ML20207N820
| ML20207N820 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/03/1988 |
| From: | Hannon J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20207K620 | List: |
| References | |
| NUDOCS 8810190377 | |
| Download: ML20207N820 (79) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION a
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TOLEDO EDISON COMPANY EE THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.123 License No. NPF-3 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated May 18, 1988, complies with the standards and requirements of the Atomic Energy Act of 19C4, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) 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 comon defense and security or to the health and safety of the public; and E.
The issuance of this anendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this licenst. amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:
SS10190377 estoo3 gDR ADOCK 05000346 PDC
. (a) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.123, are hereby incorporated in the license. The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGUL TORY COMMISSION ohn N. Hannon, Director Project Directorate III-3 Division of Reactor Projects - III, IV, Y, & Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: October 3. 1988 l
.s l
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ATTACHMENT TO LICENSE AMENDMENT NO.123 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
Revised Pages 2-2 3/4 1-29 3/4 2-3 1
1 2-3 3/4 1-29a 3/4 2-3a 2-5 3/4 1-29b 3/4 2-3b 2-7 3/4 1-29c 3/4 2-3c B 2-1 3/4 1-31 3/4 2-9 B 2-2 3/4 1-34 3/4 2-10 j
B 2-3 3/4 1-35 3/4 2-11 B 2-5 3/4 1-36 3/4 2-12 B 2-8 3/4 1-37 3/4 2-13 3/4 1-14 3/4 1-38 3/4 2-14 3/4 1-16 3/4 1-39 3/4 3-8 3/4 1-17 3/4 1-40 3/4 4-1 3/4 1-18 3/4 1-41 3/4 5-7 3/4 1-26 3/4 2-1 B3/4 1-2 3/4 1-28 3/4 2-2 B3/4 1-3 3/4 1-28a 3/4 2-2a B3/4 5-2 3/4 1 '8b 3/4 2-2b 5-4 4
3/4 1-28c 3/4 2-2c
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i
2.1 SAFETY LIMITS REACTOR CORE c
2.1.1 The combination of the reactor coolant core outlet pressure and l
outlet temperature shall not exceed the safety limit shown in Figure i
2.1-1.
L APPLICABILITY: MODES 1 and 2.
E t
ACTION:
[
Whenever the point defined by.the combination of reactor coolant core l
outlet pressure and outlet temperature has exceeded the safety limit, be in ROT STANDBY vithin one hour.
l REACTOR CORE 4
r 2.1.2 The combination of reactor THERMAL POVER and AXIAL F0VER IMBALANCE shall not exceed the safety limit shovn in Figure 2.1-2 for the various 3
combinations of two, three and four reactor coolant pump operation.
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APPLICABILITY: MODE 1.
l ACTION:
l Whenever the point defined by the combination of Reactor Coolant System flow, AXIAL POWER IMBALANCE and THERMAL POVER has exceeded the appropriate safety limit, be in ROT STANDBY vithin one hour.
i REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The Reactor Coolant System pressure shall not exceed 2750 psig.
APPLICABILITY: MODES 1, 2, 3, 4 and 5.
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ACTION:
i MODES 1 and 2 -
Whenever the koactor Coolant System pressure has I
exceeded 2750 psig, be in HOT STANDBY vith the l
Reactor Coolant System pressure within its limit
[
1 within one hour.
j J
Whenever the Reactor Coolant System pressure has
[
MODES 3, 4 and 5 exceeded 2750 psig, reduce the Reactor Coolant System l
I pressure to within its limit within 5 minutes.
]
l DAVIS-BESSE, UNIT 1 2-1 1
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Figure 2.1-1 Reactor Core Safety Limit 2500 RC High Pressure Trip (618,2300) 2300 RC High Temperature Trip ACCEPTABLE 2200 OPERATION l'
e
[
(618,2124.6)
(633.4, s'
2100 2129.8) s df RC Pressure (606.79,1983.4)
Temp Trip 2000 RC Low Pressure Trip Safety Limit 1900 (621.4.1929.8) 1800 (608.2.1729.8) 1700 e
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590 600 610 620 630 640 650 Reactor Outlet Temperature,*F DAVIS-BESSE, UNIT 1 2-2 Amendment No. AA, 33, 4y, fl.123
l Figure 2.1-2 Reactor Core Safety Limit
% RATED THERMAL POWER 120
(-44.0,112.0) 4 PUMP LIMIT (33.0,112.0)
(-49.0,100.0)
,100
(-44.0,90.0)
(33.0,90.0) 3 PUMP LIMIT (47.1,87.2)
- 80
(-49.0,78.0)(
)(47.1,65.2) 60 UNACCEPTABLE UNACCEPTABLE OPERATION ACCEPTABLE OPERATION OPERATION FOR SPECIFIED RC PUMP COMBINATION
-40 20 l
e t
I i
1
-60
-40
-20 0
20 40 60 AXIAL POWER IMBALANCE, 1 Reautred Measured Flow to Ens;re Pu os Ocerating Reactor C0olant Flew. oc*
Co*Dliance, ;em 4
380,000 3S9,500 3
2 S 3, 86 0 290.957 DAVIS-BESSE, UNIT 1 2-3 Amendment No. 17, 16. 33, it, 4}', ??, 7F, 123
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM SETPOINTS 2.2.1 The Reactor Protection System instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY: As shown for each channel in Table 3.3-1.
