axial. power shapes.
p These. factors are selected conservatively with respect to L
their effect on.the applicable transient results.as determined by the current analysis model.
This transient model,, evolved over many years, has been substantiated in operation as a conservative tool.for evaluating reactor
. dynamic performance.
Results obtained from a General l
Electric boiling water reactor have been compared with
' predictions made by the model.
The comparisions and results.are summarized in References 1, 2, and 3.
i The absolute value of the void reactivity coefficient used L
in the analysis is conservatively estimated to be about p
25% greater than the nominal maximum value expected to occur during the core lifetime.
The scram worth used has been derated to be equivalent to approximately 80% of the total scram worth of the control rods.
The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay a<d slowest insertion rate acceptable by Technical SpecificGtions as further described in reference 4.
The effect of scram L
worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion.
.The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20% insertion.
By the time the rods are 60% inserted, approximately four dollars l-of negative reactivity has been inserted which rtrongly L
turns the transient, and accomplishes the desired effect.
The times.for 50% and 90% insertion are given to assure
. proper' completion of the expected performance in the earlier portion of the transient, and to establish the l
' ultimate fully shutdown steady-state condition.
L For analyses of the thermal consequences of the transients a MCPR > limits specified in specification 3.5.K is l:
conservatively assumed to exist prior to initiation of the transients.
This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than
- .would result by using expected values of control parameters and analyzing at higher power levels.
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. Amendment No'.'75 19
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? ^4.Ei Io summary :
The licensed maximus power level is J,293 MVt.
1.
Asalyses of transients t= ploy adequately ecoservative values of the 2.
controlling reactor parage:ers.
The abnormal operational transients were analy:ed to a power level of 3 3.
is a = ore logical answer than The analytical proceiures cov used result j
the alterns:ive cethod of sagasing a higher a:a :isg pewer is con u=c-4, tion with the expected values for the parameters.
The bases for individual se: points are tiscussed below:
Heuteen Fluz Scres
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APRM h.gh Flux Scras Trip Setting (Run Mode) 1.
aystes, which is calibrated The ave:ase pcuer range coattoring (APRM) heat balance data take fission chambert pro-usist MVC).
Because is percent of rated power (3,29:he A?Pr. syste= responds directly to vide the basic input si gn al s,
the 1:stantaneous rate of During trs:sients, is less than the everage neutron flux.
tronofer from t5e fuel (rce: tor thernal power) instan:angour neutron flux due to the time toeste.nt of the fuel.
heat the thermal 1:duced by dicturbancca, Therefore, during transients iudicated by the neuttso flux pever of the fuel vill be less than :hatAnalyses repor:ed in Sectie scras trip 14 of the Pisal at the serts setting.
demons: s:ed that vi:h a 120 per:en:
Safety A=slyste Repor:se: ting, none of the ab:or=al opers: ions! ::acelents sr.
a:4 there is a substan:is =argin f rem f uel the fuel safe:y 11:1:
(Iow-biased scras provides eves addi:io:al da aye.
Therefore, use of a serr.n as a function of c4: in.
Tigure 2.1.2 shows the flow biase:
core flow.
setting would decrease the unr;is pre-An increase in :he A?Rlt s: r t:
is reacnel. The tu er,rity safety 11:1:
r fuel c~.tddin:
sent before the ired Pet:ing was JeTer..n:t by an analysis of sar3 :s ::qu A?RM sc ri:
to provide e re.saunable range for =asauverist during operation.
the freques:7 of r.purious Reducing this operatin =artin sould increaseen res::or safe:y because of the scrara, which have 43 adverse e!!ee:the AFP,X at::ing was celected
- Thus, resulting thcrtal stressee.
adequa:e ar-in for the fuel cledding is:egrt:7 bacsuse 1: provides reduces ta possibility of safety 11:1: ye: ellevs operating =argin thr:
u=s s e s s s a ry a crtaa.
13 Amendment No. 76 m
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2.1 BASES I
Analyses of the limiting transients show that no scram adjustment is required to assure MCPR A 1.07 when the transient is initiated from MCPR limits specified in specificatien 3 5.k.
For operation in the startup mode while the reactor is at icw pressure, the APRM scram sstting of 15 percent of rated power provides adequate thermal nargin betweea the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor, cold water frem 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 uniferm by operating procedures backed up by the rod worth minimiser and the Rod Sequence Centrol System. Thus, all of possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Becauss the flux distribution associated with uniform rod withdrawals deer 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 uniform rod withdrawal approach to the scram leve, the rate of power rise is no more than 5 percent of rated power per =inute, and the APRM system would be more than adequate to assure a scram before the power could exceed the sr."ety limit.
The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressurer is greater than 850 psig.
- 3. IRM Flux Scra-: Trio Setting The IRM System consists of 3 chambers, t in each of the reactor protection system logic channels.
The IRM is a 5-decade instrument which covers the range of power. level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in si:e.
The IRM scram setting of 120 divisions is active in each range of the IRM.
For 21
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.1 5 AS ES_
3.
IRM T1um Scram Trfe.Cetting (ContinueJJ t
example, if the inst..aent were on range 1, the scram setting would be at 120 divisions for that range; likewise, if the instrueent was on range 5 the secas setting would be 120 divisions on that range. Thus, as the IxM is ranged up to acconmodate the increase in power level, the scram setting is also ranged up.
A 120 divisions on the IKM instruments remains in effect as long as the scran at reactor is in the startup mode. In addition, the APRM 15% scram prevents higher power operation without being in the RUN mode. The IRM scram provides The protection for changes which occur both locally and over the entire core.
nost significant sources of reactivity change during the power increase are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of v,ithdrawing control rods, that heat flux is in equilibriu= with the neutron flux and an IRM scram would result in a reactor shutdown well before any safety limit is exceeded. 7or the case of a single control red withdrawal error, a range of rod withdrawal accidents was analyzed. This analyels included starting the accident at various power levels. The most severe es,e involves an initinI condition in which the reactor is just suberitical and the IRM system is not yet on seale. This con'dition exist: at quarter red density. Quarter rod density is illustrated in paragraph 7.5.5 of the FS AR.
Additional conservatism was taken in this analysis by assu=ing that the IRM channel closest to the p
withdrawn red is bypassed. The rasults of this analysis show that the reactor is' scrammed and peak power li=ited to one percent of rated pewer, thus maintaining MCPR above 1 07. Based en the above analysis, the IKM provides protection against local control red vichdrawal errors and continuous withdrewal of control rods in sequence.
l
- 8. APRM Control Rod Black Reactor power level =ay be varied by =oving control rods or by varying the recirculatien flow rate.
Ihe APRM sys tem provides a control red block to prevent rod withdraval beyond a given point at constant recir-cuciation flow rate. and thus to protect against the condition of a HC7R Icss than 1.07.
This red block trip setting which is automatically varried sich recirculation loop flow rate, prevents an increase in the r actot pcwer level to excess values due to control red with-d e sv 41.
The f1:w variable trip setting prevides substantial =argin 22 Amendment No. 35, M l
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W 2.1 BASES from fuel damage, assumir.4 a steady-state ' operation at the trip setting, over Lehe entire recirculatinn flow range. The margin to'the Safety Limit increases as the flow' decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is
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at 100% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the. core is established by specified control red sequences and is monitored continuously by the in-core LPF3 system.
C.
Reactor water Lov'tevel' Scram and Isolation (Except Main Steamlines)
The^ set point for the lov level scram is above the bottom of the separator skirt.
This level has-been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR subsection 14.5 show that scram and isolatica of all process lines'(except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than 1.07 in all cases, and system pressure does not reach the safety valve settings. The scram setting is approximately 31 inches below normal operating range and is thus adequate to avoid spurious scrams.
D.
Turbine Sten Valve' Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and' heat flux increases that would result from closure of the stop valves.
With a trip setting of 10% of valve closure from full open, the resultant p,
T-increase in heat flux is such that adequate thermal margins are maintained even.durirg the worst case transient that assumes the turbine bypass valves remain closed. '(Reference 2) 5.
Turbine Centrol Valve Seram 1.
Fast Closure Scram This turbine control valve fast closure scram anticipates the pressure.
neueren flux, and heat flux increase that could result frem fast closure of the turbine contcol valves due to load rejection coincident with failures of the turbine bypass valves. The Reactor Protection System initiates a scram when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of control valve fast closure. This is achieved by the action of the f ast acting solenoid valves in rapidly reducing hydraulic control oil' pressure at the main turbine control valve actuator dise dump valves.
This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-tvice 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
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produce transients very similar to that for the stop valve. No signifi-cant change in MCPR occurs. Relevant transient analyses are discussed in References 2' and 3 of the Final Safety Analysis Report. This scram is bypassed when turbine secam flow is below 30% of rated, as measured by_ turbine first state pressure.
Anendment No. 76 23
2,1 EASES 2.
Scram on loss of control oil pressure The turbine hydraulic control system operates using high pressure oil. There are seversi points in this oil system whers a loss of oil preseure could result in a fast closure of the turbine control valves. This fast closure of the turbine control valves is not protected by the genera, tor load rejection scram, sizes failure of the oil system would r.ot result in the fast closure solenoid valves being actuated.
For a turbine control valve fast closure, the core vould be protected by the APRM and high reactor pressure scrams. However, to provide the same margins as provided for the generator load rejection scram on fast closure of the turbine control valves, a scram has been added to the reactor protection system, whic; senses f ailure of control oil pressure to the tur-bine control system. This is an anticipatory scram and results in react +r shutdown before any significant 1 :rease in pressure or neutron flux occurs. The transiest respecse is very similar to that resulting fr ia the generator load rejection.
F.
Main Condenser Lev Vacuum scram To protect the main condenser ag. inst overpressure, a loss of coo-denser vacuum initiates automatic closure of the turbine stop valves and turbine bypass valves. To anticipate the transient and autematic scram resulting f rom the closure of the turbine stop valves, low con-densar vacuum ini'tiates a scram. The lov vacuum scraa set point is selected to initiate a seras belc.e the closure of the turbine stop valves is initiated.
C. & H.
Kain. Steam Line Iss.mtion on Lev Pressure and Main Steae Line Isolation Scram The low pres sure isolation of the =41n steam lines at 825 psig was provided to protect against rapid resator depressuri:ation and the resulting rapid cooldown of the vessel. Advan: age is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdova so that high power opera-tion at lev reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reac-tor at pressures lover than 825 peig requires that the reactor mode switch be in the STARTUF position, wher e protection of the fuel cladding integrity saf ety limit is provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of ceutron flux scram protsetion over the entire rangs of applicability of the fuel cladding inceprity safety limit.
In additien. the isolation valve closure ee rse anticipates the pressure and flux transiaste that occur during normal or inadvertent isolation valve closure. Vith tbs scrans set at 10 percent of valve closure, neutren flux does not increase.
2b
a 9
2.s m y
- 1. J. 6 1.
Pa se t er low matee level set estat f or tat t f at ion of ttat! and setC, c l e s s a '- enin steam isolation valves, and s t a r s tar (Jf.!
and core spray puis pe.
These systess mJintain adequate coolant inventory and ptnvide core cooling with the objective of preventing e.acessive clad temperatures.
The des tta of etiene systens :e adequately perf ors the latended func-tien is based en the specified low levet scene ses point and initis.
tien met points, fransient analyses tepersed in Sectlen 16 of the 13 AA deoenstrate that these tend 1ttens resielt in adequate selety settint f or both tt.s fuel and the systes pressure.
L.
Referencis, 1.
(Inford. P. 8.
Malytical Methods of Plant Tr ans tant tva twettoes f or the Centr al Elect ric letting' Vater teacter " KtDo-10802, reb.,1913.
- 2. Generic Reload Tuel Application, Licensing Topical Report.
NEDE-240ll-P-A, and Addenda.
3.
" Qualification of the One-Dimensional Core Transient hodel for Boiling Water Reactor", NEDO-24154, NEDE-24154-P, October 1978.
4.
Letter from R. H. Buchhol: (CE) to P. S. Check (NRC), " Response to NRC request for informstion on CDJ$ computer =odel," September 5, 1350.
e 25 Amendment No. 76
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Revised 1-17-79 l
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4 1,2 BAS 3 "herete-e,.*ollowin; ar.y trun:icnt pressure seniter higher in the vessel, iftd that is severe enogM to suse cenetra that this sr.fety 1141% was v c.a e,
a :siculatica v111 be perfereed using att availakis in'erratice to cat.:r.
eine if the esfety limit was vicisted.
?D?"CCi$_
1.
Plent Se* sty Anslys'is (IT.TPTSARSreticaIL.d Sectita II:
/.3:2 '.%1ler and Pressurt Vessel Cod:
2, C'.3 Pipirc Code, "cetica 331.1 3
(F.:P F.Aa at,et:r *. ::et one Appurtet.anets !!achsni:s1 ;4s ga h,
Seesee:1ca *. 2)
~
5-Generic Reload Fuel Applicag4 81 3 IOPital Reporr.,
." ' ~
N Q E-240ll-P-A and Addenda.
