ML19269E572
| ML19269E572 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/23/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19269E573 | List: |
| References | |
| NUDOCS 7906290300 | |
| Download: ML19269E572 (29) | |
Text
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f UNITED STATES-
'4 NUCLEAR REGULATORY COMMisslON tw
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ARKANSAS POWER & LIGHT COMPANY DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE - UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 43 License No. DPR-51 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment by Arkansas Power and Light Company (the licensee) dated Noverber 9,1978, as supplemented February 27 and April 26, 1979, and February 23,1979, as supplemented March 19, 20 and 30,1979, comply 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 confomity with the applications, tb: provisions of the Act, and the rules and regulations of tne Conrnission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the co. mon 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 Comission's regulations and all applicable requirements have been satisfied.
2146 122
.e 7906290geo'
2-Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this licen 2,
No, DpR-51 is hereby amended to read as follows:
- 2. c, (.2) Technical Soecifications The Technical Specifications contained in Appendices 43, are A and B, as revised through Amendment No.
The licensee hereby incorporated in the license.
shall operate the facility in accordance with the Technical Specifications.
The license is further amended by revising paragraph 2.c.(3) to read as follows:
The licensee may proceed with and is required to complete 2.c.(3) the modifications identified in Paragraphs 3.1 through 3.19 of the NRC's Fire Protection Safety Evaluation (SE) 22, 1978 and supplements on the facility dated August These modifications shall be completed as thereto.
specified in Table 3.1 of the Safety Evaluation Report In addition, the licensee may or supplements thereto.
proceed with and is required to complete the modif Safety Evaluation Report, and any future supplements.
These modifications shall be completed by the dates identified in the supplement.
This license amendment is effective as of the date of its issuance.
3.
FOR THE NUCLEAR REGULATORY COMMISSION e
(
Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
- Changes to the Technical Specifications Date of Issuance: May 23, 1979 2146 123 em-.,
,n
ATTACFNENT TO LICENSE AMENCMENT NO. 43 FACILITY OPERATING LICENSE NO. CPR-51 00CKET NO. 50-313 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 9
9 9b 9b 11 11 12 12 14b 14b 15 15 35 35 35a 35a 47 47 48 48 48a 48a 48b 48b 48bb 48bb 48bbb 48bbb 48c 48c 48cc 48cc 48ccc 48ccc 48d 48d 48dd 48dd 48ddd 48ddd 48f 48f 48g 48g 48h 48h 53d 53d 53e 66n 66n 2146 124
Using a local qualsty limit of 22 percent at the point of minimum DNBR as a ba.sts for curve 3 of Figure 2.1-3 is a conservat sve criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.
The DNBr. as calculated by the BAW-2 cora elation continually increases from point of minimus. Gi3R, so that the exi t CNBR is always higher and is a funct ion of the pressure.
The maximum thermal power for three pump operation is 85.4 percent due to a oower level trip produced by the flux-flow rat o (74.7 percent flow x 1.057 =
73.9 percent power) plus the ma21 mum calibration and instrumentation error.
The maximum thermal power for other reactor coolant pump conditions is pro-duced in a similar manner.
For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor cool ant pump situation.
Curves 16 2 of Figure 2.1-3 are the most restrictive because any pressure / temperature point above and to the left of this curve will be above and to the left of the other curve.
REFE RENCES (1) Correlation of Critical Heat Flua in a Bundle Cooled by Pressurtzed Water, BAW-10000A, May, 1976.
(2) FSAR, Section 3. 2. 3.1.1.c D *
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120 UNACCEPTABLE
-18 112 25.112 OPERATION I
ACCEPTABLE 100 4 PUUP 45,95 45,92 OPERt. TION
-l8,85.4 25.'85.4 80 ACCEPTABLE 2
3 & 4 PUMP 45,68.4 45,65.4 OPERATION 60 25,58.5 18,58.5 ACCEPTABLE 3
41.3.41.3 2,3 & 4 PUMP 40 OPERATION
~~
20 i
40
-40
-20 0
20 40 60 Reactor Power imoalance, 5 CURVE REACTOR COOLANT FLOT (GPM) 1 374,880 2
280,035 3
134.441 CORE PRaiECTION SAFETY LIMITS Figure 2.1-2 Amendment No.
, 43
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2.3 LIMITING SAFF"Y SYSTD4 SETINGS, PROTECTIVE INSULHDTATION Applicability Applies to instru=ents monitoring reactor pover, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet te=perature, flev, number of pu=ps in operation, and high reactor building pressure.
Objective To provide automatic protection action to prevent any combination of process variables frem exceeding a safety limit.
Specificatien 2.31 The reactor protection syste= trip setting li=its and the permissible bypasses for the instru=ent channels shall be as stated in Table 2.3-1 and Figure 2 3-2.
Bases The reactor protection syste= consists of fcur instru=ent channels to moniter each of several selected plant conditions which vill cause a reactor trip if any one of these conditions deviates free a pre-selected operating range to the degree that a safety limit =ay be reached.
The trip setting limits for protection systes instrumentation are listed in Table 2.3-1.
The safety saalysis has been based upon these protection system instru=entation trip set points plus calibrstion and instru=entation errors.
Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.
During normal plant operation with all reactor coolant pu=ps operating, reactor trip is initiated when the reactor pcver level reaches 105 5 percent of rated power. Adding to this the possible variation in trip set points due to calibration and instrument errors, the maximun actual power at which -
a trip would be actuated could be p2%, which is the value used in the.
safety analysis.
A.