ACTION:
With a Reactor Protection System instrumentation setpoint less conserv-ative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
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1 DAVIS-BESSE. UNIT 1 2-4 l
U M
=
Table 2.2-1 Reactor Protection System Instrumentation Trip Setpoints E
m Functional unit Trip setpoint Allowable values 1.
Manual reactor trip Not applicable.
Not applicable.
2.
High flux
$104.94% of RATED THERMAL POWER vith
$104.94% of RATED THERMAL POWER with four pumps operating four pumps operatingt
$80.6% of RATED THERMAL POWER with
$80.6% of RATED THERMAL POWER with l
three pumps operating three pumps operatingt
)
3.
RC high temperature
$618'F f618'FS i
e 4.
Flux - aflux/flovI)
Four pump trip setpolat not to Four pump allowable values not to l
exceed the limit line of Figure exceer" the limit line of Figure 2.2-18.
2.2-1.
For three pump operation, for three pump operation, see Figure see Figure 2.2-1 2.2-1
)
5.
RC low pressure (I) 11983.4 psig 11983.4 psig* 11983.4 psig**
gg 6.
RC high pressure f230C psig
$2300.0 psig* f2300.0 psig**
7.
RC pressure-temperature 2(12.60 T F - 5662.2) psig 1(12.60 T F - 5662.2) psigt out out
[
8.
HighfluxgeberofRC
$$5.1% of RATED THERMAL POWER with
$55.1% of RATED THERMAL POUER with y
oumps on one pump operating in each loop one pump operating in each loopt i
g-
<0.0% of RATED THERMAL POWER with
<0.0% of RATED THERMAL POWER with
[
Ivo pumps operating in one loop and ivo pumps operating in one loop and "g
no pumps operating in the other loop no pumps operating in the other loopt
$0.0% of RATED THERMAL POWER with no 30.0% of RATED THERMAL PUUIR with no 1
pumps operating or only one pump pumps operating or only one pump op-operating eratingt i
9.
Containrent pressure high f4 psig
$4 psigt i
1
4 E
Table 2.2-1.
(Cent'd) m e
h IIITrip may be manually bypassed when RCS pressure $1820 psig by actuating shutdown bypass provided that:
U, T'
a.
The high flux trip setpoint is 3.iZ of RATED THERMAL POWER.
E b.
The shutdown bypass high pressure trip setpoint of <l820 psig is impesed.
-e c.
The shutdown bypass is removed when RCS pressure >l820 psig.
- Allowable value for CHAlWEEL FUNCTIONAL TEST.
- Allowable value for CRANNEL CALIBRATION.
BAllowable value for CHANNEL FUNCTIONAL TEST and CHAletEL CALIBRATION.
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c
o Figure 2.2-1 Trip Setpoint for Flux -- AFlux/ Flow
% RATED TkERMAL POWER UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION 120 Curve shows trip
(-17.0.108.0)
(17.0,108.0) setpoint for an approximately hn ortrI M =+1.00 M =-2.27 g
[4 PUMP 2
100 l
pump operatter.
(-30.6,94.4) O LIMIT l
(283.860 gpm).
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The actual set-I 80 (17.0,80.6) point will be
(-17.0.80.6) calculated by the l
Reactor Protecticn (30.6,77.1) 3 P[ UMP System and will be l
l directly pr0por-(- 30.6,67.0) e.XAMPLE tional to the
,. 6 0 l
l actuel flew with l
l lACCEPTADLE OPERATION FOR three pumps.
lSPECIFIED RC PUMP (30.6,49.7)
COM8INATION l
I
-40 g
1 l
l l
l l
[
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l 20 g
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t l
I t
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-80
-60
-40
-20 0
20 40 60 80 AXIAL POWER IMBALANCE. %
DAVIS-BESSE, UNIT 1 2-7 n
n o
$2
l Figure 2.3 2 Allowable value for Flux-A Flux /Tiov M
DAVIS-BESSE, UNIT 1 2-8 Amendment it. JP.W,d,45
2.1 SAFETY LIMITS BASES 2.1.1 AND 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which vould result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper coundary of the nucleate boiling regime vould result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNL is not a directly measurable parameter during operation and therefore THERHAL POVER and Reactor C6olant Temperature and Pressure have been related to DNB through the B&V-2 DNB correlation. The DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30.
The value corresponds to a 95 percent probability at a 95 percent confidence level that DNB vill not occur and is chosen as an appropriate margin to DNB for all operating conditions.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power 112% when the reactor coolant flov is 380,000 GPM, which is approximately 108% of design flow rate for four operating reactor i
coolant pumps.
(The minimum required sensured flov is 389,500 GP.M.)
l This curve is based on the following hot channel factors with potential l
fuel densification and fuel rod boving effects:
1 r,. 2.83, t,. 1.21, r"g. 1.65 s
The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdravn to minimum allovable ;ontrol rod withdraval, and form the core DNBR design basis.
DAVIS-BESSE, UNIT 1 B 2-1 Amendment No. 41.
A3, 41,123
SAFETY LIMITS BASES The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and account for the effects of potential fuel densification and potentiel fuel rod bow.
1.
The 1.30 DNBR limit produced by a nuclear power peaking factor of F=
2.83 or the combination of the radial peak, axial peak, and l
o pcsition of the axial peak that yields no less than a 1.30 DNBR.