29 hendment No.
35, 4 [ 59 FEB 25 1980
f 2.2 BASES REACTOR C00MNT SYST1M INTEGRITY f
To meet the safety design basis, thirteen relief valves have been Lnstalled on the unit with a total capacity of 83.9% of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steenline isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, res its in adequate matgin to the code allevable overpressure limit of 1375 psig.
To meet the operational design, the analysis of the plant isolation transient (generator load reject with bypass valve f ailure to open) shews that 12 of the 13 relief valves limit peak system pressure to a valve which is well below the allevable vessel over-pressute of 1375 psig.
l t
30 Amendment No. 76
(;
r
-I,I!!ITI!!C CO:!DITIO!!S FOR OPERATION SintvEII.I.ANCE 10;O'JIREMEhiS 3.1 FIACTOR PROTECTION SYSTEM 4.1 PEACTOR PROTECTION SYSTEM Applicability' Apolicability
. Applies to the instrumentation Applies to the surveillsace of and associated devices which the instru=entation and asso-initiate a reactor scrau.
ciated devices which initiate reactor scra=.
Obie--ive Objective To assure the operability of the To specify the type and frequency reactor protection system.
of surveillance to be applied to the protection instru=entation.
Specification Specification When there is fuel in the vessel, A.
Instru=entation systens shall
~
.the setpoints, =ini=um nu=ber of be functionally tested and trip systens,.'and =ini=u= number calibrated as indicated in of instru=ent channels that must Tables 4.1.A and 4.1.3 respec-be cperable for each position of' tively.
the reactor = ode switch shall be as given in Table 3.1.A.
1
?
C.
When it is deter =ined that a i
chcnnel is failed in the ensa.#e condition, the other R?S channe that =enitor the same varichle i
shall be functienally tested i==ediately before the trip sys tem centaining the failure is tripped. The trip syste: con-taining the unsafe failuro n.'.:.
untripped for short periods of ti=e to allow functional tectin of the other trip system.
The trip syste= =ay be in the untripped position for no = ore than eight hours per functional db test period for this testing.
Amendment No. 76 31
9 1
[-
9 r.
L.'
Pact DELEng a
32
...n a
f: -
'=
. 4.1 8 ASF.S '
Tbc frequency of calibration of. the 'ADPM F]ow Biasing Network has been established as'each refueline, outage. There are several instruments wnich must be calibrated 'and it will take several hours to perform the
.calibrationLof the entire network. While the calibration is being per-formed a zero flow signal will be sent to half of the APRM's resulting
-in'n; half scram and rod block condition. Thus, if the calibtstion were performed during operation, flux shaping would not be possible. Based
.on experience at-other generating stations, drift of instruments, such aus chose in the Flow Biasing Network,- is not significant and therefore.
-to avoid: spurious scrams, a calibration frequency of each refueling out-age is established.
Group (C) devices are active ~only durint e given portion of the opera-tional cycle. For example, the IpJt'is active during startup and inactive during full-power operation. Thus.the only test that is meaningful is the one performed just prior. co shutdown or startup:
1.e.,
the tests that-tre performed just prior to use of the instrument.
Calibration frequency of the instrument channel is divided into two groups. -These are as follows:
-. l. Passive type indicating devices that can be compared with like units on a continuous basis.
2.
Vacuu:a tube or semiconductor devices and detectors that drfit or lose sensitivity.
Experience with passive. type instruments in generating stations and sub-stations indicates that.the specified calibrations are adequate. For
-those devices which employ amplifiers, etc., drift specifications call for drift to be less than 0.4%/ month: 1.e., in the period of a month a drift of 4%'would occur and thus providing for adequate nargin. For-the APRM system drift of electronic apparatus is not the only considera-tion in determining a calibration frequency. Change in power distrihu-~
tien and less of chamber sensitivity dictate a calibration every seven days. Calibration on this f requency assures plant operation at or belew ther=al limits.
A ecenarison cf Tables 4.1. A and 4.1.B indicates that two instrument
. channels have been. included in the latter table. These are; mode switch in shutdown'and manual scram. -All of the devices or sensors associated with these scram functions are simple on-of f st: itches and, nence, calibratien during operation is not anolicable, i.e., the switch is either en or of f.
47 Amendment No. 76 L
$5 s
- v' en
)
7r7
~ N.1 - BASES -
P 4
J 'a LThe sonsitivity of LPRM detectors -decreases with exposure to. neutron flux
. at a-slow and approximately ' constant rate. The APRM system, which uses the LPRM readings to: detect a change in thermal power, will be calibrated every ryen _ days using a heat balance 'to compensate for this change in sensitivity. The RBM system uses-the LPRM reading to detect a localized chsnge in thermal power. It applies a correction factor based on the APRM
-1 output signal to determine the. percent-thermal power and therefore any
' change in.LPRM' sensitivity is compensated _for by,the APRM calibration. The technical specification: limits of.CMFLPD, CPR, and MAPLHGR are determined by the use of the process computer or other backup methods. These methods
-use LPRM readings and TIP. data.to determine'the power distribution.
Compensation in the process computer for changes in LPRM sensitivity will
.be made by performing a full core.TIP. traverse to update the computer calculated LPRM correction factors every 10n0 effective full power hours.
As a minimum the : individual LPRM meter readings will be adjusted at the beginning* of. each' operating cycle before reaching 100 percent pcwer.
w 48 Amendment No. 76
TABLE 3.2.C 1HSTRtalD. TAT 10't 11 TAT lillTIATES ROD BIDCK.S F
a al.t x..
Operabis Fer Trip Level Settfor, Tunction 1rty Sys (5)
< 0,66u + 421 (2) et u
2(1)
AFR1 Upscale (Flow Biss) f 4 122
^*
2(1)
AFut Upecale (Sta tup Mode) (8) w 2(1)
AFRM Dovascalt (9) 1 31 76 u
(10 )
2(1)
AFRM Inoperative b
< 0.66W + % (2) 1(7)
Rpt Upscale (Flow Blas) 1 3Z 1(F)
R3M Downecale (9)
(10 )
1(7)
RBri Inoperative g
_ 108/125 of full scale 3(1)
Itti Upscale (8) 3 3/123 ef full scale 3(2)
IRM Dovuscale (3)(8) 1 3(1)
IRH Detector not in Startup Position (8)
(11)
(10*)
3(1)
IU( Inoperat ive (0) j 1 x 10' counts /sec.
2(1)(6)
LRH Upecate (8) 2(1)(5)
SMt Dawnscale (4)(8)
L 3 counteleec.
i 2(1)(5)
S2:1 Detector not in Startup Position (4)(8)
(11)
(102) 2(1)(6) 52M Inoperative (8)
< 1')I dif ference in recirculattom flows 2(1)
Flow Blas Ceeparator
,.1101 recirculation flow 2(1) riou Bias Upecale N/A kod Block Lo RSCSHeatrairgte:
lle7 psig turbine 1 )
2 rirst stsee pressure (approx:.mately 301 power)
(PS-85-61A &
PS-85-610)
f_
9 E
NOTES FOR TABLE 3.2.C i:
- 1.
-For-the startup and run positions of the Reactor Mod > Selector Switch, there chall be two operable or. tripped trip-systems for each function. The SRM,
?IRM, and APRM (Startup Mode), blocks need not be operable in "Run" mode, and the APRM (flow biased) and iU3M rod' blocks need not be operaale in "Startup"
-mode. ~If'the first. column cannot.be met for.one of the two trip systems, this' condition may exist for up to seven days provided that during that time the operable; system is functionally tested immediately.and daily thereafter.
If.'this condition-lasts longer than seven days, the system with the inoperable channel shall be. tripped.
If the first column cannot be met for both trip :syste=s, both trip systems shall be tripped.
2.-
W.is the recirculation loop flow in aercent of design.
Trip level setting is in percent of rated power (3293 M4t).
See Specification 2.1 for APRM centrol rod block setpoint.
- 3. ~ IBM deknscale is bypassed when it is on11ts lowest range.
~
- 4. ' This function is bypassed when the count rate is 3L 00 cps and IRM above 1
range 2.
-5.
One instrument channel, i.e.
one APRM or IRM or R3M, per trip system may be bypassed except.only one of four SRM =ay be bypassed.
6.
IRM channels A, I, C, 0, all in range 3 bypasses SRM channels A a C' functions.
(
IRM channels 3, F, D, H, all in range S bypasses SEM channels 3 & D functions.
(
.7.. The.following operational restraints apply to the R3M only, s.
Both RBM channels are bypassed when reactor power is di 30%.
b.
The RBM need not be operable in the "startup" position of the reactor mode selector switch.
.c. _ Two 33M channels are provided and only one of these may be bypassed from the console. An RBM channel =ay be out of service for testing and/or maintenance provided this condition does not last lenger than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
-in any. thirty day period.
d.
If minimum ~ conditions for Table 3 2.C.are not net, administrative
! controls shall be i==ediately imposed to prevent control red withdrawal.
. Amendment No. 76.
74 w x:
g TADM 3.2.F Surveillance Instruisentation a
n t'inimu.a I of f
Operable Insttument Tyre Ir.llcat.lon Channels Instnernent #
Insttument, arut liar.>,e Notes 2
!!11 94 Dryvell and Torus 0.1 ' 2uf.
(1) 2 Ilydrogen Il H 104 Concentration 2
5 2
PdI-64-I 's /
Drywell to Suppression Indicator (1) (2) 5)
PdI-64-138 Cliamber Dif ferential 0 to 2 paid pressore-
- r t
S it NO?TS FOR T A.B! E 3. 2. ?
(1)- From and af ter the date that one of these parameters is reduced to one indication, continued operatien is permissible during the succeeding thirty days unless such instrumentation is sooner made operable..
(2) From and af ter the date that-ene of these parameters is not indicated in the centrol room, centinued operation is permissible during the eu=ceeding seven days unless such instrumentatica is sooner made operable.
(3) If the require =ents of notes (1) and (2) cannot be met, and if one of the indications cannot be restored in (6) hours, an orderly shu:down shall be initiaced and the reactor shall be in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(a) These surveillance instrurents are considered to be redundant to each other.
(5) From and af ter the date that both the acoustic monitor and the temperature indication on any one valve fails to indicate in the control roos, continued operation is permissible during the succeeding thirty days, unless one of the two monitoring channels is sooner =ade operable. If both the pri=ary and seconds y indicstion on any SRV tail pipe is inoperable, the torus temperature vill be =enitored at least once per shift to observe any unexplained te=perature increase which might be indicative of an open SRV.
j i
I I
SQ.
AmeMment No. ff, 71
. 31TINC CONDITIONS FOR OPEhTION SURV7.!LI.t. IC? REQUIFJ_tOIT3 3.3.A REACTIVITY CONTROLS _
6.3.A REACTIVITY CONTROLS Control rods wLth scras b.
A second liceosed operstor c.
times greater than thosa shall verify the confor-paraitted by Specifica-
=snee to Specification tion 3.3.C.3 are inoper-3.3.A.2.d befors a rod may able, but if they can be be bypassed is the Rod inserted with control rod Sequence Control Systas.
drive pressure they need not be disarmed electri-c.
hhen it is initially deter-cally.
nined that a control cti is inespable ei nor-si inscrtian d.
Control rods with a failed an attempt to fully insert
" Full-in" or "Tull-out" the control red shall be position swLtch may be by-nsde.
If the contral red passed in the Rod Sequence cannot be fully inscrtcd,.2 Control Synten and conat-shutdot.m nargin test shall dered operable if the actus1 he nsde to demonstrate und rod position is knavn.
These this cor.dition tSat the :Or-rods must be moved in sequence can be cade subcritica' ie.-
to their correct position, sny reactivity conditicn (full in on insertion or full during the re-ainder of the out on withdraval),
operating cycle with the afinlytically deternined, Control rods with inoperable highest worth control rod' e.
accumulators or those whose capabic of ui thdrausl, full position cannot be positively wi,thd :vn, sai all other d.ecernined shall be consi-control rods capable of l
deced inoperable.
insertion fully instrted.
f.
Inoperable control rods shall d.
The control red accumulators be positioned such that Speel-shall be dete nined operabic fication 3.3.A.1 is =et.
In at least once per 7 days by addition, dueing reactor power verifying that the p'r-s;ure operation, no more than one and level detector < are not te control rod in any 5 x 5 array the alar =ed condition, nay be inoperable (at least 4 operable control rods =uat seperste any 2 inoperable i
ones).
If chie Specifica-tion cannot be met the reac-I' tor shall not be started, or if at power, the reactor shall be Stought to a shut-down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Control Rods B. Centrol Reds 1.
Each control rod shall be 1.
The c:uplins inesgrity shall be coupled to its drive or verified for each withdrsvn emn-completely inserted and the trol ad as follevs:
121
' JAN 10 SM
f.tMITINC C0wntT10N3 FOR OPERATION SURVE!!,t.ANCF. REQUIREMINTS 5 Cenerei Rode 4.3.3 Control Rods seetrol rad directional Verify that the control red control valves disarmed 8'
electrically. This require.
is following the drive liy ment does not apply in the observing a response in the refuel condition when the nuclear instrumentation each reactor is vented. Two con-time a rod is moved when trol rod drives may be removed the r.eactor is operating as long as Specification above the pre-set power 3.3.A.1 is met, level of the RSCS.
b.