Overpower trip based on flov and imbalance The power level trip set point produced by the reactor coolant system flow is based on a pover-to-flev ratio which has been established to accom=odate the =cs; severe ther=al transient considered in the design, the less-of-coolant flew accident fres high pcver. Analysis has de=en-strated that the specified power to flev ratio is adequate to prevent a DN3R of less than 1.3 should a low flow condition exist due to any ele -
trical =alfunction.
1; g3, Arendrent No.
f-2146 127
=
The power level t rip set poi nt produced by the power-to-flow ratto provides both har.F powc r level and low flom protection in the event the reactor power level snere.ses or the react or coolant flow rate dec reas es. The power level t rip set point produced by the power to flow rat to provides overpower DNR prot ect ion fur all modes of pump operation.
For eve ry flow rate there as a maximm permissable power level, anJ for every rower level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are ss follows:
1.
Trip would occur when four reactor coolant p.mps are operat ing af power is 103.7 percent and reacter flow rat e as 100 percent or fios rat e i s 9A..o. perc ent and pow e r level is 100 percent.
2.
Trip would occur when three react o r cool s.-t neps are operating if power is 73.9 percent and reactor flow rat e is 74.7 percent or flow rat e is 70.9 percent and power level is 75 percent.
3.
Trip would occur when one reactor coolant pump is operating an each loop (total of two pumps operating) If the power is 52.0 percent and reactor flow rat e is 49.2 percent or flow rate is 46.3 percent an d t he pow e r le ve l 15 49.0 percent.
l The flux / flow rat ios secount for ths
- n. u s m.;. cat tbration and instrmentation errors and the m.uimum vari st s on f rem the as e rage value of the RC flow signal in such a manner that the reset or prot ect tve system receives a conservative indication of the RC flow.
No penalty in reactor coetant thw through the core was taken for an open cote vent valve because of the core vent vahe surveillance program during each refueling outage.
For safety analysa calculations the maximus cali-bration and inst rumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor t hermal limits from being exceeded. These therma l Itmt ts are either power peaking kW/ft limit s or DNBR limits. The reactor power i:abalance (power in top half of core manus power an the bot tom hal f of core) reduces the power level trip produced by the power-to-flow ratto so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power, level trip associated reactor power-to-reactor power imbalance boundaries by 105 7 percent for a 1 percent flow reduction.
B.
Pump monitors In conjunction with the power imbalance / flow trip, the pump moni-itors prevent the minimum core DNBR from decreasing below 1.3 by trip-ping the reactor due to the loss of reactor coolant pump (s).
The pump monitors also restrict in operation.
the power level for the number of pumps C.
ACS Pressure Durint, a startup seeiilent from low power nr a slow rod withdr:wal f rom high powce, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit Amendment *No. [ *'.y g D
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7;45 379
THERMAL POWER LEVEL. 5
. 120 UNACCEPTABLE OPERATION 10,105.7 105.7 15,105.7 Mg = 0.747
. 100 y2 =
-0. 913 ACCEPTABLE
-31,90 4 PUMP 30,92 OPERATION 10.78.95 80 15.78.95
/
ACCEPTABLE 3 & 4 PUMP 30.65.25
-31.63.25 OPERATION 60 10,52.0 52 15.52.0 40
-31,36.3 ACCEPTABLE 30.38.3 t
2,3 & 4 PUMP i
OPERAT!ON 20 E'
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20 40 60 Power imostance, 5 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS Figure 2.3-2 Amendment No.
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Reactor Protection Syste s Tria Set:ing li-its a
D A-Ej 0.ie peac tar Cool uit 5%cp i
o Tour neact.sr Coolant Pawap s Three Reactor C.nlant rurps og.cratics in fach toap Operat ing (Norinal ty.:r.s t ing (% amin a,1 (f.--
i a a l Op. r a : a ng shi.e.Im.n
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,,,g re fr. i Jue *o y
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-t. il a me * ( s) ras N.u t e.a r power bascJ on NA NA 55%
p ur p c.on i t o r s, i of eypasscJ l*J r a t t.I.
van ingh re.cior coolant 2355 2355 2355 i nc' il
% st m,iicss.a.
- psis, cu a I
a low reactor coolant sys-1500 1800 1800 t r ai piessure, psig. nin Pyp nseJ va i.at,l e to. reactor (ll.75Tou t -5103) (1)
(ll.75Tou t -5103) (1)
(ll. Nou t -I' 3 03) ( 3 )
co >l.in s y s t ea pr e ssu re,
N asscJ psig.
m.
ac u tor coolant t e mp.
619 619 619 619 I, r.a a llegh reactor building 4 (18.1 psia) 4(18.7 psia) 4 (13. 7 psia) 4(li.? psi piessure, psig, saa (1 ) T is in degrees.'ahrenheit (F).
(3) Autonatically set when other sez icnt s of the CPS (as specified) are b;p:ssed out (2) Reactor coolant syste s flow, %.
(,4 ) The pucp monitor s also produce a t rip on: (a) loss of two reactor coolant par.ps in one reactor coolant loop, and (b) la.s of one or 16o reactor coolant pumps during t6o-nurip operation.
CD
Mininte volunes (including a 10". safety fsetor) as specified by Figure 3.2 1 for the horic acid addition tank or 35,659 gallons of 2270 ppm boron cs boric acid solution in the horated water storage tank (31 will cach satisfy this requirement. The specification assures that adequate supplies are available whenever the resttor is heated above 200 F so that a single failure will not prevent borstion to s cold condition. The minimum volumes of boric acid solu-tion given include the boron necessary to account for xenon decay.