I 2.
The combination of radial and axial peak that causes central fuel melting at the hot spot. The limits are 22.0 kv/ft for batch 1F and l
20.5 kv/ft for batches 6, 7 and 8.
l Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.
The Lpecified flow rates for the two curves of Figure 2.1-2 correspond to the analyret minimum flow rates with four pumps and three pumps, I
respectively.
The curve of Figure 2.1-1 is the most restrictive of all possible rea; tor coolant pump-maximum thermal power combinations shown in BASES
, Figure 2.1.
The curves of BASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to +22%,
whichever condition is more restrictive. These curves include the potential effects of fuel rod bov and fuel densification.
I
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DAVIS-BESSE, UNIT 1 B 2-2 Amendment No. AA, AS, A5, A1, A0,123
SAFETY LIMI'tS BASES For the curve of BASES Figure 2.1 a pressure-temperature point above and to the lef t of the curve vou,ld result in a DNBR greater than 1.30 or a local quality at the poitit of minimus DNBR less than +22% for that particular reactor coolant pump situation.
The 1.30 DNDR curve for three pump operation is less restrictive than the four pump curve.
2.1.3 REACTOR COOLAFT SYSTEM PRESSURE 4
The restriction of this Safety Limit protects the integrity of the 1
Reactor Coolant System from overpressurisation and thereby prevents the release of radicnuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressuriser are designed to Section III of the ASME Boiler and Pressure Vessel Code which permits a maximum I
transient pressure of 110%, 2750 psig, of design pressure. The Reactor Coolant Systes piping, valves and fittings, are designed to ANSI B 31.7, 1968 Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component design pressure.
therefore consistent with the design criteria and associated codeThe Safety Limi requirements.
i The entire Reactor Coolant System is hydrotested at 3125 psig. 125%
of design pressure, to demonstrate integrity prior to initial operation.
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l DAVIS-BESSE, UNIT 1 B 2-3 Amendment No. II.
JJ. 43, 123
e e
2.2.
LIMITING SAFETY SYSTEM SE MINGS BASES 2.2.1.
REACTOR 'ROTECTION SYSTEM INSTRUMENTATION SETPO!NTS The reactor protection system instrumentation trip setpoints scecified in Table 2.2-1 are the values at which the reactor tries are set for each param-eter. The trip setcoints have been selected to ensure that the reactor core and reactor coolant systen are prevented frem exceeding their safety limits.
The snutdown bypass p.ovides for bycassing certain functions of the reactor orotection systen in order to cemit control red drive tests, zero cower *HYS-ICS TESTS and certain startuo and snutdown procedures.
The purpose of tne shutdown eypass high eressure trip is to prevent nomal coeration with snut-down bypass activated.
This hign cressure trip setcoint is lower tnan the nomal low pressure trip setcoint so tnat the reactor must ce triccec before the bytats is initiated.
The high flux trip set:oint of <5.C'.
- revents ary significant reactor power frem being erecuced.
Suf ficienI natural circula-tien would be available to renove 5.C' of RATED TFEMAL POWER if ncne of tne reactor coolant cun:s were ocerating.
1 1
Manual Reacter Trio The manual reactor trio is a redundant channel to the automatic reactor orotec-tien system instrurentation chanteels anc provices ranual reactor tric cacacil-ity.
Hien Flux A nigh flux trip at
- gh power level (neutron flux) Orovices react: P core : -
tection against rea vity excursions which are too racid to te cettectec Oy tencerature and pressure protective circuitry.
Durir; nomal station oceration, reactor tri: i s initiate: *nen -ae react:-
cower level reacnes 104.g4 of ratec ower.
Due to transient overs o:t.
es:
I calance, anc irstrument errors, tne maxinua actual :ower at wnics a tri:
would ce actuatec cculd me 112*., wnicn was use: in tee safety analysis.
0 AVIS-BESSE, U!;IT 1 Aneneent ::o. J3,61 4
!.IMITING FAFETY SYSTEM SETTINGS BAS 85
)
RC High Temperature The RC high temperature trip i 618'F prevents the reactor outlet s
temperature from exceeding the design limits and acts as a backup trip for all power excursion transients.
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_ Flux -- 6 Flux /Flov The power level trip setpoint produced by the reactor coolant system flov is based on a flux-to-flow ratio which has been established to accommodate flov decreasing transients from high power where protection is not provided by the high flux / number of reactor coolant pumps on trips.
The power level trip setpoint produced by the power-to-flov ratio presides both high power level and lov flow protection in the event the reactor power level increases or the reactor coolant flow rate d6 creases.
The power level setpoint produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation.
For every flov rate there is a maximum permissible power level, and for every power level there is a minimum permissible lov flow rate.
Examples of typical power level and lov flow rate combinations for the pump situations of Table 2.2-1 that would result in a trip ara as follows:
1.
Tri) vould occur when four reactor coolnt pumps are operating if povur is 108.0% and reactor coolant flov rate is 100% of full flov rate, or flov rate is 92.59% of full flov rate and power level is 1001.
\\
2*
Trip would occur when three reactor coolant pumps are operating if power is 80.68% anc reactor coolant flow rate is 74.7% of full flov i
rate, or flow rate is 69.44% of full flow rate and power is 75%.