When the rod is fully with-drawn the first ti=e after e.ach refueling outage or after.sintenance, observe that the drive does not go to the overtravel po;ition.
2.
The control rod drive 2.
The control rod drive housing housing support systes shall be in place during reactor support systets shall be inspected after reassembly and the results Power operation or when the of the inspection recorded.
reactor coolant system is Pressurized above ot=oepherie 3.a.
Prior to the start of control prewaure with fuel in the reac.
rod withdrawal'at startup the cap-tor vessel, unless all control ability of the Rod Sequence System roda are fully inserted and (RSCS) to properly fulfill its functions Specifica tion 3.3. A.1 is met.
shall be verified by the following checka.
3.
a.
Whenever the reactor is in the startup or run modes Cequence portion - Select a sequence below 2 S rated power the and attec:pt to withdraw a rod in the Rod Sequence Control System remaining sequences. Move one rod (RSCS) shall be operable in a sequence and select the remain-except the RSCS constraints ing sequences and attempt to move may be suspended by neans of s rod in each. Repeat for all the individual rod bypass scouences, switches for 3roup notch portion - For each o' the 1 - special criticality six comparator circuits go through tests. or test initiate: comparator inhibit; 2 - control rod sctam timing verify: reset. ^n *-" W arremne per 4.3.C l.
test is allowed en continue until When RSCS is bypassed on completion is indicated by individual rods for these illumination of test complete light.
exceptions EbH must be oper-able per 3.3.3.3.c and a second licensed operator may not be used in lieu of RLH.
122 Amendment No. 76
f O
~
SimVP.1IJ.ANCE f a ou r itr>tt u 3 f,fMITINC CONDITTOMS Fon OPFRATTON
,4.3.R Control Rnde
. 1. 3. 11 Contral_f, ode b.
Prior to attaining 20% rated power b.
During the shutdown procedure no rod movement is permitted during rod insertion at shutdown the herween the testing performed tests in 4.3.3.3.a shall be performed above 20% power and the rein-to verify RSCS capability, statement of the RSCS re-neraints at or above'207.
c.
The capability of. the Rod Worth power. Alir,nment of rod Minimizer (RWM) shall be verified groups shall be accomplished by the following checks:
prtor to performing the tests.
Whenever the reactor is in the startup or run modes 1.
The correctness of the c.
control rod withdrawal below 20% rated power the
- Rod Worth Minimizer shall be sequence input to the operable A second licensed RWM computer shall be operator may verify that verified befo e reactor startup or shutdown.
the operator at the reactor consola is following the 2.
The RWM computer on linc control rod program in lieu diaccostic test shall be of RWM except as specified successfully performed.
in 3.3,B 3.a.
3.
Prior to startuc. proper annuncistion of the selec-tion error of at least one out-of-sequence control rod shall be verified.
4 Prior to startup, the red d.
If Specificationa 3.3.B.3.4 throuph.c cannot hs met the block function of the RUM ronctor sha!! not be ntseted.
shall be verified by movinn nr if the reactnr la in tha an out-of-sequence control run or ntartup modea at len8
- rod, than 20% rated power. it shall be brought to a shut-5.
Prior to obcAining 20" rated down condition tenediately, power during rod insertion ~
at shutdown, verify the latching of the proper rod group and proper annunciation after insert errors.
d.
When the RWM is not operable a second licensed operator will verify that the correct rod program is followed except as specified in 3.3.3.3.a.
123 Amendment No. 76
1.8tilftna en utftnNs rot OPERAT!n1 51JtvrlLt.ANer, acquitocwT3
)
3.3.3 centrol a.de 4.3.8 Concret Rode 4.
Centrol rods shall not be withdrawn for startup er refweling unless at least tse source ranze channeta have an observed count rate eque", to er greater than 4
Peter to control red withdrawel threr, couts per second.
for startup or during refueling.
verify that at least two source S.
.Durin: operation with range channels have as observed limiting control rod pat-count rate of at least three torns. as determined by the counts per second.
designated quali!1sd person.
mel, either S.
Vhen a limiting control rod a.
goth RtM channeta shall pattern entsts, an instruzent be operabia:
functional test of the R3M er shall be perf orwed prior to i
withdraval of the destssated b.
Control rod withdrawal red (s) and at least, esce per shall be blocked.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thermatter.
C.
Scram Insertion Times 1.After each refueling outage all operable rods shall be scras time tested from the fully vichdr.vn position with the nuclear system pressure above 800 psig This testing shall be completed prior to C. Scras f asereten Times exceeding 40% power.
Below 201 p o w e r',
only. rods is those sequences 1.
The average scram insertion (A
time, based en the deenersi-fullywithdrkhnin t$e) region or 3 and 3 which 12 and A g were sation of the scram ptiot valve f
M dmuY u W M eolenotds as stae sero, of all dessity shall be scram time tested, eperable control rods in the The sequence restraints imposed upon reacter power operation condi-the control rods in the 100-50 tien shall be no greater then:
percent rod density groups to the
- tneerted Fr**
Aws. Scree Inser.
preset pnwer level may be removed Fully Vithdrawn_
tien ?tnes (see) by use of the individual bypses switches associated with those 0.373 control rods which are fully or 20 0.90 partially withdrawn and are not SO 2.0 within the 100-50 percent rod density 90 3.500 groups.
In order to bypass a rod.
the actual rod axial position must be known; and the rod must be in the correct in-sequence position. As required by 3.3.3.3.a a second licensed operator may not he used in lieu of RRi for this testing.
124 Amendment No. 76
3.3/4.3 3Asts:
3.
The Itod Vorth Minimizer (P.'.M) and the Rod Sequence Control System (RSCS) re:si rt.-t virhdravs:s and Ine.crtionn of centrol rode to pre-r.pce tf led r.c<tuencen.
All pat terns str.ociated vitt.
these ccquences have tieu ch.tracterl-t Ic th.it, swuelnr. the worst single devtstion from the sequwnce the drop of any control rod f rom the fully inserted position to the position of the control rod drive would not cause the reactor
- to Austain a power excursion resulting in any pellet average enthalpy in excess of 280 calories per tram. An enthalpy of 280 calories per gram is veil below the icvel at which rapid fuel dispersal could occur (i.e.,' 425 calories per gram). Primary sy'stes damage in this accident is not possible unless a significent amount of fuel is rapidly dispersed.
Ref. Sections 3.6.6, 7.7.A. 7.16.5.3, and 14.6.2 of th. TSAR and N Go-10527 and supple =ents thereto.
'In perfornin, the fdnetien described above, the IC M and F3CS are:
not required to inpose any restricciens at core power levels in excess of 20 percent of rated. I'aterial in the cited referer shows that it.is impossible to reach 280 calories per gram in the event of a control red drop occurring it power greater ensa 2C percent, regardless of the rod pattern. This is true for all normal cnd abnormal patterns including those which maxi =1re individual centrol ros verth.
At power icvels belov 20 percent of rated, abnormal control rod patterns could pr: duce red vorths high enough to be of concern ralt.tive to the 280 calorie ;er gran srd drop 11 nit.
In ghis ranre the 10!M and the RSCS constrain the control red sequenees and pattarna to those wht:h involve only'secept:ble rod worths.
i The Rod Vorth Minimirer and the f.ed Sec;;ence Control Systen provide autenatic supervision to assure that out of sequience control rods vill not be vithdrawn or inserted; i.e., it Liniti operater deviations f rt:s planned withdrsval sequences. Re f.
Section 7.16.5.3 of the ySAR.
Ihey serve as a backup to proceduc#
control of control red sequences, which linit the ua: ir:un resect -
vity worth of control rods. Except during specified exceptions, when the Rod Worth Minimi.:er is out of service a second licensed operator can nanus 11y fulfill the control red pattern con-formance function:: of this syste:3. In this case. the RSCS is backed up by independent procedural controls to assure confor=ance.
Because it is allovable to bypass certain rods in the RSCS during specified testing below 20 percent of rat'ed power in the startup or run modes, a second licensed operator is not an acceptable substitute for the RWM during this testing.
Amendment No. 76 129
r T.he functions of the KkN end RSCS caka it unnec csary to specif y a lleense limit on rod worth to preclude unacceptabic consequences in the event ei a control rod drop. At lov powers, below 20 percent, these devices force adherence to acceptable rod patterns. Above 20 percent of rated power, se constraint on rod pattern is recuire1 to aasure that rod drop accidert consequences are acceptable. Control rod pattern constraints above 20 percent of rated power are imposed by power distribution requirements, as defined in Sectiona 3.5.1, 3.5.J. 4.5.1, and 4.5.J of these technical specifications. ?cwcr level for automatic bypass of the RSCS function is sensed by first stage turbine pressure.
4 The Source Range. Monitor (SRM) ayste:s performs no sucomatic safety system function; i.e., it has no scram functio'n.
It e
130 JAN 14 @
g e
e
- /t.3 nas?%:
Joe. praetde the apsrator with a visual indteattaa of neu-tron livel. The consemiesten u! re.sttete;, accidsc.ts art functionn of the tr.itial nevtren (1Js.
"he requitettnt of at lesit 3 Count.s i.er actiend easures that ar.y trer.t; tent, should it ocrur, nnqins ni ur above the Initt el value of 30 " of rate 1 3.nv e r i.. ? d i n L1.~f in e l y s e s ni transient s f ru.:
~
cold'eandittune. Ona oprabin 1711 chinsel vnald he a,detuste to reunitor tbs approach te ttiticality siang hesoe,rnc9us patterne of scatter:J control red v!thstav41. A sir.i m of two operible SL't's are providad as an added tor.ser r.stian.
S.
The Red Block Mon *1 tor (Mat is dest; sed to automatically prevent fuel dar. age in the evt t of ertoetsus rod vithdraval f rors tocatto.a of high sover de.otty during high po.cr le rel operation. Two channels are providet, tr.d one of these tiay
'ee bypassed froo the censole f or ruintenance and/or testing. 3 Trippins of one of the channels will bloc *a errar. cows red
+
vlthdrawat soon ensuuh to prevent fuel dasap..The spect-fled restrictions with 4ne channel cut of rervice conserva-tivsly as.ute that f ue'l damage vill not occur due to rod withdrawal tirors when this cond!. tion estats.
A limiting control rod pattern is a pattern which resulin in the core. being on a thermal hydraulic ilmft, (ie, HCPR given by See.c. 3.s.K or LHCR of 13.4 kv/fc.
During use of such patterns, it is judged that testing of the R5ft system prior to wit *.-
drawal of such rods to assure its operability will assure that improper withdrawal does not occur.
It is normally the responsibility of the llu c l e a r Engineer to identify these litaiting patterns and the designated rods either when the patterns are initially established or as they develop due to the accurrence of inoperable control rods in other than limiting patterns.
Other personnel qua'lified to per-form these functions may be designated by the plant superintendent to perform these functions.
Scram insertion Times The control rod system is designated to bring the reactor suberitical at the rate fast enough c prevent fuel d a r.a r. c 1c, to prevent the HCPR from becoming less than 1.07.
The limitinC power transient is given in Refsrence 1. Analysis of this trancient shows that the negative reactivity rates resulting from the scram uith the average response of all the drives as given in the above specification provide the required protection, and HCFR remain, greater than 1.07.
On an omrly 8WR. sone degradation o'f control rod scrac petforei.1nce decured during plant startup and was deteroined
- i. he c a u m...' by 1.11 Amendment No. 76
<a K:
o 1~
Q' ' i.
't 3.3/4.3 8A5Fi:
part iculate matertal (pr abably construction Jewrfs) 7.n,Aing an interns 1 control rod drive filter.
The design of the preatne control rod drive (Model 7K351643) relocation of the filter to a totation outis gro sly improved by the of the scra-drive path; i.e., it tan no lenser interfere with scram perfot snce, even if completely blecked.
The degraded performance of the original drive (CP.31RD 1 *.44) under dirty operating renditions and the insensit'vity of the rede = taned drive (CK37%D814&S) has been de=onstrated by a
' sert e of engineering t ent s unde r o1mu tated rr ector optar. ting condLtions.
The ouecessf ul aerfornenen of the n.w d ri v.: unJe:-
actu21 operating conditions has also been deecettrated by toes'atently Roost in-service teet renutto for planco unin. the frase and may be inferred frts plaste using t!.e old. r redel n.v e
, driv wlth a modified (larger sc reen s tae) ints s?.a1 fil r.c r which is 1 se' pruce to plugging.
lancJ, reports in various operattag plaste. Data han been do:u.. nted by s u6.til-Thcen include Oys te r Creek, Monticello. Dresden: 2 an ? Dreedor. 3.
App r. ::1uc t ely 5000 drive tests have been recorded to date.