The principal method of adding boron to the primary system is to pump the con-centrated boric acid solution (8700 ppm boron, minimum) into the makeup tank using the 25 gpm boric acid pumps.
The alternate method of addition is to inject boric scid from the boro-ated water storage tank using the nakeup pumps.
Concentration of boron in the boric acid addition tank may be higher than the concentration which would crystalli: at ambient conditions. For this reason and to assure a flow of boric seid is available when needed this tank and its 0
associated piping will be kept 10 F above the crystallization temperature for the concentration present. Once in the makeup system, the concentrate is suffi-ciently well mixed and diluted so that normal system temperatures assure boric acid solubility.
REFERENCES (1)
FSAR, Section 9.11 9.2 (2)
FSAR, Figure 6-2 (3)
FSAR, Section 3.3 D *
- 0)
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' T V 6
_ o w h\\ w Ju. A.Y a
2146 131
- cene. ment 80.
43; 3
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i Figure 3.2-180RIC ACIO ADDITION TANK VOLUME AND CONCENTRATION REQUIREMENTS VS RCS AVERAGE TEMPERATURE 7000 (579'F,6138 gal) 6000 5000
?
w" OPERATION ABOVE AND T0 (532*F,4427 gal) (579*F,4392 gal) g THE LEFT OF THE CURVES h
IS ACCEPTABLE 4000 g
(500*F 3826 gal) m
[
(532*F,3166 gal) 2 3
3000 a
o j
8700 PPM BORIC AC10 u
~
(400*F,2103 gal) 12000 PPM BORIC ACIO 1000 (300*F,877 gal)
(300*F 628 gal) 1 1
I
- I O
200 300 400 500 600 1
RCS Average Temuerature (F)
Amend =ent No. y[, 43 35a
e
- 6. If a control rod in the regulat ing ur maial ptmer shaping groups i s dec la red s nope rable per Speci fic
.on 4. 7.1. 2. operat ion above 60 percent of the thermal pomer allomable for the reactor coolant pump cambination may continue provided the rods in the group are posittened such that the rod that was declared' inoperable is con-tained within allowable group average position limits of Specifica-tion 4.7.1.2 and the withdrawal limits of Specifiest ion 3.5.2.5.3.
3.5.2.3 ne worth of single inserted control rods during criticslity are limite) by the restrictions of Specification 3.1.3.5 and the Control Rod rosition Limit s defined in Specification 3.5.2.5.
3.5.2.4 Quadrant tiIt:
1.
Except for physics tests, if quadrant tilt exceeds 4.92% power shall be reduced immediately to below the power level cutoff (see Figure 5 3.5.2-1A and 3.5.2-1B1 Moreover, the power level cutoff value shall be reduced 2% for each 1% tilt in excess of 4.92% tilt. For less than 4 pump operst ion, thernal power shall be redueed 2% of the thermal power allowable for the reactor coolant pump combin-ation for each 1% t ilt in excess of '4.92%.
2.
Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be reduced t.o less than 4.9'2% except for physics tests, or the following adjust-ments in setpoint.- and limits shall he made:
a.
The protect ion system maximwn allowahle set points (Figure 2.3-2) shall be reduced 2% in power forcaeh 14 tilt.
b.
The control rod group and APSR withdrawal li: sits 'shal'1 be reduced 2% in power for each 1% tilt ln excess of 4.92%.
c.
ne operational imbalance limits shall be reduce'd 2% in power for each 1% tilt in excess of 4.92%.
3.
If quadrant tilt is in excess of 25%, except for physics tests or diagnostic testing, the reactor will be placed in the hot shutdown condition. Diagnostic testing during power operation with a quad-rant power tilt is permitted provided the thermal power allowable for the reactor coolant pump combination is restricted as stated in 3.5.2.4.1 above.
4.
Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15% of rated power.
3.5.2.5 Control rod positions:
1.
Technical Specifiestion 3.1.3.5 (safety rod withdrawn!) does not prohibit the exercising of individual safety rods as required by Tahic 4.!-2 or apply to inoperable safety rod limits in Technical Speci ficat ion 3.5.2.2.
2.
Operat ing red group overlap shsil be 20% 1 between two sequential l
5 groups, except for physics tests.
Amendment No. A, g 1, g 4
D
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3.
Except for physics tests or exercising control rods, a) the control rod withdrawal limits are specified on Figures 3.5.2-1A, 3.5.2-1B and 3.5.2-1C for four pump operation and on Figures 3.5.2-2A, 3.5.2-2B and 3.5.2-2C for three or two pump operation and b) the axial power shaping control rod withdrawal Itmits are specified on Figures 3.5.2-4A, 3.5. 2-4 B and 3.5.2-4C. I f any of these control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours.
4 Except for physics tests, power shall not be increased above the power level cutoff (see Figures 3.5.2-1) unless the xenon reactivity is within 10 percent of the equilibrium value for operation at rated power and asymptotically approa:hing stability.
3.5.2.6 Reactor Power Imbalance sha!! be monitored on a frequency not "to exceed two hours during power operation above 40 percent rated power. Except for physics tests, imbalance shall be maintained within the envelopes defined by Figures 3.5.2-3A, 3.5.2-38 and 3.5.2-3C.
If the imbalance is not within the envelopes defined by Figures 3.5.2-3A, 3.5.2-3B and 3.5.2-3C corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met.
3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.
Bases The power-imbalance envelopes defined in Figures 3.5.2-3A, 3.5.2-3B and 3.5.2-3C are based nn 1) LOCA analyscs which have defined the maximum linear heat rate (5cc Fig. 3.5.2-4) such that the maximun clad temperature will not exceed the final Acceptance Criteria and 2) the Protective System Maximum Allowable SetW ints (Figure 2.3-2).
Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary. Operation in a situation that would cause the final acceptance criteria to be approached should a LOCA occur is highly improbable because all of the power distribution parameters (quadr 2nt tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.' Conserva-tism is introduced by application of:
a.
Nuclear uncertainty factors b.
Thermal calibration c.
Fuel densification effects d.
Hot rod manufacturing tolerance factors e.
Fuel rod bowing The 20 t5 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.
Control rods are arranged in groups or banks defined as follows:
' Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument and calibration errors. The method used to define the operating limits is defined in plant operating procedures.
Ncf.'j/,[,
p Amend =ent 48 6
x
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I; mop linnetinn I
Sa fe t y 2
Sa fe ty 3
Sa fe ty
'I 5afoty 5
Regulating 6
Hegulating 7
Xcnon traasient override 3
ApSit (axial power 3haping bank) 1hc rod position limits are based on the most limiting of the following th ree cri teria: LiCCS power peaking, shutdown taargin, and potential ejected rod worth. As discussed above, compliance with the CCCS pcwer peaking criterion is ensured by the rod position limits. 1hc minimwa availabic rod worth, consistent with the rod position limits, provides for achicvini; hot shutdown by reactor trip at any time, assuming tite highest worth control rod that is wi thdrawn remains in the full out pon-tion (1). 1hc rod position limits also ensure that inserted rod groups will not contain s ingic rod worths greater than 0.65*. lik/k at rated po'..cr.
1hese values have been s'iown to be safe by the safety analys ts (2) of the hypothetical rod ejcetion accident. A:1aximum single inserted control rod worth of 1.0f. ak/k is allowed by the rod positions iiuits at not :ero power. A single inserted control rod worth of 1.0". AL/L at 1)cginning of l i fe, ho t, zeto power would result in a lower transient peak ther.a1 jiower and, the re fo re, less severe environmental consequenec8 than a 0.050 Ah,'k cjccccd rod worth at rated powcr.
Control rod groups are wi thdrawn in settuence beginning with grot:p 1.
G mup s 5, 6, and 7 are overlapped 20%. 1hc normal position at power is for groups 6 and 7 to be partially inserted.
1he quadrant ;iower ti lt limits set forth in Specificat ion 3.5.2.
have beca established within the thermal analysis design base using the dc tinition
~
o f.p tad ra n t powe r til t gi sea in Technicul Specificaticus, Section 1.o.
1he;c limi ts in con.innetion with the control rod pos it ion l in.i t s in Specif-icat. ion 3.5.?.5.3 ensure that des inn peak heat r,t e e t i i eri a a re not cweeded during normal operat ion when including the effect:. of pot ent ial f:ic i iie r. i-fi ca tion.
the iptad rant tilt and axial imb;ilance monitoring in Spi e fications
- 1. 5..' l. 4 aad 3. 5. 2.6, respect ively, will nor:aally be performed in the plant cum-pute:. 1he two hour frequency for monitoring thesc <piaatttics will provide
.idequate surveillance when the computer is out of service.
Unring the physics tes t i ng program, the high flux trip setpoints are ad;.;inis-t ra t i ve ly set an fulinws to ensure that an additional safety margin is ;5ro-vided:
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2146 140 Amendment No.
43 43ccc e
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A N O -l, CYCLE 4 Figure 3.5.2-3A Amendment No.
, 43 45d 2146 141
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2146 143
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Amendment No.
.,.43 4Sh 2146 146
3.5.5 Fire Detection Instrumentation Acolicability This specification applies to fire detection instru=en:ation utili:ed within fire areas containing safety related equipment or circuitry, for the purposes of protecting that safety related equipment or circuitry.
Objective To provide i==ediate notification of fires in areas where there exists a potential for a fire to disable safety related syste=s.
Soecification 3.5.5.1 A sinimum of 30*a of the heat /s=oke detectors in the locations specified in Table 3.5-5 shall be operable.
3.5.5.2 If less than 50'5 of the fire detectors in any of the loca:ior.s designated in Table 3.5-5 are operable, within one hour establish a fire watch patrol to inspect the :ene(s) with the inoperable instrument (s) at least once per hour and restore the equipment to operable status within 14 days or prepare and submit a report to the Co= mission pursuant to Specification 6.12.3.2(b) within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s)to operable status.
Bases The various detectors provide alarms that notify the operators of the existence of a fire in its early stages thus providing early initiation of fire protection. The detectors in :he =ain and auxiliary control rooms also provide auto =atic fire protection initiation.
The detectors required to be operable in the various areas represent 1/2 of those installed.
Operability of the fire detection instrumentation ensures that operable warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stiges. Prompt detection of fires will reduce the potential for damage to safety related equipment.
In the event that a portion of the fire detection instru=entation is inoperable, the establish =ent of frequent fire patrols in the affected areas (s) is required to provide detection capability until the inoperable instrumentation is restored to operability.
53d
' Amendment No.
,N 2146 147
Table 3.5-5
+
Safety-Related Areas Protected By Heat /Scoke Detectors 1.
Each of the four reactor building cable penetration areas.
2.
Each of the four cable penetration rooms.
(Zones 149-E, 144-D, 112-I, 105-T)*
3.
Each of the two emergency diesel generator rooms.
( enes 86-G, 87-H) 4.
North switchgear room.
(Cone 99-m) 5.
South switchgear room.
(Cone 100-N) 6.
Main control room.
( ene 129-F) 7.
Auxiliary control room ceiling.