Note that the value of 80.6% in Figure 2.2-1 was truncated from the
{
calculated value of 80.68%.
For safety calculations the instrumentation errors for the power level were used.
Full flow rate in the above two examples is defined as the flow calculated by the heat balance at 100% power. At the time of the calibration the R::S flow vill be greater than or equal to the value in Tame 3.2-2.
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DAVIS-BESSE. UNIT 1 3 2-5 Amendment No. II/.
3J, 45 A1. W. 123
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LIMI?!N3 SAFETY SYSTEM SEMINGS BASES The AX!AL POWER IMSALAtCE bouncaries are estaelished in order to crevent reac-ter themal limits fecrn being exceeded. These themal limits are either power ceaking kW/ft lir'its or Df0R limits. The AX!AL POWER IMBALAfCE reduces the power level trio produced by a flux-to-flew ratio such that the bounca-ries of Figure 2.21 are produced.
RC Pressure - Lew. 49en, and Pesssure Teaeerature The high and Icw tries are provided to limit tne pressure range in wnien reac-ter cceration is cemitted.
During a slow reactivity inserti:n startue a::icent from icw power or a slen
{
reactivity insertien from migh ocwer, the RO nign cressure setcoint is reacnec eefoFe the nign flux tric set:oint. Tne tric set:oint for RC high cressure, 2300 esig, has been establishec to raintain the syste1 pressure be-i Icw tne safety limit, 2750 osig, for any design transient.
The RC nicn cres-sure trip is backed up by the pressurizer cc:e safety valves for RCS over pressure protection, and is therefore set lower than the set cressare for these valves,1 2525 psig. The RC high pressure trip also backs up the high j
flux trip.
The RC low pressure,1983.1 osig, and RO cressure-tem:erature (12.60 tou:
5662.2) osig, trip setooints have been established to maintain the DN3 ratic greater than or caval to 1.30 for tnose design ac:idents that result in #
cressure reduction.
It also crevents react:r c:eration at cressures tele =
- ne valid range of OfS correlation linits, :retectiec against Dra.
l Hien Flumnumeer of Reacter Ceelant Peres On In conjunction with the f1'.x -
- 1. flux / flow tri: :ne nign flux /nu :er :' rea:-
ter r: Slant pun:s on trie :revents the ninimum : ore OfaR fren cecreasim; belew 1.30 dy triccing the reactor cue to tne less of reactor coelant I
cuno(s).
The puno nonitors also restri:t tre cower level for tne nuneer :f l
cures in c:eration.
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5 I-i DAVIS-BE35E. U i!T 1 Anend cr.
"o. 12. H. 22.61
- e t
1 L,IMITING $AFETY SYSTEM $(TTINGS i
BASFS 4
Containment Hieh Pressure l
a i
.The Containment High Pressure Trip Setpoint < 4 psig, provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-l 1
coolant accident, even in the absence of a RC Low Press: ire trip.
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i BAVIS-BESSE UNIT 1 827 l
1 i
0 Bases Figure 2.1 Pressure / Temperature Limits at Maximum Allowable Power for Minimum DNBR 2300 i
2200 (636.3.2159.8'/
(633.4,2129.8)
/
2100
/
/
4 Pump
/
S r/
E 2000 ACCEPTABLE
/
OPERATION
/
/ (UNACCEPTABLE 625.7,1959.8) c-(621.4.1929.8) 1900
/
/
OPERATION
/
7 -
3 Pu p
/
1800 j
(608.2,
!(614.3.1759.8) 1729.8) 1700 I
i i
1 595 605 615 625 635 645 Reactor Outlet Temperature. 'T Re"quireo Measured Flow to ensure Pures Flew, gem Power Comoliance, gem i
4 380,000 112%
399,500 3
263,860 90.51 290.957 t
DAVIS-BESSE, UNIT 1 B 2-8 Amendment No //, if, //,
W. 123
REACTIVITY CONTROL SYSTEMS BORIC ACIO PUMPS - OPERATING LIMITING CONDITION FOR OPERAT!0N 3.1.2.7 nt least one boric acid pump in the boren injection flow path required by Specifict. tion 3.1.2.2a shall be OPERABLE and capable of being powered from an OPERA 8tE essential bus.
APPLICABILITY: MODES 1. 2. 3 and 4.
ACTION:
With no boric acid pump OPERABLE. restore at least one boric acid pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDSY and borated to a SHUTDOWN MARGIN equivalent to li ak/k at 200'F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least one boric acid pump to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, l
SURVEILLANCE '4EOUIREMENTS 4.1.2.7 In addition to the Surveillance Requirements of Specification 4.0.5. at least ene boric acid pump shall be demonstrated OPERABLE at least once per 316vs by:
l a.
Starting (unlessalreadyoperating)thepumpfromthecontrol
- room, b.
Verifying that the pump develops at least 931 of the discharge pressure for the applicable flow rate as detemined from the manufacturer's Pumo Perfomance Curve at a discharge pressure t 65 psig.
c.
Verifying pamp operation for at least 15 minutes, d.
Verifying that the pump is aligned to r eceive electrical power from an OPERABLE essential bus.