Tollowing identification of the " plugged filtet" problen, vety f requent scram testJ were necepstry to ensure proper perfrrmarce However, the more frequent scram tests are ecw cL-nideret' tetrily unnecessary and envise for the f alleving ressoriet 1.
Cet<.t'ic scrari perfors nee has been identified as due to ar oboaructed drive filter in type "A" drivec.
ETM? are of the nov "3" type design whose scrc4The driv:r.1s.
is unaffected by filter condition perforr.anc e 2.
The dirt Jood is primarily released during startup of the rescrer when the reactor and its syntese are first to (;ous and, preen' ire and thermal stressee.
subjected Special atten-
. tion end scavures are now being taken to assure cleaner
-systeN-Ree ctore with drives identical or similar (shorter g e rsk a'*
refueling cycles with no suddc5:. smaller platon arese) have operated thr eerfor ance.
or erratic changes in scram
.ufficient to detect ancealous drive perfora4 te.This preoperalfen 3.
\\
h, 72-hour cutage liste which 1sitisted the start of the frequent scher then quantifyingscrim testing is' arbitrary, havin no logical basis a " major outage" uhich =ight bly be eaur.;d by an event so severe as to pocsibly affett reasona-drive performance.
Thle requirement is unwise because it "an line" to avoid the additiont.1 testing due a 72-ho
,4 -
i.
132 l;
l 4
1,
.I i
r
, u jj
>S y }
g
.Q i
s.
e LfMITINC CovDITtoN3 FOR OPERATIDN SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CCRZ AND CONTAINKENT COOLING SYSTEMS SYSTEMS Applicability _
Applicability Applies to the operational Applies to the surveillance status of the core and contain-requirements of the core and ment cooling systems.
containment cooling systees when the corresponding lLaiting condi-tion for operation is in effect.
Ob_lective Objective To assure the operability of To verify the operab/lity of the the core and containment cooling core and containment cool'ng systeau under all conditions for ayatems und6r all conditicas for which this cooling capability is which this cooling capability is an essentist respones to plant an essential response to plant abnormalities.
abno rm alitie s.
g.
\\'
5peciftention Specification A.
Core Spray System (CSS)
A. Core Soray Systea (CSS) 1.
The CSS shall be opera-1.
Core Spray Systen Testing.
ble:
Itm Trecuency (1) prior to reactor startup from a a.
Simula ted once/
cold condition.or Automatic Opera tin 5 Actuation Cycle (2) when there,is Lt r' -'
test a
disted fuel..: th! -
vensel and when ens b.
Pump Opera-Once/
reactor vessel pres-bility conth sure is greater than atmospneric pressure.
c'. Motor Once/
- /
except se specified Operated monen in specification Valve 3.5.A.2.
Cperability d.
System flow Onca/3 rata: Each months loop shall deliver at least 6250 gym against a system head corres-ponding to a I
143 m,
Amendment No. 76 l
.h 0
i LDC'"!NG CONDI"'!oNS FCR CPIRATION SURVEILIJ.NCS DECUITS#ENTS 3.3.A Core Sorav System (CSS) 4.5.A Core Spray System (CSS) 2.
ll One c3S loop is inopera-los pai dif-ble, the reactor may remain ferential V
in operation for a period pressure not to exceed 7 days provi-between the ding all. active components reactor ves-in the other CSS loop and the gel,nd ghe UK system (LPCI mode) and the pgg=gry een.
diesel generators are operable.
tainment.
3.
If specification 3.5. A.1 e.
Check Valve Once/
or specification 3.5.A.2 Operating cannot be met, the reactor Cycle shall be shutdevn in the Cold Condition within 24 2.
When it is determined that one
- hours, core spray loop is inoperable, at a time when operability is 4
When the reactor. vessel required, the other core spray pressure is atmoopheric loop, the RHR$ (LPCI mode), and and irradiated fuel is in the diesel gerterators shall be the reactor vessel ag demonstri Ced to be operable least one core spray loop immediately. The operable core with one operable pu.np and sprav 1Np shall be dernonstrated associated diesel generator to oe operable daily thereafter.
shall be operable, exespt with the reactor vessel head re=cved as specified is 3.5.A.5 or prior to reactor startup as specified in 3 5.A.1.
5 When irradiated fuel is in the reactor vessel and the reactor vessel head is renoved, core spray is =ct required provided v:rk is not in pres ess whi:h has the potential to drain the vessel, p; ovided the fuel pool gates are opes and the fuel pool is r.aintained above the lov level alar = point,
and provided one RF2SV pu=p and associated valves supplying the standby cola:t supply are cperable.
144
r A
4-t,TtCTI*fC CO'fDITIC:ss rom 0?mTIO!!
SURVEILt.ntCE BE' UIS2:CCT*-
).).A meeldunt Heat Removat Svetem 4.5.1 Restdual Weat Recovst Svaten (RHR$)_ (LPCI sad Cont ainment (RIOT 3] (L?CI and Containment Cooling)
-Coolin g) 1.
The RilR5 shall be operablet 1.
a.
Simulated onee/
Automatic OptratinA (1) prior to a reactor Actuation Cycle startup f reci r Cold Test Condition; or (2) when there is irra-b.
Furup opera-Cace/
disted fuel in the bility msnth reactor vessel and when the reactor vessel pres-c.
Motor Opera-Ouec/
sure is Areater than ted valve month atmospheric, except an operability specified in spect!1ca-tiene 3.5.3.2, throagh d.
Pucrp Tiov Rata once/3 3.5.R.7 monthe 2.
Vith the reactor veneel pres, e.
Test Check Valve Once/
sure less than 105 pair.. the operating C Cl*
Y etHR$ m a y h e r ea.ow ed f r osi s e r-vice (except that tun RnR pumps-centsinnent coolin;;. code and Each 1.PCI pump shall deliver 9000 agnociated heat c:tchangers reust gpc1 against an indicated s'fstem renain operabic) for a period pressure of 125 psig. Two LPCI numps not to exceed 24 hovre while in the same loop shall deliver being drained of suppression 15.000 gpm against an indicated chanher qualit7 vater and system pressure of 200 psig.
. filled with primary coolant quality water pr=vided that 2,
6 air test on t'ne d w il m during cooldom cuo locps v'_th torus heade s and no:: lea chall one pu.mp per loop or one loop vich h
- a c d m/3 Ws.
r.
two pumps, and associated diesel water test may be pd.:nd on generators, in the core spray systen the torus header in lieu of t'.u are operr.ble.
air test.
3, If one Ritu pumo (LFCI code) 3,
- g. hen it is datermined that one RHH is innperehle. the reactor O OE #
r ey remain in nperation 'or a time when operability is required, perted not to esceed 7 days the remaining RHR pumps (LPC1 mode) provided the rem ining RNR and active components in bnth 3ccess pucoa (LPCl mode) and both pathe of the RHR$ (LPCI r' ode) and neceen p e t tia of the RHR$
the CSS and the dit scl r.cncrators (t.FCI node and the CSS and the d ie sel) ce nt ra tors sh/ill be demonstrated to be opera-remain
- operable, ble 1 : mediately and daily thereafter.
145 Amendment No. 75
G PAGE DELETED 146
7 e-SURVEIIIN*dE Td'QUIREMINTS
~ 1,Ih1 TIC CCNDITIONS 70R OPERATION 3.5.T Reactor Cora Isolation Coolini 4.5.F Reactor Core Isolation Coo 11r.1 j
2.
When it is determined that the
-2.
If the RCICS is inoperable.
RCICS is inoperable, the FJCIS the reactor may remain in shall be demonstrated-to be operatien_for a period not operable i= mediately.
to exceed 7: days if the KTCIS is operable during such time.
3.
If. specifications 3.1.F.'
or 3.5.F.2 are not met, sn orderly shutdown shall be initiated and thu reactor shall be dupressurizecd to less than 122 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
G.
Automatic Dcoressurization
.C.
Automa t ic,_pqp res suriza tion Systes (A* S)_
J System (ADS) l.
Fo's; of the six valves of 1.
During each operating tycis the following tests shall be the Automatic Deprc9suri-zation System shall be perfor=ed en the ADS:
operable:
a.
A simulated aute: art:
acttiati:n test shal* Fu (1) prior to a startup f rom a Cold Condition, perfor ed prior to stare:::
or, after each refueling out-age. Manual surveillante of the relief valves is (2)' whenever there is irra-covered. in 4.6.3.2.
diated fuel in the reactor vessel and the tesetor vessel pressure is greater than 105 puig, e.< cept as speef fied in 3.5.G.2 and 2.
'When it is decernine taa: n se 3.5.c.3 below.
than two of the ADS val.es s;c 2.
If three of the six ADS valves incapable of aut:catt: :;e r.i t i m.
are known to be intapable of the ;il'CIS Shsil be der.or.stra r e.t autcmatic >perat!on.' the to b's opersbie 1:=adiaccly a:id dsfly thereafter as len; as reactor may remsin in opera-Specification 3.5.C.: applies, t ion f or.* period not to ex.:ced 7 days, provid?4 the lifCi system is operable.
(Note thtt the prenorc roltef functbin of these valvas is ast.ured bv actiott 3.6.0 nf these e,ceifications and rhat this spect 'ic..t ton anly at p ties to the Aim f unct len.)
If more than three of tne six A"S valves are knem te he incap-able n f aut on ic i.e operation, an itswitat e orderly P.hutdown shall br initiat.d. "irh the 157 rest: tar t o a he t sh"'aie..m con-dicion in 6 hourr* and in a cold shutdown condition in the follewing 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
1 l
knendmeht No.
59 1
FEB 2 51990 1
7
.E
~
LIMITI'NG CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS -3.5.0 Automacie nepressurization 4.5.C Automatic nepressurization System.
System 3.
If specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown will be initiated and the reactor vessel pressure shall be reduced to 105 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
11.
Maintenance of Filled D scharge PiDe H.
Maintenance of Filled J
Disch&rce Pice Whenever the core spray systems, LPCI, HPCI, or The following surveillance RCIC are required to be requirements shall be operable, the discharge adhered to assure that the piping from the pump discharge piping of the discharge of these systems core spray systems, LPCI, to the last block valve HPCI, and RCIC are filled:
shall be filled.
The suction of -the RCIC and HPCI pumps 1.
Every month prior to the tes:ing shall be al'igned to the condensate of the RHRS (LPCI and Con:ainnes:
storage tank, and the pressure suppres-Spray) and core spray system. :he sien chamber head tank shall normally discharge piping of these syste=s bc Aligned to serve the discharge piping shall be vented from the high poin:
of the RHR and CS pumps. The condensate and wa:er flow determined.
head tank eay be used to serve the RHR and CS discharge piping if the PSC head 2.
Following any period where the L?CI tank is unavailable. The pressure or core spray systems have not been indica: ors on the discharge of the RH:'.
required to be operable, :he dis-and CS pumps shall indicate not less char;:e piping of the inoperable sys-than listed belev.
tem shall be vented fro: :he high Pl-75-20 48 psig point prior to the return of :he Pl-75 43 43 psis, system to service.
Pl-74-51 48 psi:;
P1-74-65 48 psig 3.
Whenever the HPCI or RCIC sys:e= is lined up to take suction fro: the condensate storage tank, the dis-charge piping of :he HPCI and RCIC shall be vented from the high pein:
of the system and water flew observed on a conthly basis.
4 When the RHRS and the CSS are re-quired to be operable, the pressure indicators which monitor the dis-charge lines shall be conitored
. daily and the pressure recorded.
Amendment No. 76 158
1II l
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
~,$,1 Average Planar Linear Heat' Generation 3
4.5.I Maximum Average Planar Linear Heat Rato Generation Rate (MAPLHGR) j
During steady state power operation, the The MAPLHGR for each, type of fuel as a Maximum Average Planar Heat Generation function of average planar exposure Rate (MAPEGR).for each type of fuel as shall be determined daily during a function of average planar exposure reactor operation at 2125% rated shall not exceed the limiting value thermal power.
shown in Tables 3.5.I-l through 3.5.I-5.
If at any time during steady state operation it is determined by. normal.
surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 =inutes to restore operation to within the prescribed limits.
If the APLHGR is not returned to within the prescribed 1,imits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
J.
Linear Heat Ceneration Rate (LHGR)
J.
Linear Heat Generation Rate (LHGR)
During steady state power operation, the The LHGR for 8xS, 8x8R, and P8x8R fuel linear heat generation rate (LHGR) of shall be checked daily during reactor any rod in any fuel assembly at any operation at 225: rated thermal power.
axial location shall not' exceed 13.4 Kw/ft.
If at any time during steady state operat3on it is decernined by normal survefilance that the linitine value for LUCR is being exceeded, action shall be
'6'. s initiated within 15 minutes to restore operation to within the prescribed limits.
If the LHCR is not returned to within the prescribed.imits within two (2) neurs, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor oper-ation is within the prescribed limits.
Amendment No. 76 159
.W l
LIMIT!!!G CO::0ITIO!:S FOR OPIPATIO!!