3.
Auxiliary control room floor.
9.
Each diesel generator fuel vault.
10.
Cable spreading room.
(Zone 97-R) 11.
Battery charger and inverter rooms and hallway.
(:one 98-J)
- 12. Spent fuel area.
(Zone 159-B) 13.
Computer transformer rccm.
(~one 167-3) 14 Controlled access area.
(~one 123-E) 15.
Tank room.
(:ene 68-P) 16.
Electrical equipment room.
(Cone 104-5) 17.
North upper piping penetration room.
(~one 79-U) 18.
South upper piping penetration room.
(Zone 77-V) 19.
Condensate desinerali:er area.
(Cone 73-W) 20.
Compressor room.
(Cone 76-W)-
21.
Radwaste processing area.
(Cone 20-Y) 22.
Storage and pipe area.
(Cone 34-Y) 23.
Pipe area.
(:one 40-Y) 24.
South lower piping penetration room.
(Cone 46-Y) 25.
Penetration Ventilation room.
(:one 47-Y) 26.
North lower piping penetration room.
(~one 53-Y) 27.
East decay heat removal pump room.
(;one 10-EE) 23.
Nest decay heat removal pump room.
( ene 14-EE) 29.
Intake structure.
Cone numbers reflect nomenclature in the Fire Ha ards Analysis and are listed for clarification only.
{ }kh AmendmentNo..k]
53e
3.18 Fire Suceression Serinkler Systems Aeolicability This specification applies to the following fire suppression sprinkler systems protecting safety-related areas:
a.
Each of the four reactor building cable penetration areas.
b.
Each of the four cable penetration rooms.
c.
Each of the two emergency diesel generator rooms.
d.
Cable spreading room.
e.
Each of the two diesel generator fuel vaults.
f.
Hallway-El 372. ( ene 98-J)
- 2 Condensate deminerali:er area.
Obiective To assure that fire suppression is available to safety-related equipment located in the above-listed areas.
Scecification 3.18.1 The above-listed sprinkler systems shall be operable at all times.
3.13.2 With one or more of the above-listed sprinkler systems inoperable, establish a continuous fire watch (or operable smoke and/or heat detection equipment with control room alarm) with backup fire suppression equipment for the applicable area (s) within one hour. Restore the system (s) to operable status within 14 days or prepare and submit a Report to the Cc= mission pursuant to Specification 6.12.3.2(b) within the nex: 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system (s) to operable status.
Bases Safety related equipment located in various areas is protected by sprinkler systems. The operability of these systems ensures that adequate fire suppression capability is available to confine and extinguish a fire occurring in the applicable areas.
In the event a system is inoperable, alternate backup fire fighting equipment or operable de:ection equipment is required to be made available until the inoperable equipment is restored tc service.
- To be i=plemented no later than July 30, 1979.
Amendment No. ^
66n
~
/
UNITED STATES
[^:3
- 1 NUCLEAR RECULATORY COMMissiCN I b #I S
WASHINGTON, D. C. 20553
$t i
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SAFETY EVALUATION SY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENCMENT NO. 43 TO LICENSE NO. OPR-51 ARKANSAS POWER AND LIGHT CCMPANY ARKANSAS NUCLEAR ONE, UNIT NO.1 00CKET N0. 50-313, 1.0 Introduction By letters dated Neverrter 9,1978, as supplemented February 27 and April 26,1979 (References 1 and 2, respectively), Arkansas Pcwer and Light Company (AP&L) requested amendment of Appendix A to Facility Operatin License No. OPR-51 for Arkansas Nuclear One - Unit No.
1 (ANO-1. Section 5 sumarizes the proposed amendment to the Technical Specifications (TS)phanges of this The AP&L submittal of November 9,1978 was presented to support operation of Cycle 4 folicwing the refueling perfomed at the end of.
Cycle 3.
As such, the analysis presented in the submittal was based on the intended ex::osure for Cycle 4 of 387 effective full power days (EFFD). Informaticn submitted describes the fuel system design, nuclear design, themal-hydraulic design, accident analyses, and startup test program.
The refueling of ANO-1 for Cycle 4 will result in a core loading consisting of 64 fresh Mark 8-4 assemblies, 57 once burned assemblies, and 56 twice burned assemblies. The fuel management has been changed from a conventional three fuel batch out-in scheme to a three fuel batch in-out-in scheme. The key feature of this scheme is the extensive use of fixed burnable poison in fresh reload fuel which will be loaded in the core in 'erior rather than on the core periphery. The maximum fuel batch exposure at the end of Cycle 4 is predicted to be 28,300 MWD /MTU and hence is considerably less than the value of 33,000 MWD /MTU used in staff environmental considerations. This report addresses Cycle 4 operation only. Operation of successive nominal 18 montn fuel cycles-will result in fuel batch exposures in excess of 33,000 MWD /MTU during Cycle 6.
The environmental impact of extended fuel burnup is to be addressed prior to Cycle 6 cperation.