DAY!S-BESSE. UNIT 1 3/4 1 13
REACTIVITY CONTROL SYSTEMS BORATED VATER SOURCES - SHUTDOVN LIMITING CONDITION FOR OPERATION 3.1.2.8 As a minimum, one of the following borated vater sources sbs11 be OPERABLE:
a.
A boric acid addition system with:
1.
A minimum available borated water volume of 600 gallons, l
2.
Between 7875 and 13,125 ppm of boron, and 3.
A minimum solution temperature of 105'F.
b.
The borated vater storage tank (BVST) with 1.
A minimum available borated water volume of 3,000 gallons, l
2.
A minimum boron concentration of 1800 ppm, and 3.
A minimum solution temperature of 35'F.
APPLICABILITY:
HODES 5 and 6.
ACTION:
Vith no borated water sources OPERABLE, suspend all operations involving CORE ALTERATION or positive reactivity changes until at least one borated vater source is restored to OPERABLE status.
SURVEILLANCE REQUIREMENTS 4.1.2.8 The above required borated water source shall be demonstrated OPERABLE:
l a.
At least once per 7 days by:
1.
Verifying the boron concentration of the vater, 2.
Verifying the available borated vater volume of the l
f source, and 1
DAVIS-BESSE, UNIT 1 3/4 1-14 AmendmentNo.p/,123
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I REACTIVITY CONTROL SYSTEMS l
l SURVE!LLANCE REQUIREMENTS (Continued) 3.
Verifying the boric acid addition system solution tempera-ture when it is the source of borated water.
b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the BWST temperature when it is the source of borated water and the outside air
~
temperature is < 35'F.
l l
l l
l DAVIS-BESSE. UNIT 1 3/4 1-15 t
e Figure 3.1-1 Minimum Boric Acid Tank Available Volume as Function of Stored Boric Acid Concentration -- Davis-Besse 1 8500 8000 l
l '
5" 7500
\\
ACCEPTABLE 5
7000 g
OPERATION 6500
_o 6000 3
5500 1
5 UNACCEPTABLE
\\
a3 5000 OPERATION h
I l
4500 E
4000 3500 l
l 1
3000 7000 8000 9000 10,000 11,000 12,000 13,000 14,000 Concentration of Boric Acid Solution, ppm B DAVIS-BESSE, UNIT 1 3/4 1-16 Amendment No. W 123
REACTIVITY CONTROL SYSTEMS BORATED VATER SOURCES - OPERATING LIMITING CONDTTION FOR OPERATION 3.1.2.9 Each of the in11oving borated water sources shall be OPERABLE:
s.
The boric acid addition system with:
1.
A minimum available borated water volume in accordance l
with Figure 3.1-1, 2.
Between 7875 and 13,125 ppe of boron, and 3.
A minimum solution temperature of 105'F.
b.
The borated water storage tank (BVST) with:
1.
An avsilable borated vater volume of between 482,778 and l
550,000 gallons, 2.
Between 1800 and 2200 ppa of boron, and 3.
A minimum solution temperature of 35'F.
APPLICABILITY:
MODES 1, 2, 3 and 4.
. ACTION:
I Vith the boric acid addition system inoperable, restore the storage
(
a.
system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least MOT STANDBY and borated to a SHtTIDOVN MARGIN equivalent to 1% Ak/h at 200'F vithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> restore the boric ecid addition system to OPERABf,E status within the next 7 days or be in COLD SM11TDOV'1 vithin the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, b.
Vith the borated water storage tank inoperable, restore. the tank to OPERABLE status within one hour or be in at least NOT STANDBY vithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOVN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
DAVIS-BESSE, UNIT 1 3/4 1-17 Amendment No. M.
(7, 123
REACTIVITY C0h* TROL SYSTEMS SURVEILLANCE REQUIREMENTS l
i 4.1.2.9 Eac!t borated water source shall be demonstrated OPERABLE:
I a.
At leaFt once Per 7 days by:
i r
1.
Verifying the boron concentration in each water source, 2.
Verifying the available borated water volume of each water l
sourec, and j
3.
Verifying the boric acid addition system solution temperature.
f b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the BVST temperature when the outside air temperature is < 35'F.
f r
1 I
l L
i I
i DAVIS-BESSE, UNIT 1 3/4 1 18 Amendment No. 123 l
f 1
N EACTIVITYCONTROLSYSTEMS SAFETY ROD INSERTION LIMIT
]
LIMITING CONDITION TOR OPERATION 3.1.3.5 All safety rods shall be fully withdrawn.
APPLICABILITY:
1* and 2*f.
ACTION:
With a maximum of one safety rod not fully withdrawn, except for sur-veillance testing pursuant to Specificatien 4.1.3.1.2, within one hour either:
a.
Ft ly withdraw the rod or b.
Declare the rod to be inoperable and apply Specideation 3.1.3.1.
SURVE!LLANCE REQUIREMENTS 4.1.3.5 Each safety red shall be detemined to be fully withdrawn:
a.
Within 15 minutes prior to withdrawal of any regulating rod during an approach to reactor criticality, b.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereaf ter.
- 5ee 5pecial Test Exception 3.10.1 and 3.10.2.
fWith K,ff g 1,0.
DAVIS SESSE, UNIT 1 3/4 1 25
I l
l I
REACTIVITY CONTROL SYST".MS REGULATING ROD INSERTION
7,$,$
LIMITING CONDITION POR OPERATION l
l 3.1.3.6 The regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2a, and.2b, 3.13a, and -3b.
group overlap of 23:51 shall be maintained between sequential withdrawn l
A rod t
j groups 5, 6 and 7.