SUPNEILLA::ct RCOUIRI":I::TS 13 5.K Mini um Critical Power Ratio U.S.K. Minimum Critical Power
-(MCPR)
Ratio (MCPR)
The minimum critical power ratio-(MCPR)
reactor power operation at>25% rated ficw,.shall _ be equal to or greater than thermal pcwer and fo11 cuing any shown in' Figure 3.5.K-1 multipiled by change in power level or distribution the Ff shown in Figure 3 5.2, where:
that would cause operation with a limiting control rod pattern as T = 0 cr. Tave~ 73
, whienever is described in the bases fc' 7 A-TS greater Specification 3.3
- 2. The MCPR limit shall be determined
-r"l A'=0.90 see (Specification 3 3.C.1 scram for each fuel type SX8, 8X33, P3XER, ti=e'li=it to 20% insectica frem from figure 3.5.K-1 respectively fully withdraOn) using:
y r
2.
I3=0.7 %1.65 '
N (0.053) /~Ref 5
- a. T= 0.0 price to initial scram
~
n time =easurements for the cycle,
~
erferred in accordance with e
specification 4.3.C.1.
7, ave: L: I
- b. Tas defined in specificatica n
n = number of surveillance red tests 3.5.K fc11owing the conclusien of performed to date in cycle (including each scram time surveillance test 30C test).
required by specifications 4.3.C.1 and u.3.C.2.
,,I i=
Scras time to 20% insertien frem
_ fully withdrawn of the ith red.
The determination of the limit must be completed within 72 hcurs of each scram time surveillance
!! =
total number of active reds ceasured required by specification 2.3 0.
,in specification 4 3.C i at 30C k
.If at any time during steady state operation it is determined by normal survellance that the limiting value for MCPR is being exceeded, acticn shall be initiated within 15 minutes te restore operation te within the prescribed
-l limits.
If the steady ~ state MCPR'is net returned to within the prescribed limits within two (2) hours, the reacter shall be brought tc the Cold Shutdown ccaditien within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, surveillance and correspending action,shall continue until reacter operatien is within the prescribed limits.
Amendment No. 76 160
w,
's Limiting Conditions for Coeration Surveillance Requirements 35 Core'and-containment cooling Systems 4.5 Core and containment Cooling Systems L. 'APRM Setooints' L.
APRM Setooints 1.
Whenever the core thermal FRP/CMFLPD shall-power is > 25% of rated, the be determined ration of FRP/CMFLPD shall be daily when the
> 1.0, or the APRM scram and -
reactor is > 25P. of
, red block setpoint ecuations rated thermal listed in sections 2.1.A and power.
2.1.3 shall be multiplied by FRP/CMFLPD as follows:
-S< (0.66W + 54%) FRP CMFLPD S 1(*
+
R3 PD 2.-
When it is determined, that 3.5.L.1 is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition.
3 If 3 5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to < 25% of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
M.
Recorting Requirements If any of the limiting values identified in Specifications 3.5.I, J, K,.or u.3 are ex-ceeded and the specified remedial actior.L is taken, the event shall be logged
.and reported in '. 30-day written report.
160A Amendment No. 76
at
- L t
8 G/
.1 jM5.JiTLinearHeatGenerationRate(LHGR)
This specification assures that'the linear heat generation rate'in any
.c roduis-less than 'the design linear heat generation if fuelipellet ydensifiertion is postulated.
The LHGR lforf 8xB, 8x3R,' an[ P8x8R-fuel 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, MT?F would have to be 1 greater than 10 which is precluded by a considerable =argin when
-employing any permissible control. rod pattern.
3 5.K Minimum Critical Power Ratio-(MCPR)_
-At: core thermal. power le'vels less than or equal'to 25%, the reactor will be. operating at1 mini =um recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns
-which may be employed at this point, operating plant experience and thermal hydraulic' analysis indicated that the resulting MCPR value is in excess of requirments by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in
.a more. conservative mode relative to MCPR. The daily requirement for
_ calculating MCPR above 25% rated thermal power is sufficient since power
~ distribution shifts.are very slow when there have not' been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod ' pattern is approached ensures that MCPR will be known following a' change'in power or power shape (regardless of magnitude) that could place operation at a thermal limit.
3.5.L APRM Seteeints The. fuel cl' adding integrity safety limits of section 2.1 were based on a
- total peaking factor within de' sign limits (FRP/CMPLPD 21.0).
The APRM instruments must be adjusted to ensure that the core thermal limits are not exceeded in a degraded situation when entry conditions are less conservative than design assumptions.
3 5.M Reoortinz - Recuirements The LCO's. associated with monitoring the fuel rod operating conditions are required to be met at all times, i.e., there is no allowable time in which the' plant.can knowingly exceed the limiting values for MAPLEGR, LHGR, and MCPR.. It is a requirement, as stated in 9pecification 3 5.I, J,1cnd K, that if at any time during steady state power operation it is determined that the limiting values for MAPLEGR, LHGR, or MCPR are exceeded,. action is then initiated to restore-Operation to within the prescrabed limits..This action is initiated as soon as normal surveillance indicates ~that an operating limit has been reached. Each
-event involving steady state operation beyond a specified limit shall be
~
reported within 30_ days. It must be recognized that there is always an action which would return any of the parameters (MAPLHGR, LMGR, er MC?R)
=to within prescribed limits, namely power reduction. Under most
~
circumstances,. this will not be the only alternative.
169
- Amendment--No. 76-
4.$
Core and Containment Cooling Svetems Sutvet11a c Frecuencies The testing interval for the core and containment 4 s11nr systems is ba sed on industry practice, quantitative reliability analysis, judgement and practicality. The core cooling systems have not been designed to be fully testabic during operatic.
For example, in the case of the HPCI automatic initiation durtnf power operation would result in pumping cold water into j
the reactor vessel which is not desirable. Co=plete ADS testing during power operation causes an undesirable losa-of-coolant inv en to ry.
To increase the availability of the core and containment coolin5 System, the components which make up the system; 1.e.,
instrumentation, pu,ps, valves, etc., are tested frequently. The pumps and motor operated injection valves are als) tested each nonth to assu:e their operability.
A sinulated aute-atic actua-each cycle conbined with monthly tests of the pumps and injec-tien test once tion valves is deemed to be adequate testing of these syste,s.
When components and subsystems are out-of-service, overall core and contain-ment coolinr. rettah111ty is maintained by demonstrating the operability of the remaining equipment. The der.ree of cretacility to be demonstrated depends on the nature of the reason f or the out-o f-se rvice equip *ent.
For routine out-of-service periods caused by preventative maintenance, etc., the puma and valve operability checke vill be perf ormed to der.onstrate operability of 15e remaining compenents. HovcVer, if a failure, design deficiency, cause the out4be, then the demonstration of operability should he thorough enough to assure that a generic problem does not exist.
For example, if an out-of-service period van caused by failure of a pump to deliver rated capacity due to a design deficiency, the other punps of this type night be subjected to a fl9v rate test in addition to the operability checks.
Whenever 4-CSCS system or loop is made inoperable because of - required teac or calibration, the other CSCS systems or loops that are required to be operable ghall be considered operable if they are within the required surveil-lance testing frequency and there is no reason to suspect they are inoperable.
If the function, synten, or loop under test or calibration is found inoperable or exceeds the trip level setting, the LCO and the required surveillance testing for the system or loop shall apply.
Redundant operable components are subjected to increa:cd testing during equip-ment out-of-aervice times. This adds further conservatism and increases assurance that adequate cooling is available should the need arise.
Maximus Average Planar LHCR, IRCR, and MC?R N PA?iXCR, l.HCR, and MC?R shall be checked daily to det ermine if fuel buraup, or control rod =ovement has caused changes in power distr'bution. Since changer due to bursup are slew, and only a f ew control rods are moved daily, a daily check of power distributies is adequate.
170
'q5 I
f~
4
~
f:
4 b
t-n '
- -l) 3 5.N': References I
I-:
L
- 1.. -" Fuel Densification Effects on General Electric Boiling Water Reactor.
' Fuel," Supplements 6, 7, and 8, NEIM-10735, August 1973 it.
- 2. '.Suplement 1.' to Technical Report -on Densification of General Electric -
F Reactor Fuels, December 14,' 1974 (USA Regulatory Staff).
E
'3 1 communication:
V. A. Moore to I.S. Mitchell, " Modified GE Model for
.' Fuel.Densification," Docket 50-321, March 27, 1974 t-4.1 Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.
1 5.
Letter from R. H. Suchhol:: (GE) to P. S. Check (NRC), " Response to NRC request for information on ODYN computer model," September 5, 1980.'
Mendment No. 76 169A
D
'[4 Tchla 3.5.I-1 MAPLilGR VERSUS AVERACE PLANAR EXPOSURE Fuel Type: ADB274L
' Average Planar Exposure MAptscg i
(Mud /c)
(k'J/f t) 200 11.2 1,000 11.3 5,000 11.9 10,000 12.1 15,000 12.2 20,000 12.1 25,000 11.6 30,000 10.9 55,000 9.9 40,000 9.3 Table 3.5.I-2 MAPLHGR VERSUS AVERAGE PLANAR EXPOSLTE Fuel Type: SDB274H Average Planar Exposure MA?L3GR (Mwd /:)
(44/8:)
200 11.1 1,000 11.2 5,000 11.3 10,000 12.1 15,000 12.2 20,000 12.0 25,000 11.5 30,000 10.9 35,000 10.0 9'3 40,000 171 Amendment No. 76
i Tcblo 3.5.I-3 MAPLHGR VERSUS-AVERAGE PLANAR EXPOSURE.
Fuel Type: 8DRB265H Average Planar Exposure MAPLHGR
-(mwd /e)
(kW/ft).
200 11.5 1,000
'11.6 5,000' 11.9 10,000 12.1 15,000
.12.1 20,000 11.9 25,000 11.3
-30,000 10.7 35,000 10.2 40,000 9.6 Table 3.5.I-4 i'
MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 8DRB265L and P8DRB265L Averaga Planar Exposure MAPLHG".
(mwd /t)
(kW/f t) 200 11.6 1,000 11.6 5,000 12.1 10,000 12.1 15,000 12.1 20,000 11.9 25,000 11.3 30,000 10.7 35,000 10.2 40,000 A6 i+
170 Amendment No. 76
4W +.,.
/[
.Tcblo 3.5.I-5 MAPL $CR VERSUS AVERACE PLANAR EXPOSURE R
Fuel Type: P8DRB284L, J
--GLTA-1, GLTA-2 Exposure-
' MAPLHCR~
-(mwd /t)
(kW/ft)
'200
-11.2 1000-11.3 2
5000 11.8
- n
'10,000
-12.0
-15,000-12.0 20,000 ^.
11,8 25,000 11.2
-30,000.
10.8 35,000 10.2 40,000 9.5
'172a-Amendment tio.'76 r
L
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0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 mL
..igure 1..i.s-1 r
MCPR I.ZlTS*
aNOTF.:
- 1. cad test ass ablics are categor iud m P3xfiR bundl es.
Amenument No. 76 17,.b e
e G
LIMITINC CCNDITIONS FOR OPERATION SURVEILI.ANCE REQUIREMENTS 3.6.C Coolant t.c e k a n 4.6.C Coeisnt t.eakate 3.
If the condition in 1 or 2 above stanot be set, an orderly shutde>n shall be initiated and t% reactor shall be shut-D. Relief Valves down in the Cold Condition withie. 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />.
1.
At least one saf ety valve and approximately one-half of all D.
Relief Val res relief valves shall be bench-checked or replaced with a J. When more than one relief be.nch-checked valve each opera-valve or one or : ore safety ting cycle. All 13 valves (2 valves are known to be safety and 11 relief) vill have failed, an orderly shutdown been cheeked or esplaced u;en aW be initiated and the de concletion of enry second reactor depressurized to cycle.
less than 105 peig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
Once.:tri: g each epers:in; cycle, each relief valve-shall be nanually epened until thernecouples and acoustic nonitors det.T.strean of the valve indica:e steam is flowing frem the valve.
3.
The inte:rity of the relie!!
safety valve bellows shall be continueusly =enitored.
4 At least one relief valve shall be disass:: bled and !=67e=:44 cach operatin;; cycle.
E.
Jet Punes
- y..
.T e t P;:es 1.
Vhenever the reactor is in the 1.
V5eaever there is retir: la:i n startur or run. odes, all jet flow vi:5 :he rea: or in :he pumps shall be operable. If startup or un =edes with be h it is detemined that a jet retirculati:r. pu ps ru.ning, pump is inopetsele, or if two je pu sp operabili:y shall be or more jet o west flow instru-checked' daily by verifying that ment failures occur and can-the followtst coadi:iens to.o:
not be corrected within 11 occur sicsl:aneously:
hours. an orderly shutdo m shall be initiated and the a.
The vo retircul:: ion loess reactor shall be shutdown in have a fiev i=.talance of the Cold Candi:fon within 24 13I or tore when the ;utps hours.
are operated 4: the sase speed.