Mf pn s
1mme#
2146 150
. 2.0 Evaluation of v difications to Core Desien o
2.1 Fuel System Design The 64 fresh Mark 8-4 fuel assemblies which are to be loaded for Cycle 4 are mechanically identical to previously approved and utilized fuel assemblies at ANO-1 and other Babcock and Wilcox (S&W) supplied nuclear s.eam sucply systems, NSSS. The mechanical design of the fresh fuel was not re-evaluated by the staff for Cycle 4 Modified burnable poison rod assembly retainers (Ref. 3) are to be used in Cycle 4 to insure positive retention of the burnable poison rod assemblies (SPRA's). These retainers have been previously approved for retention of Orifice Rod Assemblies (0RA's). Mechanical and thermal-hydraulic compatability of the BPRA retainers has been previously reviewed and accepted. The SPRA's weigh 49 lbs. Coolant flow past the BPRA's has been predicted by the licensee to provide a net lift force of 53 lbs. Therefore, there is a net upward force of 4 lbs. The retainers provide a downward force of 47 lbs. and hence a minimum pos-itive holddown force in excess of 30 lbs. Use of the modified BPRA retainers will therefore insure positive retention of the SPRA's. These retainers have been designed for one operating cycle and are to be replaced if BPRA's or ORA's are utilized in future cycles.
2.1.1 Cladding Creeo Collaose Fuel rod cladding creep collapse analyses have been performed for the most limiting (i.e., twice burnt exposed batch 4 fuel assemblies) fuel assembly to be used in Cycle 4 The analyses were perfomed according to the methods and assumptions described in References 4 and 5.
These analyses predict that the time to rod cladding collapse will be in excess of 30,000 effective full power hours. The maximum batcht4 assembly burnup during Cycle 4 is predicted to be 23,112 EFPH (Table 4.1, Ref. 1). We conclude that cladding creep collapse has been suitably considered.
2.1.2 Cladding Stress and Strain Stress calculations have been performed for a generic fuel rod model and strain calculations for a generic pellet model. These models and calculations have been approved for prior ANO-1 relcads. The licensee has asserted that Cycle a parameters are enveloped by these generic models. The licensee's calculations shcw that in no case does the stress exceed the yield.
2.1.3 Fuel Ther al Cesian The intreducticn of the batch 6 fuel does not introduce significant diff-erences in fuel themal per#crmance relative to the other fuel remaining in the core. The credicted linear heat rate to centerline melt (20.15kw/ft) is the same for batches 4, 5 and 6.
At the core average linear heat rate 2146 151
. (5.8kw/ft), the licensee predicted ncminal fuel temperatures of batches 4, 5 and 6 would be approximately 1300 F.
These values are typical of all PWR's. Licensee calculations were nrfomed using the approved computer code TAFY-3 (Ref. 6). It is cotee that the code TACO (Ref. 7) has also been approved for fuel temper 6ture calculations and is the staff preferred code. Based on the Cycle 4 predicted values and current approval of the analytic techniques used to make these predictions, the staff considers the fuel themal design acceptable and provides for no reduction in the margin of safety.
2.2 Nuclear Design Figure 3-1 of pef e ence 1 indicates the core loading trrangement far ANO-1 Cycle 4; the initial enrichments and burnup distributions are given in Figure 3-2.
An unconventional fuel management scheme has been utilized.
An in-out-in fuel management scheme has been adopted. Fresh 3.19 w/o U235 fuel will be leaded in the core interior in a checkerboard pattern.
Next cycle this fuel will be loaded on the core periphery.
In its third resident cycle the fuel will once again be loaded in the interior of the core in a checkerboard fashion, hence the term "in-out-in".
The fresh fuel will,contain BPRA's to hold down local reactivity.
Three concentrations of baron carbide (in.an alumina matrix) will be employed to tailor the radial power distribution. By loading fresh fuel in the core interior, rather than on the periphery, neutron leakage is reduced. In turn for a fixed core enrichment the cycle length will be increased. Alternately the designer may increase the cycle length by increasing the average core enrichment. Both techniques have been used for Cycle 4.
In-out-in fuel management is believed to tax the nuclear designer's analytic capability to a greater extent than conventional three batch fuel manage-ment. To insure that achieved power distributions in the core are within the bounds assumed in the safety and setpoint analyses, monthly incore power maps are to be taken. This is a current TS requirement. Pcwer distributions are to be compared with predicted distributions and the licensee has conTnitted to report the deviations as part of the plant's monthly operating report.
Reactivity control and power distribution control will be maintained by control rods, axial power shaping rods (APSR) and soluble boron concentration control. The red locaticns are given in Figure 3-3 of Reference 1.
The core will be operated wita control rods inserted at pcwer to 250 EJPH and the APSR's deeply inserted.
The projected Cycle 5 length is 387 EFP0 with a predicted cycle burnup of 12,111 MWD /MTU.
2146 152
. Cycle 5 nuclear parameters including critical baron concentrations, control rod worths, Doppler ccefficients, moderator coefficients, xenon worth and effective delayed neutron fractions have been calculated using the approved PCQ07 code (Reference 8). These are presented in Table 5-1 of Reference 1 and compared to the Cycle 3 values. Relative to Cycle 3, predicted critical boron concentrations have increased due to the greater excess reactivity at beginning of life which is required to achieve the 18 month fuel cycle. The increased soluble boron concentration and use of BPRA;s will make the core " blacker" to themal neutrons at beginning of Cycle 4 relative to Cycle 3.
The extended cyr.le will result in more fission products and hence a " blacker" core at enei of Cycle 4 relative to Cycle 3.
Small changes in the power defect, Doppler coefficient, moderator temperature coefficient and inverse bcron worth are consistent with increased core blackness.
Shutdown margins have been calculated for BOC and ECC (Table 5-2 of Reference 1). The calculated minimum shutdown margin during Cycle 4 is 1.77% aX/K which is larger than the value of 1% aX/K assumed in cooldown accident analyses by an adequate margin.
2.3 Themal-Hydraulic Desian The themal-hydraulic design conditions for ANO-1 Cycle 4 are included in Table 6-1 of Reference 1.