APPLICABILITY: MODES 1* and 2*t.
l ACTION f
Vith the regulating red groups inserted beyond the above insertion l
limits (in a region other than acceptable operation), or with any group sequence or overlap outside the specified limits, except for r
1 surveillance testing pursuant to Specification 4.1.3.1.2, either:
i Restore the regulating groups to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, f
a.
or l
b.
Reduce THERMAL POVER to less than or equal to that fraction lif RATED i
l THERMAL POVER vhich is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or Be in at least HOT STANDBY vithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
L c.
I NOTE:
If in unacceptable region, also see Section 3/4.1.1.1.
i E
i i
i l
I t
I i
- See Special test Exception 3.10.1 and 3.10.2.
i 4Vith k,gg > 1.0.
j DAVIS-BESSE, UNIT 1 3/4 1-26 Amendment No. 11
(
77, WY, 42, 45 W. 61, 50, 125 l
\\
REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating group shall be determined to be within the insertion, sequence and overlap limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except when:
a.
The regulating rod insertie, limit alarm is inoperable, then verify the groups to be within the insertion limits at least
-.o per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; b.
TF.e control rod drive sequence alarm is inoperable, then verify the groups to be within the sequence and overlap limits at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i 0
1 DAVIS,BESSE, UNIT 1 3/4127
e Figure 3.1-2a Regulating Group Position Limits, O to 3251 10 EFPD, Four RC Pumps --
Davis-Besse 1, Cycle 6 (300,102)
(258,102) 100 - Power Level (270,'102)
Cutoff = 100%
(270,92)
SHUTDOWN E
MARGIN g 80 LIMIT (250,80) c.
a bw UNACCEPTABLE P
OPERATION OPERATION 60 g
RESTRICTED E
(170,50)
(180,50) mo
$ 40 e
f
{
(128,28.5)
Y E 20 '-
/
ACCEPTABLE j'g OPERATION W (0.0,5.0) 0 t
t 8
0 100 200 300 Rod index (% Withdrawn)
GR 5 t
_ _.t.,
O 75 100 GR 6 i
i e
i 0
25 75 100 GR 7 I O
25 100 nAVIS-BESSE, UNIT 1 3/4 1-28 Amendment No. J4. 43, 44,
{
AA,pp.123 l
Figure 3.1-2b Regulating Group Postion Limits After 3251 10 EFPD, Four RC pumps, APSRs Withdrawn -- Davis-Besse 1, Cycle 6 (266,102)
(300,102) 100 Power Level (270,l02)
Cutoff = 100%
SHUTDOWN
(
MARGIN
^
)k (250,80) 5 80 LIMIT 8
c.
a OPERATION E
RESTRICTED 60
'dACCEPTABLE 8
OPERATION 3
(176,50)
(180,50) o j
40 E
E
{
(136,28.5) r2 20 ACCEPTABLE OPERATIOh (0.0,5.0) 0
-g 0
100 200 300 Rod Index (% Wither 3wn)
GR 5 t.--,, -.
0 75 100 GR 6 L__
0 25 75 100 GR 7 '
0 25 100 DAVIS-BESSE, UNIT 1 3/4 1-28a Amendment No. Ah, M, /V2, Ah, Ai, A0,123
4 e
e 4
Figure 3.1-2c Regulating Group Position Limits, 200 10 to 330 10 EFPD, Four 3
2 RC Pumps - Davis-Besse 1, Cycle 5 P
l DELETED b
i f
l 5
i i
r 1
d I
r I
i t
I l
i 1
DAVIS-BESSE, UNIT 1 3/4 1-28b Amendment No. 47, 4N, 6V, 6W, 8(4 123 l
l
Figure 3.1-2d Regulating Group Position Limits, 330210 to 390 10 EFPD, Four RC Pumps, 3
APSRs Vithdravn--Davis-Besse 1, Cycle 5 DELETED l
F t
i i-b r
1 l
L
(
I i
[
DAVIS-BESSE, UNIT 1 3/4 1-28c Amendment No. 45, 61, de, 80, 123
, -..... -, ~ -,.,,
-. - - - - ~., - - - -.. -,. _ - - _. - - - -... _ - - - - -, - - -.
Figure 3.1-2e Del eted l
l l
DAVIS-BESSE, UNIT 1 3/4 1-28d Amendment No. $5, $1, pp, go r
i I
I l
)
}
'~
c' Figure 3.1-3a Regulating Group Position Limits, O to 325 10 EFPD, Three RC Pumps --
Davis-Besse 1, Cycle 6 100 S
80 (258,77)
(300,77) 3 SHUTDOWN f
MARGIN
)(270,69.5)
LIMIT w
){
(250,60.5) 5 60 S
E
[
UNACCEPTABLE OPERATION OPERATION RESTRICTED
('170,38)
T 40 8
(180,38) tS b
j 20 (128,21.8)
ACCEPTABLE OPERATION 0
O 100 200 300 Rod Index (% Withdrawn)
GR 5 '
O 75 100 GR 6 i
i 0
25 75 100 GR 7 '
O 25 100 DAVIS-BESSE, UNIT 1 3/4 1-29 Amendment No. J/, //, //,
pp,p),pp,123
~...;.