181 Amendment No. yI, 71
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 4.g.g Jet Pu s,.
3.6.E Jet Pumos 3.6.P Recirculation Pump Operation glow rat. werics !rw the value Arivsd f row loop fiev messarements by more-than 101.
The dif f ue<r to 1o'rer virsuna c.
dif f erestial pre saure read-ing on an individual jet pump varies Iree the mesa of all jet pu:rp dif f eren-tial pressuras by amore than 101.
2.
Whenever there is recirculation II'" "I * *
- 1. The reactor shall not be I t ' " " E ' ' E'#" "* d ' ' " d ' " ' T '~
operated with one recirculation cireala !on pump 1) opunting 1o00 out of service for more "ith th' '9"'11**f 1 'I
than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With the reactor O' C h*"
- PI'"""
operating, if one recirculation diffff'"C1'l 7f***"f' ' hall be loop is out of service, the checked daily and the differen-plant shall be placed in a hot tial wressure of an individual shutdown condition within jet pop in a loop shall not 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is vary f rem the seen of all jet sooner returned to service, pu=p dif f erential pressures in that loop by morr than 101.
3, Tolled.r.g one pt=p ep-stien, the discharEe valve of the low epeed pts.p =ay r.ct be epened 4.6.F Recirculation Pumo Operacion urless the speed of the ft:ter
. Recircul tion pump speeds shall 1.
p.:=p is les s tha.n Scf. of its be checked.m.nd logged at least rated :yeed.
once per day.
3.
Steady state operation with both recirculation purcs out of ser-vice for up to 12 tirs is per-ettted. During such interval restart of the recirculation pros is pemitted, provided the loop discharge temperature is within 75CT of tne saturation temocrature of the reactor vessel water as determined by dorne pressure, c.
screteural intrert:v 1.
Table c.6.A tosether vi:5 sup-C.
Structural Inteerity pleme nt a r7 not es, speciften sne 1.
The structural integrity of the Prirury systes shall be 1 82 Amendnent No. 75
e v
3,6/4.6
.BA % ES :
The basis for the equilibrium coolant iodine activity limit is a coeputed dose to the thyroid of 36 rem.it the exclusion distance during the 2-hour period follouing a steam line bresk. This cose is comput'ed wit 5 the conservative assumption of a release of 140,000 lbs of coolant prior to closure of the isolation valves, and a X/Q value of 3.4 x 10*' Sec/m3 The maximum activity limit during a short term transient is established from consideratinn of a maximum iodine inhalation dose less than 300 rem.
The probability of a stean. line break accident coincident with an iodine concen: ration transient is s:gnificantly lower than that of the accident alone, since operation of the reactor with iodine levels above the equilibrium value is limited to 5 percent of total operation.
The sar pling f requenc ies arc establi shed in order to detect the occurrence of an iod.ne transient snach r..ay ex;ced the eqviltbrium concentration limit, and to assure t....; the maximum coolant lodine concentrations are not excecced. Additional sampling is required f ollowing power changes.nd of f-gas transient s, since present data indicate that the iodine peaking phenomenon is related to these e'ents.
3.6.C/4.fi.C Coolant i.eakare A11ovahic leckace rates of coolan frem the reactor coolan: system have been based on the predicted and experimentally observed behavior of cracks in picco and on the ability to makeup coolant system 1cakage in the event of lona of offette a-c power. The normally expected background leakage due to equipment (?sich and the detection capability f or determining coolant sys-tem leakate were also considered in eccablishinn the limits. The behavior of cracks ir, pipine. nyacena has been experittentally and analytically inves-titated as part of the tfSAEC opensored Reactor Pri-ary Coolant Sys:cm Rupturc Study (t he Pipe Rupture Study). Verk util!:ing the data obesined in this Jtudy indicates that leskate from a crack can he detected before the ers'ck nrows to n danc.erous or critical size by meenanically or :hernally induced cyclic lo.tding, or stress corrosion crackinr. or some other mechanism characterf ted by gradual crcck growth. This evidence sugtests that for leak-age somewhat greater than the limit specified for unidentified leakage, the probsbility is small that imperfections or cracks associated with such leak-age would grow rcpidly, tiewever, the establishment of allowable unidentified leakage greater than that given in 3.6.C on the basis of the data presently ava!!able veuld be premsture because of uncertainties associated vi:h the data.
For'leakete of the order of $ gpm, as specified in 3.6.C. the experi-mental and analyttcal dets sugr.ent a reasonable carr.in of safety that such loakane magnitude would not result from a crack approaching the critical site for rapid propagation. Leakage less than the magnitude specified can be 218
c 3.6/4.6 BAS ES detected reasonably in a matter of few hours utilizing the available leakage-detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down 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 sumps.
The capacity of the drywell floor sump pump is 50 gpm and
'the capacity of the drywell equipment sump pump is also 50 gpm.
Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
REFERENCES 1.
Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10) 3.6.D/4.6.D Relief Valves To meet the safety basis thirteen relief valves have been installed on the. unit with a total capacity of 83.9% of nuclear boiler rated steam flow.
The analysis of the worst overpressure transient, (3-second closure of all main steam-line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate-margin to the code allowable overpressure limit of 1375 psig.
t To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open).shows that 12 of the 13 relief
. valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1375 psig, i
219 Amendment-No. 76
3.6/4.6 s AsEs:
trperience in relief valve operation shows that a testins af 30 percent of :he valves per year is adequate :s detect failures or deteriora:Lons. D e relief valves are benchtested every second operating cycle to ensure : hat their set points are within the
- 1 percent tolerance. De relief valves are tested in place once per opera:ing :ycle to establish that they will open and pass steam.
De requ:remen:s es:ablished above apply when the nuclear tys:am :a= be pressurized above ambient conditions. De s e r e qui re.r.en ts are amplicable nuclear system pressures below normal epera:ing pressures because atabnor:.s1 opera:ional transier.ts :culd possibly s:ar: a: :hese erndi: ions such that eventual overpressure relief would be needed. However, these tranalents are auch less severe, in terms of pressure, than those a:nreing at rated conditions. De valves need not be functional when the vessel head is re.oved, since the nucisar system cannot be pressurized.
RITnZNCES 1.
Nuclear Systes Pressure Islief Systes (3TNF TSAA Suosection 4.4) 2.
Amendment 12 in response to Aic Question 4.2 of December 6,1971.
3.
" Protection Agains: Overpressure" (ASMT Soiler and Fressure Vessel Codt, Section !!!, Article 9) 4.
3rovn Terry Nuclear Flar.t Design Deficisocy P.s;er:--Terge: Roch Safety-Relief Vilves, transmi::sd by J. E. Otiteland to T. E. K:assi, Au gu s t 29. 1973.
3-Gene:10 Reload Tuel Applicati:n, 1.1:ensing !cpical Report, ICE-N011-?-A and Addenda.
- 3. 6. t./4. 5. E Je : Nms Tsilure o! a jet puap nos:le asserely holddovn r.eehasism. noetle sarsmbly and/or riser, would increase the cros s-sectional !!ov area for blevt.svn folleving :he destgr* basis double-ended line bras's.
Also, failure =f :he dif fuser veuld siidnase the capability to te!!aod the core to two.:hteds heigh: level it.;11ovtag a recir:.alation line breas. Berefors, if a failure occurred. repairs must be made.
- ith the evo recircula: ion pumos Be det ec:.on t echnique is as follows.
=
balanced 1.1 speed to wi:hin ; $ percent, :he flow rates in bo:n recir:ula-tion loops vill be veri fied by control room sonitoring instru=ents.
If the evo (Icv rata values do soc di!!ct sy more :han 10 per:ene, riser and no::le essembly inagri:y has been viri'ied.
220
/
Amenc ent No. j#, 59
- C..- m....
e
~
a d
3.(./4.4 nA,5fft if they de dif fer by 10 percent or cose, the core flow rete measured by the jet' pump diffumer differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or eiere (with the derived value higher) diffuser measuremente vitt be taken to define the location within the vessel of fatted, jet pump nosate (or riser) and the unit shat down for repaire. If the potential l
blowdown flew sees is increased, the system resistance to the recirculation puas to ateo reduced;.hence, the ef fected drive pump vill "run out" to a euhatantially higher flow rate (approximately 115 percent to 120 percent r
for e nint.le nostle failure). If the two loops are halenced in flev at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant reocese instrumentation. In addition, the effected jet puep would provide a leakar,e pa th past the core thus reducing the core flow rate. The reverse flow thtour,h the inactive jet pump would still be indicated by a positive dif ferential pressure but the net effect would'be a slight decrease (3 per-cent to 6 percent) in the total core flow measured. This decrease, together with the loop flow increase, would result in a lack of corre.'ation between mea.nrcd and derived core flow r' ate.
Finally, the affected jet pump diffuser dif f erential pressure signal would be reduced because the backflow would be less than the normal forward flow.
A nantle-rt=cr system failure could also tenerate tha coincident failure of let pu.o diffuser body: however, the converse le not true. The lack of a
any substantial stress in the jet pump diffuser body makes failure impossible without an initial nottle-riser system failure.
I b6,r/4.6.P Recircula tion Ptmn Operation Ste.nly state operation without forced recirculation will not he per tit ted for r ove th m 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
And the start of a recirculation pump from the nat' ural circulation condition vill not be permitted unicss the temperature dif f creuce between the loop to be started and the core coolant temperature i
is less than 75 T.
This reduces the positive reactivity inscetion to an l
acceptably lov value.
l l
221 Amendment No. 76
7 m.
-o p_
).6/3.6 g.A,g.gg 9,
V F
' These tests will inclut.e stroking of the snubbers to ve:Lf y.
l proper pL s ton movement, lock-up and bleed.
Ten percent or ten p
snubb<a r s whichever is less, represents an adequate sample for r
P such tests.
Observed-f ailures on these samples should require testing of ' additional _ units.
Those snubbers designated in Table 3.6.H as ba.ing in high radiation a reas or especially dif ficult to t-remove need not b.e selected for functional. tests provided operability was previcusly verified.
Snubbers of rated capacity greater than 50,000 lb. are exem::t frem the functional. testing.' requirements because of the inpracticability of testing such large snits.
REFERESCES 1.
Re po rt,
H.
R.
EricXsen, Bergen Paterson to K.
R. Goller, trRC, October 7, 197:1, subject:
Hydraulic Shock Sway Arrestors 225 u._
9 LIMITING CONDIT CNS TCR OPERATICN SURVE LLANCE REQ U t RI.INTS I
- 3. 7 : q,Q r,qa f gj.1*J 3'f gj to. 7 CONT A I NM ENT S YST!*85
. yA}J,g3bility Aeolicability Applies to the operating status Applies to the primary and of the primary and secondary secondary containment containment systens.
integrtty.
Ceiectiva obwet t v_e
-To assure the. Integt
- .y of the peamary and secondary To verif y the Integrity of the containment systems, primary and secondary containmenc.
!.EtsAfif AtloJ.1 seeetficatten A.
Primary containm nt A.
At any time that the irradiated fuel is in 1.
Pressure sueeressien the reactor vessel, chameer and the nuclear systes is pressurited a.
The supp ession above atmospheric charter water level pressure or work is being done which has be checked once pe,-
the potentiel to day. Whenever heat drain the vessel, the is added to the pressure suppression sucoression Dool by pool water level and testing of the ECCS tenperature sna11 :e or relief valves the maintained within tne pool temoerature shall fellcwing limits be c:ntinually monitored except as specified arid shall be observed in 3.7.A.:.
and logged every 5 minutes until the heat a.
Minimu:n vacer level =
addition is teminated.
-6.25" (differential pressure control
>0 psid)
-7.25" (0 psid differen-tial pressure centrol' b.
Maximum vater level =
-1" 227 Anendment No. 76
l LIMITING CCNDITICNS FCR OPERATICN SURVEILI.ANCE REQUI?J.MENTS 3.7 COWAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS 6.
Dryvell-Suppression Chaaber 6.
Dryvell-Suppression Cha=ber Differential Pressure Differential Pressure a.
Differential pressure a.
The pressure diff er-between the dryvell and ential between the soppression chamber shall dryvell and suppression be maintained a: equal chamber shall be; recorded to or greater thsn 1.1 at le2st once each shift, psid except as specified in (1) and (2) below:
(1) This differencial shall be established within 24 heurs of achieving operating te=perature and a
pressure. The differential pressure
=ay be reduced :o less than 1.1 psid 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prio :o a scheduled shutdown.
(2) This differential
=ay be, decreased :o less than 1.1 psid-for a =ax1=um of Ecur hours during required cperability testing of the U.?CI syste=,
RCIC systes and the dryvell-pressure suppression cha=ber vacuum breakers, b.
If the differential pressure of specifica-tien 3.7.A.6.a cannot be
=ain:ained and :he differential pressure cannot be restored within the subsequen: six (6) hour period, an orderly shutdown shall be ini:-
iated and the reactor shall be in the Cold Shu:doim condition v1 thin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 733, i
/cendment No. 76 j
2 li
'o y
LIMITINC CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
~
3.7' CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS H.