Only the minimum departure frem nucleate boiling ratio at Meady state differs from the Cycle 3 value.
The small difference, a 0.02 reduction in predicted DNBR at steady state 112% overpcwe'r, is attributable to the assumption in the thermal-hydraulic analyses that the hot assembly contained a BPRA. This assumption is consistent with beginning of Cycle 4 power distribution calculations (Fig. 5-1, Reference 1) which predict that the hot assembly will.in.. fact contain a BPRA.
2.3.1 Removal of Orifice Rod Assemblies Orf fice rod assemblies were removed, bypass flow reanalyzed, and required setpoints adjusted as part of the Cycle 3 evaluation. Relative to Cycle 3 an additional six ORA's will be removed (44 to 50 ORA's) for Cycle 4 The core bypass flow will be negligibly effected (approximately 0.1% decrease in core flow) by this change. The flux / flow trip setpoint will be reduced to acccmmodate this change. This setpoint is based on an assumed two-pump coastdown frem 102% indicated power (108% core pcwer assumed in analyses).
2.3.2 Effect of Rod Scw on Themal Desian The potential effects of fuel red bcw has been reviewed generically in Reference 9.
Based on the red bcw model acproved by the staff, ANO-1 nas apolied a ONBR penalty of 11.2% for fuel r0d bcw which has been incorporated in the variable low-pressure trip function and flux /ficw set::oint.
2146 153
. 3.0 Evaluation of Accidents and Transients General:
~
The licensee has stated that each accident analyzea in the FSAR has been examined and has found, with the exception of the moderator dilution postulated event, to be bounded by the FSAR and/or the Fuel Densification Report and/or subsequent cycle analyses.
The staff has concluded that the consequences of hypothesized events are no worse than those stated in the FSAR or previous submittals, that is, part 20 and part 100 dose rate limits will cot be exceeded in the event of an anticipated operatir.g occurrence or accident respectively.
The remcval of CRA's in Cycles 3 and 4 has resulted in increased bypass flow and corresponding decrease in core flow (approximately 1% decrease).
This effect has been considered in: calculating the steady state DNBR conditions. The licensee has assumed that the incremental DNBR degradation during an anticipated operating occurrcnce (A00) or accident has not been substantially altered by these changes. Hence, the FSAR analyses are bounding. This approximation is considered acceptable.
Soecific Analysis:
The licensee has stated (Reference 1) that the generic B&W ECCS analysis (Reference 10) is applicable to ANO-1, Cycle 4.
Based on the minimal core changes for Cycle 4 the staff accepts this assertion.
The conclusion presented in the FSAR is that, in the event of a steam line break (SLB) accident, a small fraction of the 10 CFR 100 dose rate would be reached. The supporting analysis assumed a 1% ao safeguards allowance (shutdown margin). The predicted minimum shutdown margin during Cycle 4 is 1.77% 46.
On these bases the consequences of a hypothesized SLS are considered acceptable for Cycle 4 operation.
The larger initial soluble boron concentration at beginning of Cycle 4, relative to the reference analyses, will result in a slightly larger reactivity insertion ' rate for the postulated moderator dilution accident.
For in assumed 500 gpm dilution rate the reactivity insertion is pre-dicted to be 1.235 x 10-5 aK/X/sec. A value of 1.227 x 10-5 ax/X/sec was assured in the Cycle 4 analysis. The higher insertion rate will result in a f aster reactor trip on high reactor coolant system pressure.
The licensee has predicted that the reactor ccolant pressure will increase by less than 10 psi relative to tne FSAR analyses leaving a margin of approximately 300 psi to the safety limit. This change is insignificant.
2146 154
. The dropped rod accident analysis reported in the FSAR is based on an assumed dropped rod worth of 0.65% aK/K, and a peak post dropped enthalpy
, rise, FaH, equal to the design value,1.78. Turbine runback to 60% of rated power was assumed not to function. The licensee has predicted that the maximum dropped rodworth is 0.2% aK/K. Post drop values of FaH have not been provided by the licensee. The peak enthalpy rise would increase by less than 20% following the drop of a control redworth only 0.2". aK/K.
Following the rod drop and assuming no turbine runback, the core will return to rated power. Since the core is typically operated with an initial enthalpy rise approximately 15% (or greater) less than the design peak, and even if the core was initially at the design peak and the peak were to increase by 20% there would still exist margin to DNBR limits (at 100% power),
the dropped rod analysis is considered adequate for Cycle 4 The most limiting transient considered as a part of the original licensing process was the postulated loss of AC power. The loss of all AC power would result in loss of reactor coolant pumps and leave forces flow as well as loss of normal feedwater. The postulated loss of feedwater is considered less limiting than the loss of AC power assuming no other single or multiple failures. Therefore, loss of feedwater has not been reviewed as a part of the proposed license amendment.
j The maximum ejected rod at hot full power, Cycle 4, is predicted to be 0.55% a6 or 0.893. The FSAR analysis assumed a rod worth of 0.65". ad or 0.925. The predicted Doppler coefficients during Cycle 4 are substantially less than the values used in the FSAR analyses. These are conservative changes relative to the FSAR analyses. The delayed neutron fraction (seff), is predicted to be smaller than assumed in the FSAR. The effect of the smaller value of Seff is a slower decay of the neutron flux once the peak value is reached. This is a non-conservative change. The above cited conservatisms are substantially larger than this non-conservatism. FSAR calculations were run using a point kinetics design model assuming the design three dimension peak and compared to two dimensional space-time kinetics calculations. The design model was shown to be conservative. Post ejected rod peaking factors have not been presented for Cycle 4 nor for the FSAR analyses.