Figure 3.1-3b Regulating Group Position Limits After 325 10 EFPD, Three RC Pumps, APSRs Withdrawn -- Davis-Besse 1, Cycle 6 100 p 2
80 (266,77)
(300,77) y SHUTDOWN MARGIN (270,77) n-g UNACCEPTABLE LIMIT (270,69.5)
=
OPERATION 0
l250,60.5)
S 2"
OPERATION (176,38)
RESTRICTED 40 5
(180,38)
E 2
u*g 20 (136,21.8) c.
eCCEPTABLE OPERATION O
0 100 200 300 Rod Index (% Withdrawn)
GR 5 L
8 0
- /5 100 GR 6 i 0
25 75 100 GR 7 '
O 2S 100 l
1 l
DAVIS-BESSE, UNIT 1 3/4 1-29a Amandment No. ll, $3, 44, 97, f),79 123 1
O e
Figure 3.1-3c Regulating Group Position Limits, 200:10 to 330g10 EFPD, Three RC Pumps--Davis-Besse 1, Cycle 5 i
DELETED
\\
4 i
I 1
t l
i t
i l
t i
t i
l t
l l
i i
DAVIS-BESSE, UNIT 1 3/4 1-29b Amendment No. YY, l
3N, 47, 4"/, WY, t
i A9, AO,123
?
Figure 3.1-3d l
Regulatory Group Position Limits, t
330 10 to 390 10 EFPD, Three RC Pumps, 2
APSRs Vithdravn--Davis-Besse 1, Cycle 5 t
DELETED i
~
t 4
L I
t i
1 f
DAVIS-BESSE, UNIT 1 3/4 1-29c Amendment No. 11, t
11, 45, 61, M, no 123 s
l
Figure 3.1-4 Control Rod Core Locations and Group Assignments --
Davis-Besse 1, Cycle 6 X
/
N I
A 3
4 6
4 C
2 5
5 2
D 7
8 7
8 7
E 2
5 5
2 F
4 8
6 3
6 8
4 G
5 1
1 5
H W-6 7
3 4
3 7
6
-Y K
5 1
1 5
L 4
8 6
3 6
8 4
M 2
5 5
2 N
l 7
8 7
8 7
0 l
2 5
5 2
P l
l l
4 6
4 R
l l
I I
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 Groue No. of Rods Function 1
4 Safety 2
8 Safety 3
4 Safety X
Group Number 4
9 Safety 5
12 Control 6
8 Control 7
8 Control 8
8 APSRs Total 61 i
l l
l DAVIS-BESSE, UNIT 1 3/4 1-31 Amendment No. AA. A. f.
7)'. W. 123
INTENTIONALLY LEFT BLANK DAVIS-BESSE, UNIT 1 3/4 1-32 Am n a nt No. M
REACTIVITY CONTROL SYSTEMS XENON REACTIVITY 4
LIMITING CONDITION FOR OPERATION 3.1.3.8 THERMAL POWER shall not be increased above the power level cutoff specified in Figure 3.1-2 unless one of the following conditions is satisfied:
Xenon reactivity is within 10 percent of the equilibrium a.
value for RATED THERMAL POWER and is approaching stability, or b.
THERMAL POWER has been within a range of 67 to 92 percent of RATED THERMAL POWER for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the soluble poison control mode, excluding xenon free start-ups.
APPLICABILITY: MODE 1.
ACTION:
With the requirements of the above specification not satisfied, reduce THERMAL POWER to less than or equal to the power level cutoff within 15 minutes.
SURVE!LLANCE REQUIREMENTS Xenon reactivity shall be determined to be within 10% of the 4.1.3.8 equilibrium value for RATED THERMAL POWER and to be approaching stability or it shall be determined that the THERMAL POWER has >een in the range of 87 to 92% of RATED THERMAL POWEF. for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, prior to increasing THERMAL POWER above the power level clitoff.
DAY!S-BESSE, UNIT 1 3/4 1-33
REACTIVITY CONTROL SYSTEMS AXIAL POVER SRAPING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.9 The axial power shaping rod group shall be limited in physical insertion as shown on Figures 3.1-Sa, -5b, and -Sc.
l APPLICABILITY: H0 DES 1 and 2*.
ACTION Vith the axial power shaping rod group outsi., the above insertion limits, either:
Restore the axial power shaping rod group to vithin the limits a.
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or b.
Reduce THERHAL POVER to less than or equal to that fraction of RATED THERHAL POVER vhich is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or c.
Be in at least HOT STANDBY vithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.9 The position of the axial power shaping rod group shall be determined to be within the insertion limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except when the axial power shaping rod insertion limit alarm is inoperable, then verify the group to be within the insertion limit at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- Vith K,gg > 1.0.
DAVIS-BESSE. UNIT 1 3/4 1-34 Amendment No. JJ.
47, y). $J, ff.
SV, 123
O e
Figure 3.1-Sa APSR Position Limits, O to 3251 10 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 6 RESTRICTED REGION
_^
100'- (0,102)
(100,102) 80 2
w 8
a.
a h
60 E
PERMISSIBLE
[
OPERATING REGION w
E a:
so 40 e
8 t
S.
bg 20 c.