Containment Atmosphere H.
Containment Atmosohere Monito ring (CAM) System -
Monitoring (CAM) System -
.H Analyzer H Analyzer 2
4 1.
Whenever the reactor is 1.
Each hydrogen analy:er cot in cold shutdown, two system shall be demon-e.deaandent gas analyzer strated OPERABLE at systems shall'be operable least once per quarter for monitoring the drywell.
by performing a CHANNEL and the torus.
CALIERATION using standard gas samples containing 2.- With one hydrogen analyzer a nominal eight volu=e inoperable, restore at percent hydrogen balance least two hydrogen nitrogen.
analyzers to OPERABLE status within 30 days or 2.
Each hydrogen analyzer be in 'at least HOT system shall be demonstrated SHITrDOWN within the next OPERA 3LE by performing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
a CHANNEL FUNCTIONAL TEST monthly.
3.
With no hydrogen analyzer OPERABLE the reactor.
shall be in HOT SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
249
- Amendment No. 76
oi
.o 1 TA3LE 3.7.A PRIMARY C3tfTAINNEIT ISOLATION VALVES Nar2 tier of Power Maxtaun Action on operated Valves _
Operating.
Normal Initiating CrH Valve Identification Inbc.a rd Ou tiso a r d Tine (sec.)
Position S1Roel 1
'tato steamline isoletton valves 4
4 3<T<5 0
CC (FCV 14,26,3 7,657 jl-15, 27, 38, & 52) 1 Main stearaline drain isolation 1
1 15 C
SC valves FCV-1-55 & 1-56 1
Reactor W ter sample line isole-1 1
5 C
SC --
tioin valves
,f
- a 2
kilRS uliutdown cooling supply s
isolation valves < FCV-74-48 6 47 1
1 40 C
SC o
2 RilRS - 1.PCI to reactor FCV-74-53, 67 2
30 C
-SC 2
React on' v'i.ssel licad spray isola-tion.v'21ves FC'v-/4-7 7, 7tl 1
1 30 C
SC 2
HitRS f luali and drain vent to suppreauton cluralier.
4 20 C
SC l'C V-74-102, 103, 119, & 120 2
Su preaston Cliauber Urdt'n 2
15'
C SC FCV-/4-51, 58J
/
2 Drywall equipa.ent drain 'disclurge
,15 0
GC lisolstion valves ycy-77-t SA, s 15g 2
e 2
Urysell (toor Jrata directurgo tuolation' valves FCV-1/-2A & 20 2
15 0
E I
.m
\\
~
Z%f[ ' {W " ',.
v y
m.
+ _, ~
y g
.y x
a
- 3,
- n -
L ',
+
-, j~-
q,; *p
,(.
8 4;
- ^ ^' *' ; /
y g n. ><'
,n
^~
y Q
..g TAtt.E 3.7.A (Continued)
>c e..x. -
j 19 a
Action es g.
Maxianme#
a, Mucher of Power Operated Valves _
' Operating
' Normal Initiating s s
2 Inboard Outboard Time (sec.)
Posittaa. jMit ' h;.
- rt
, A. Q. W0 Velve-lJentifiestion
-[Q-D 'M'
=-
Grous u j' m '
'n CC A s
Reactor water cleinup syntes suppl.r 30 O
3 1
L.,.
isolatton valves FCV-69-1, & 2 s.
~-
,~ m
-x GC
- 0' 3
Reactor water cicanup sy stem 60~
.y
-1
- ~
return 13alation valves ECV-69.,
. GC
- 1 ~/
FCV73-81(EypakaroundFCV 73-3) i
^ 10 '
o'.
CC 4
w 1
1 20 0l "p;
~,,s.
-w 4
urCIS steamitae tr,olation valves I'CV-73-2 & 3 1
,l'-
15
_. x0?-
~CC 5
RCICS etca. aline isolation valves FCV-71-2 6 3 u
La 6
Dsyvell nitrogen purge 1
5 C
SC inlet 1sals-tion valves (FCV-76-13) 1 5
C SC 6
Su),pression char.b tr. ni trogen purge inlet 1solattan valves (ICV-76-19) y
~.
~
w SC 6
Daywell Matu E=haust isolation 2
2,5 s (,
velves (FCV-6%"I4. sad 10) u x
s 3C 6
Suppraeuton che=Ier p in exhaust c
2 2.5 LC _-
isolation v 14=a (r&-64-32 ed 31)
._o C
SC 6
Dr)well/ Suppression Chaeler purga 1
2.5 7,
inlet (icv-64-17) s y
3C inlet C
crywell atu.saphere p.
ge 1
2.5 QCV-44-14)
J S
f hvu.f h
'a~ & f o g. f f-l- ~a
,]
g f
y#, 7 9.,_;~g y ;.,;_.
wy
+
~ ~
~
,; ): -
gj
+
c.'M
. : n
'A,
' M-y 5
~
g,; -.
p',
w
-TABLE 3.7.A (Continusd)-
+ ' +
4f,ps e'
,.: m; -.
.ht e 1 SC
'~
e 3
./
- g.,
,7
'W,,,,,,'r s
6 litywell 04ygen Sample Line Esives j
3
. Analyzer A,(FSV-76-51, S2) i I
' !;A.
Note 1 SC
~
s 6
Sample Return Valves'- Analyzer A (FSV-76-57, 58) 1 I
t:A -
O' 3 SC o
v.
E
~ Tords liydrogen Sample Line Valves 6.,
a
!G
. Note 1 SC Analyzer n (PS"-76-65, 66) 1 s
6 Turus Oxygens Sample 1.ine Valves-Analyzer B (FSV,
b i, 64)
I 1
NA' Note 1 SC' 6
Drywell liydrogen Sample Lit e Valves-Analyzer h (FSV-16-59, 60)
I I
NA
!!u t e ': t SC 6
Drywe!! Oxygen Sample 1.ine Valves-Analyzer n (FSV 61, 62) i NA Note !
SC 6.
Sample Rettern Valves-Anel>zer B (FSV-76-67, 68) 1 I
!;A 0
SC Note I:
Analyzers are sucla Eliat onie is sampling da ywell laydrogen and oxygen (valves f rom drywall open-valves from turus-closed) wii!!c the ottier is sampling torus hydrogen and oxygen (valves f rom (Orus open - valves from drywell closed) l
g.
4
.e W
. g-ag TABLE 3.7.A (Continued) e, Nunber of Power Maximum _
Action on P
Operated Valves Operating Normal Initiattrg Group Valve identtfication
'nboard Outboard Time (sec.) Position Stenal y
6 Suppression Chanter purge inlet (FCV-64-19) 1 2.5 C
SC 6
Drywell/Supprtssion Chanber nitro-gen purge inlet (FCV-76-17) 1 5
C SC 6
Drywell Exhaust Valve Bypass to Standby Gas Tieatinent System (FCV-64-31) 1 s
C SC 6
Stppression Chanter Exhaust Valve Bypass to Standby Gas Treatment U
System (FCV-64-34) 1 5
C SC 6
Drywell/ Suppression Chamber Nitrogen Purge Inlet (FCV-76-24) 1 5
C SC 7
RCIC Steamitne Drain (FCV-71-6A, 68) 2 5
0 GC 7
RCIC Condensate Punp Drain (FCV-71-7A, 78) 2 5
0 GC 7
HPCI Hotwell punp discharge isola-tion valves (FCV-73-17A, 17B) 2 S
C SC d
7 HPCI steamline drain (FCV-75-57. SB) 2 5
0 GC 8
TIP Guide Tubes (5)
I per guide NA C
GC tube
= _ _ _ _ _
~.
'Ib T
i C
TABIE 3.7.A (Costinued) i Acties on-Number of Pov:r Maxime
_ Operated Valves Operattag-Normal 1mittattag Position Stamal Cruut Valve. Identific.ition InbcArd_
DJ*.bdard
[$te [*cC.),
. Staudby itsguld control system
' theck valves CV 63-526 & 525
'1 1
NA C
Process.
2 NA 0
Process CV-3-558, 572, M4. & 568 Control ro.! hydraulic return 1
1 NA 0-Process i
clieck valves CV-85-576 & 573 All25 - 1.I'CI to teactor check 2
MA C
Process valveu cv-74-54 & 68 Ye w
3 4
J:
O s
-v
-e
'2 0
+
i.
TABLE 3.7.D;(Continued)~
['.,
..' Identification Medium Method Vel *. e
. Tost Test
'4 ster (2)
Applied between 69-1,69-500 and tu-505
- ^
69 2 RWCU: Supply Water (2)
Appi teit betwe.n 6'3-2,69-500 and 10-505 71 PCIC Steam Supply Air (1)
Applied between ?l-2 and.71-3 73-81 HPCI Steam supply.sypass Air (1)
Applied between 73-2 and 73-3 71 -3.
.RCIC.Sceam' Supply-Air (1)
Applied between 71-2 cnd 71-3 71 -RCIC Pump. Discharge Water (2)
Apolied between 3-66. 3-568,69-579. 71-39, and 85-576 2.
HPCI Steam Supply Air (1)
Applied between 73-2 and 73-3 73 HPCI Stdsm Su[jly ALr (1)
Applied between 71-2 Jnd 73-3 73 44' HPCI Pump Discharge' Water (2)
Applied between 3-67 3-554, and 71-44 74-47
- R 2hutdown suction Water (2)
Applied between 74-47.74-754 74-49. md 74-661 74-48
' RHR Shutdown Suction War -19 Suppression Chmeber Purge Nitrogen Applied betwen 76-17, 76-18, 76-19 Inlet 16-24 Dryvell/ Suppression Chamber Air (l)
Applied batwess 64-17, 64-18, 64-19, Nitrogen Furge Inlet and 76-24 17 2A Dryve11 Floor Drain Sump Wter( )
Applied between 77-2.A and 77-25 77-25 Dryw11 Floor Drain Susp Water (2)
Applied berveen 77-2A and 77-23 77-15A Dryvell Equipment Drain Sump Water (2)
Applied betvesa 77-15A and 77-153 77-153 Dryvell Equipment Drain Sump W ter U Applied berveen 77-15A and 77-155 C1}
to-254A Eadiation Monitor Suction A.ir Applie j betwas 90-254A, 90-2543, and and 90-255
.0-2545 Eadiation Monitor Suction Air $ I Applied betwen 90-254A, 90-2545, and 90-235
-255 tadiation tianitar Suction Air (2)
Applied berveen 90-25AA, 90-2543, and 90-225 261
1 f.
TABLE 3.7.D (Continued)
Valve Test Test
' Valves Identification _
Medium Method 76-49.
Containment Inerting Air Applied between inboard block valve and 76-49.
76-50' Containment Inerting Air Applied between inboard block valve and 76-50.
L 51 Containment Inerting Air Applied between inboard block valve and 76-51.
76-52
.. Containment Inerting Air Applied between inboard block valve and 76-52.
'76-53 Containment Inerting Air Applied between inboard block valve and 76-53.
76-L4 Containment Inerting Air Applied between inboard block valve and 76-54.
76-55 containment Inerting Air Applied between inboard block valve and 76-55.
76-56 Containment Inerting Air Applied between inboard block-l valve and 76-56, 76-57 Containment Inerting Air
. Applied between inboard block valve and 76-57.
76 50 Conrainment Inerting Air Applied between inboard block valve and 76-58.
76-59 Con:ainment -Inereing -
Air Applied between inboard block.
valve and 76-59.
76-60 Containment Inerting Air Applied between inboard block valve and 76-60.
76-61 Containment Inerting Air Applied between inboard block valve and 76-61.
76-62 Containment Inerting Air Applied between inboard block valve and 76-62.
76-ni Cantainment Inerting Air Applied between inboard block valve and 76-63.
76-64 Centainment Inerting Air Applied between inboard block valve and 76-64.
76-65 Containmer.c Inerting Air Applied between inboard block valve and 76-65.
76-66 Containment Inerting Air Applied between inboard block valve and 76-66.
76-67 Containment Inerting Air Applied between inboard block valve and 76-67.
76-68 Containment Inerting Air Applied between inboard block valve and 76-68.
261a
TAst2 3.7.D (Continued)
Valee Test Test
,velves tdentificatios_
Hedi.
Mathed_
90-237A a ndiation Monitor DischarSe Air Applied betvoes 90-2371 and 90-2573 90-2373 Radiation Meetter Dietharge Air Applied bemos W37A W MM 84-8A Containment Atmospheric Dilution Air Applied betvesa $4-8A and 84-600 84-88 Containment Atmospherie Dilution Air Applied between $4-83 and 84-401 84-at Containment Atmospheric Dilution Air Applied betvoes $4-3C and $4-403 84-8D Containment Atmospheric Dilution Air Appliad between 84-4D and 54-402 84-19 Containment Atmospheric Dilution Air Applied betwees 64-32 _64-33, 64-29, 64-30, and 84-19 (1) Air / nitrogen test to be displatenent flow.
(2) Vater test to be_ injection. lors.or downstraa:a collection.