Hence, a direct comparison cannot be made of these values. The conclusion of this analysis presented in the FSAR is that there exist substantive margins for this accident to limiting enthalpy deposition values. On this basis the applicability of the~FSAR analysis to Cycle 4 is accepted.
4.0 Startuo Tests Startup tests are described in Reference 1.
These tests are consistent with the startup tests perfomed in association with other recent Saw reloads. We have reviewed the tests in terms of their intended purpose and consider them acceptable. AP&L has agreed to provide a startup test report (Reference 2).
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. 5.0 Evaluation of TS Chances Proposed modifications to the ANO-1 TS are described below:
(1) TS Fig. 3.5.2-1A, B, C, Fig. 3.5.2-2A, B, O Rod position limits insure that values of shutdown cargin, ejected rod worth, peak linear heat rate, and peak enthalpy rise, realized during the operating cycle are less than or equal to the values used in the safety analyses. They have been modified, relative to Cycle 3, to accomodate changes in predicted peaking factors with rods inserted in the core which in turn have been altered by the revised fuel management.
The core is to be run at full power with control group 7 inserted and group 6 inserted as a bite bank till 250 EFPD and thereafter essentially unrodded with group 7 used as the bite bank. The Cycle 4 limits at full power are scmewhat more restrictive than the Cycle 3 limits.
(2) TS Fig. 3.5.2-4A, B, D The axial power shaping rods (APSR) are to be inserted near the bottom of the core throughout the cycle. APSR limits for Cycle 4 are more restrictive than the Cycle 3 limits.
(3) TS Fig. 3.5.2-3A, B, C Operational power imbalance limits in conjunction with rod position limits insure that the peak linear heat rate as a function of core height limits are not exceeded. Relative to Cycle 3, limiting bottom peaked axial power shapes are to be excluded for Cycle 4.
H6nce, these Cycle 4 limits are, as the red position limits, more restrictive than the Cycle 3 values.
(4) TS 3.5.2.5 Control group overlap is to be reduced from 25% to 20% cverlap. The change reduces the extent over which an overlap region must be considered at any given power. Overlap regions exhibit higher planar peaking than non-overlap regions.
Items (1), (2), (3) and (4) are considered together and may be traded against each other; it is the convolution of these limits which deter nines the peak linear heat rate and enthalpy rise. We find that these proposed changes are acceptable and do not decrease the margin of safety.
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(5) TS 3.2 and Figure 3.2-1 The increased boron acid addition tank volume is required to insure shutdown of the reactor to cold conditions (200 F) using soluble boron. The Cycle 4 core at beginning of cycle is simply more reactive (higher core average enrichment, lower core average burnup) than the Cycle 3 core.
(6) TS 2.3.1, Table 2.3-1, Figure 2.1.2 and 2.3.2 These small changes reflect revision of the flux / flow setpoint frcm 1.06 to 1.057. This proposed revision accomodates the addition of BPRA retainers and removal of an additional six ORA's.
Items (5) and (6) are considered separately. We find the proposed changes are acceptable and do not decrease the margin of safety.
6.0 High pre _ssure Injection (HPI) System Modifications _
On April 28, 1978, the Commission issued an Order modifying License No. CPR-51 to require that certain operating procedures be implemented until facility modifications could be implemented to alleviate the ECCS small break problem. By letter dated October 27, 1978, supplemented by letter dated January 3,1979, the licensee proposed certain facility modifications at ANO-1 to mitigate a small break LOCA without requiring operator action. By letter dated March 1,1979, we accepted the licensee's proposed modifications.
__The proposea modifications would ensure that the proper flow uplift in the HPI lines and assure the minimum ECCS flow to the reactor coolant system. By letter dated May 14, 1979, the licensee verified that the proposed modifications were implemented and tested to assure the proper flow split to the redundant legs of the HPI system. Thus the actions required by the Comission's Order dated April 28, 1978 have been completed.
7.0 Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level
,and will not result in any significant environmental imcact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standooint of environmental impact and, pursuant to 10 CFR 651.5(d)(4), that an environmental impact statement, or negative declaration and environ-
= ental impact appraisal need not be prepared in connection with the issuance of this amendment.
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8.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment dres not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant ha:ards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the preocsed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the corron defense and security or to the health and safety of the public.
Cated:
May 23,1979 2146 158
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References:
(1 )
W. Cavanaugh, III, Proposed TS, November 9,1978, AP3L ltr (File 1511.1).
(2)
D. A. Rueter, Cycle 4 Reload Report Questions, February 27,1978, APil letter (File 024.6,1511.i), and April 26,1979 (File 0242.6,1511.1).
(3) BPRA Retainer Design Report, BAW-1496, B&W, May 1978.
(4) Program to Deternine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084, Rev. 1, B&W, November.1976.
(5) ANO-1, Fuel Densification Report, BAW-1391, B&W, June 1973.
(6)
C. D. Morgan and H. S. Xao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-1004, B&W, May 1972.
(7)
R. H. Stoudt, et al., TACO - Fuel Performance Analysis, BAW-10087, B&W, June 1976.
(8)
H. A. Hassan, et al., B&W's Version of PDQ07 - User's Manual, BAW-10ll7, B&W, June 1976 (9) Memo to D. B. Vassallo (NRC) from D. F. Ross (NRC), Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thema.
Margin Calculations for Light Water Reactors, February 16, 1977.
(1C)
R. C. Jones, et al., ECCS Analysis of B&W's 177-FA Lowered Loop NSS, BAW-10103A, Rev. 1, B&W, July 1977.
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