Valve Test Test V31res id entifies t ion
!!ediun Method
$4-20 M.ain Exhaust to Standby Cas Treatment AirC1)
Appited bat nen 14-20,64-141, Nitronen(l)64-140, and 64-31 Applied bet even 34-4A-and 3 84-600 Main Exhaust to Standby Cas Treatment.84-601 ff sin F.th.iun t to Standby Cas Trest. ment Nitroa.cn Appliud 'actueen 34-43.:nd 34-601 84-402 t'a tn Exhaunt to Standby C49 Treatment
- !L c ena,en Applied between 54-ac aius34-603
- M-601 Ma n n 1:xhaust to Standby Can Treatment Mittoaen Applied betnen 84-40 and $4-602 64-14 nryvell t*ressurization Comp. 3ypass Air (1)
Applled between 64-141,64-140, 64-30, and 84-20
}
64-140 Dryvull Pressurization, Camp. Disc.
Air Applied between 64-141.64-140, 64-31, and 34-20
)64-139 i)ryvell Pressurization, Comp. Suction Air Applied between 64-139,64-141, and 64-34 1), Air / nitrogen test to be displacement flow (2) Water test to be injection losa or downstream collection.
62 Amendment NO. 75
\\
TABL E 3. 7. E SUFFRE55 ION CHAMBEA INFLUENT LItiES STOF-CllECK CLOSE ISOLATICN VALVES Valve Tcat Test Valves Identification Medius Method 71-14 RCIC Turbine Exhaust Water Apply between 71-14 and 71.'
.71-32 RCIC Vacuum pu:ap Discharge Water Apply between 71-32 and 71-$92
(
73-23 MFCI Turbine Erhaust Water Apply between 73-23 and 73-603 73-24 ItPCI Turbine Exhaust Drain Water Apply baeveen 73-24 and 73-609 TA3LE 3.7.F C!lECK VALVES CN SUPPRESSICN CI! AMBER IliTLUE.VI LINES Valve Test Test valvec Identifiestion Medius Mathed 71-580 RCIC Tu-5ine Exhaust Vater Apply betveen 71-14 and 71-530
/1-59:
nCIC Vacuu.m ?u=p Diocharge Vater Apply between 71-32 Sud 71-$92 73-603 HPCI Turbine Exhaust Water Apply between 73-23 and 73-603 73-609 itPCI Cthaust Drain Water Apply between 73-24 and 73-609 263
3;.
TA3LI 3.7.H (Cc.tinued)
]
i) o t
X-1073 Spare (testable)
X-108A Power X-1083 CRD Rod Positien Indic.
X-lC9 X-110A Pever X-1103 CR3 Rod Pesitica Indic.
X-230 Containment Air Monitoring Systes 4
266
e e
6 r
_ I.7,A & 4.7.A Pebery Containment The integrity of the primary containment and c:eration of the core standby cooling system in c:moination, limit the off-site doses to values loss than those suggested in 10 CFA 100 in the event of a break tr. the primary system piping., Thus, containeen integrity is specified whenever the potential for violation of the primary reac*.or system integrity exists. Concern about such a violation exists wnen.
An ever the reactor is critical and above atmospheric pressure.
exception is made to this requirement during initial core loading and while the. low power test program is being c:ncutted and ready access to the reactor vessel is required. There will be no pressure on the system at this time, thus greatly reducing the chances of a pipe break. The reactor may be taken critical during th1s peried; newever, restrictive operating procecures will be in effect again to minimi:e the pectability of an accident Occur rig. Procedures and the Rod Worth Minimiter would limit centrol worth such that a od drop would not result in any fuel damage. In additien, in the unlikely event that an excursion did occur, the react:r building and stancby gas treatment system, which shall be ccerational during this time, effer a sufficient barrier to keep offsite doses well
' telow 10 CFR 100 limits.
The pressure su: ression pool -ater prevides the heat sink for the react:r prieary system energy release following a :ostulated rustare of The pressurt suppression :nameer water vnlume ust absore the system.
tne associated decay and structural sensible seat released during primary system blowcewn frem 1.035 osig. Since all of the gases in the drywell are purged into the pressure suppressicn enamber air 5; ace during a 1 css of ecolant accident, the cressure resulting from isothernal compression plus the vaece cressure of the licuid must. net escted 52 ;sig, the su:pression chameer maximum pressure. The design volume Of the.
suppression cham er (water and air) was ettainec by c nsidering that the, total volume of react:r coolant to be concensec is cisenargec to the su:gression chamte-arc :nat the drywell volume is : urged to the su:pression Cna.:er.
Using the =ini=u= or =ax1=u= vater levels given in the specificatiens, con-cainment pressure during the design basis accident is approxi=acely 49 psig, vnich is belov the axi=u= of 62 psig. The =axi=u= vater level indi-cation of -1 inch corresponds to a downco=er sub=ergence of 3 feet 3
7 inches and a water volu=e of 127,800 cubic feet with or 128.700 ft without the dryvell-suppression chamber differential pressure control. The =ini=c=
vater level indication of -6.25 inches with differential pressure control and inches without differential pressure control corresponds
-7.25 to a downco=er submergence of approri=ately 3 feet and a water volu=e of approximately 123,000 cubic feet. Maintaining the water level between these levels vill assure that the torus water vole =e and down-co=er submergence are within the afore=entioned li=its during nor=al plant operation. Alar:s, adjusted for instru=ent error, will notify the operator when the li=its of the torus. vater level are approached.
The =ajority of the 3odega :ests were run with a sub=erged length of i feet and with co=plete condensation. Thus, with respect to down-co=er submergence, this specification is adequate. The max 1=u=
ce=parature at the end of blevdown tested during ths Hu=boldt Say and Bodega Bay tests was 170*y and this is conservatively taken to be the li=it for complete condensation of the reactor coclant, although condensation vould occur for te=peratures above 170*F.
267 Amendment No. 76
l l
L.
I i
6ASES Should it be necessary to drain the suppres; ion chamber, this should only be done hen there is no requirement for core standby cooling systems operatibilit w
Under full power operation conditions, blowdown from an initial suppression chamber water temperature of 95'F results in a peak long term water temperature of 170'F which is sufficient for complete condensation. At this temperature and atmospheric pressure, the available NpSH exceeds that required by both the lhR and core spray pumps, thus there is not dependency or. containment overpressure.
Experimental data indicate that ex, _2ive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 200*F local.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressuirzed in a timely manner to avoid the regime of potentially nigh suppression chamber loadings.
Limiting suppression pool temperature to 10S*F during RCIC, HPCI, or relief valve operation when decay heat and stored energy is removed from the primary system by discharging reactor steam directly to the suppression chamber assures adequate margin for controlled blowdown anytime during RCIC operation and assures margin for complete condensation of steam from the design basis loss-of-coolant accident.
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include:
(1) use of all available means to close the valve, (2) initiate su:pression pool water cooling heat exchangers (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the. pool.
If a loss-of-coolant accident were to occur when the reactor water temperature is below approximately 330*F, the containment pressure will not exceed the 52 psig code permissible pressures even if no condensation were to occur. The maximum allowable pool temperature, whenaver the reactor is above 212*F, shall be governed by this specification. Thus, specifying water volume-temperature requirements applicable for reactor-water temperature above 212*F provides additional margin above that available at 330'F.
268 Amendment No. 76
.y 3,
In conjunction with the lurk I Containment Short Term Program, a plant unique enalysis was performed :( Torus Support System and Attached Piping Analysis fer the.3rowns Terry Nuclear Plant Units 1, 2, and 3," dated September 9,1976 and supplemented October 12, 1976) which demonstrated a factor of safety of at
. least two for. the weakest element in the suppression cha=ber support system and~ attached piping. The maintenance of a drywell-suppression chamber differen-tial pressure of 1.1 peid and-a suppression chamber water level corresponding to a downcomer submergence range of 3.06 feet to 3.58 feet vill, assure the
' integrity of.the suppression cha=ber when subjected to post-LOCA suppression poo!,hydredynamic forces.
Inerting The relatively small contain=ent. volume. inherent in the CI-3WR pressure suppres-
' tion contain=ent and the large a=ount of circoniu= in the core are such that the occurrence of a' very limited (a percent or co) retction of the circonium and steam during a loss-of-coolant accident.could lead to the liberation of hydrogen combined with an air atmosphere to result in a fla==able concentration in che containment. If a sufficient amount of hydrogen is generated and oxygen is available in stoichiometric qu.tntities the subsequent ignition of the hydrogen in rapid recombinacien rate could lead to failure of the containment to maintain a low-leakage integrity. The <4% hydrogen concentration minimites the possibility I
of hydrogen combustion following a loss-of-coolant accident.
269 Amendment No. 76
I E^EE.$.
/ Die occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable then the occurrence of the loss-of-coolant accident upon which the specified oxygen concentration limit is based. Permitting access to the dryvell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without 'significantly reducing the margin of safety. Thus, to preclude-the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods, when the pri=ary system is at or near rated operating temperature and pressure. The 24-hour period to provide inerting is judged to be sufficient to ' perform the leak inspection and establish the required oxygen concentration.
To ensure that the hydrogen concentration is maintained less than 4% following an accident, liquid nitrogen is maintained on-site for containment atmosphere dilution. About 2260 gallons would be sufficient as a 7-day supply,.and replenishment facilities can deliver liquid nitrogen to the site within one day; therefore.
- a requirement of 2500 gallons is conservative. Following a loss of coolant accident the Containment Air Monitoring (CAM) System continuously monitors the hydrogen concentration of the containment volume. Two independent systems ( a system consists of one hydrogen sensing circuit) are installed in the drywell and the torus. Each sensor and associated circuit is periodically checked by a calibration gas to verify operation. Failure of one system does not reduce the ability to monitor system atmosphere as a second independent and redundant system will still bc operable.
In terms of separability, redundancy for a failure of the torus system is based upon at least one operable drywell system. The drywell hydrogen concentration can be used to limit the torus hydrogen concentration during post LOCA conditions. Post LOCA calculations show that the CAD system initiated within two-hours at a flow race of 100 scfm will limit the peak drywell and wetvell hydrogen con-contration to 3.6% (at-4 hours) and 3.S% (at 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />), respectively.
This is based upon purge initiation after 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> at a flow race of 100 scfm'to maintain containment pressure below 30 psig. Thus, peak torus hydrogen concentration can be controlled below 4.0 percent using either the direct torus hydrogen monitoring system or the drywell hydrogen monitoring syste's with appropriate conservatism (si 3.8%),
as a guide for CAD / Purge operations.
Amendment No. 76 270
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C e
5,0 MAJnd 9f.%lC:8 FLA1URES 5.1 s tTr. f tAftmr5 Browns Terry unit 1 is located at Browns Terry Nuclear F1 sat site on property owned by the United States and in custody of the TVA.
The site shall consist of approntmately Sto acres on the nort.. shore of Wheeler Lake at Tennessee River Mile 294 in Limestone County. Alabama. The minimum distance from the outside of the secondaty cantainme-t building to the boundary of the exclusion area as defined in 10 CTR 100.3 shall be 4,000 feet.
5.2 REACTOR A.
The reactor core may contain 764 fuel assemblies cer.risting of 8x8 assemblics having 63 fuel rods each, and 8x8R (and F8x8R) asse-blies having 62 fuel rods each.
B.
The reactor core shall contain 185 cruciform-shsped control rods. The control material shtll be boren carbide powder (B C) compacted to ap,roximately 70 percent of theoretical n
g density.
5.1 REACTot VV.sstt The reactor vessel shall be as described in Table 4.2-2 of the TSAR. The applicable design codes shall be as described in Table 4.2-1 of the FSAR.
5.4 CONTAlvstNT A.
The principal desten pstameters f or the primary contair.*ent shall be as given in Table 5.2-1 of the r $ A ?.. The applicable desite codes shall be as described in Section 5.2 of the T:AA.
I 8.
The secondary centainment shall be sa described in Section 5.3 of the r$AR.
C.
Penetrittons to the primary containment and pipinP. passing throush such penetrations shall be destened in accordance v;th the standards set forth in Section 5.2.3.4-of the TSAR.
5.5 FUEL STntAct A.
The arrangement of fuel in the new-fuel storate fact 11ty 8shall-be such that k for dry conditions. Is less than 0.90 and flooded is Itse than 0.95 (Section 10.2 of TSA?.).
330 Atendment NO. 76
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The k gf sf :he spen: fuel s::: age ;cel shall be less than or equal :: 0.95.
C.
Leads gras::: :han 1000 ;:unds shall c.s: he car:ied ver s;a.:
fuel asser:11es s:::sd in :he s;en: fusi ;eci.
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The stati:n citas s:::::;-.s Lad 37ste:s have bes: tesig.ed
- viths:er.1 a design ': asis es : aqua?.e. vita ;;: :: 1::aiert:i::
ef 0.2g.
The :;ertti:sti tssis es thqua?.e used in the ;;t::
design assussi a g :cni S::eis t-i:n f C.lg (see Ie::ien 2.8 er
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331 Amencment No. 74
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