0 i
i 0
10 20 30 40 50 60 70 80 90 100 APSR Position (% Withdrawn) 1 DAVIS-BESSE, UNIT 1 3/4 1-35 Amendment No. )). //. f).
W. 123
Figure 3.1-5b APSR Position Limits After 3251 10 EFPD, Three or Four RC Pumps, APSRs Withdrawn --
Davis-Besse 1, Cycle 6 t
100 E
80 a
k APSR INSERTION NOT ALLOWED i
60 A
IN THIS TIME INTERVAL S
E o
40 c
0 5
c.
T 20
~
~
)
0 e
i f
i i
i i
i i
0 10 20 30 40 50 60 70 80 90 *00 l
APSR Position (% Withdrawn) l t
DAVIS-BESSE. UNIT 1 3/4 1-36 Amendment No. 77. #2. 45.
W. 57, 123
o' Figure 3.1-5c APSR Position Limits, O to 3251 10 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 6 100 80 RESTRICTED REGION g
7 g
'(0.77)
(100.77) c.
J B
E 60 S
Y ERMISSIBLE 40 5
OPERATING REGION O
b 8
a-20 0,-'
' =-= 1 0
10 20 30 40
.c,0 60 7')
80 90 100 d
APSR Position (% Withdfiwn)
DAVIS-BESSE, UNIT 1 3/4 1-37 Amendment No. W. MI. II.
W. W.
?
e
/
P L
Figure 3.1-5d f
APSR Position Limits, 330t10 to 390 10 EFPD, Three or Four RC Pumps, APSRs Vithdravn--Davis-Besse 1, Cycle 5 l
t I
l DELETED i
l I
i 1
t l
i l
1 l
i 1
i l
i.
j i
i j
-l 1
4 i
i i
)
i DAVIS-BESSE, UNIT 1 3/4 1-38 Amendment No. M',
Ar, As, Ai, A9, Ao,123 l
l
.m.,
,,.r m _.,- - __ _.
]
l l
Figure 3.1-5e APSR Position Limits, O to 25+10/-0 EFPD, Three RC Pumps--Davis-Besse 1, Cycle 5 DELETED i
i 1
j 1
DAVIS-BESSE, UNIT 1 3/4 1-39 Amendment No. 42, 17 9). f?> P9: 123
s Figure 3.1-5f
~,
i APSR Position Limits, 25+10/-0 to 200 10 3
EFPD, Three RC Pumps--Davis-Besse 1 Cycle 5 r
DELETED i
i I
l i
l i
r i
I l
t I
t I
i i
i DAVIS-BESSE, UNIT 1 3/4 1-40 Amendment No. AA/,
[
6!, H, 80,123 I
s Figure 3.1-5g APSR Position Limits, 200 10 to 330g10 EFPD, t
Three RC Pumps--Davis-Besse 1, Cycle 5 DELETED 1
I
?
t L
i t
I DAVIS-BESSE, UNIT 1 3/4 1-41 Amendment No. 47.
W. W W,123
Figure 3.1-Sh Deleted DAVIS-BESSE. Uti!T 1 3/4 1-42 Amendment No. AJ. fl. 89.
80
3/4.2 POVER DISTRIBtTTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POVER IMBALANCE shall be maintained within the limits shown on Pigures 3.2-1 and 3.2-2.
l APPLICABILITY: MODE 1 above 40% of RATED THERMAL POVER.*
ACTION Vith AXIAL POVER IMBALANCE exceeding the limits specified above, either:
Restore the AXIAL POVER IMBALANCE to within its limits within 15 a.
minutes, or b.
Vithin one hour reduce power until imbalance limits are met or to 40% of RATED THERMAL POVER or less.
SURVEILLANCE REQUIREMENTS 4.2.1.
The AXIAL POVER IMBALANCE shall be determined to be within limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when above 40% of RATED THERMAL POVER except when the AXIAL POVER IMBALANCE alarm is luoperable, then niculate the AXIAL POVER IMBALMCE at.least or.ce per hour.
l l
]
j
- See Special Iest Exception 3.10.1.
DAVIS-BESSE, UNIT 1 3/4 2-1 Amendment No. JJ, M Me M e A9 W. 123
J Figure 3.2-1 AXIAL POWER IMBALANCE Limits, Four RC Pumps -- Davis-Besse 1, Cycle 6
. 110
(-20,102)
(15,102)
" 100
(-25,92) t (15,92) 90 2
(-28,80)i y-- 80
,(20,80) c.
70 w
i5
.60 8
3-- 50 o (20,50)
(-28,50) 0 w
- 40 RESTRICTED PERMISSIBLE E
- 30 REGION OPERATING $
REGION h
r 20 33 n
_1_
)
_ q n
a
_ _ - (,
n
_1
-50
-40
-30
-20
-10 C
10 20 30 40 EO AXIAL POWER IMBALANCE (*)
I f
DAVIS-BESSE, UNIT 1 3/4 2-2 Amendment No. 11, 33, 45, WY, 77, 123
~ _.
.g i
Figure 3.2-1b Axial Power Imbalance Limits, 25+10/-0 to 200+10 EFPD, Four RC Pumps, Davis-Besse 1, Cycle 5 t
n i
DELETED h
l i
i l
i l
6 i
i i
f l
i i
t I
l DAVIS-BESSE, UNIT 1 3/4 2-2a Amendment No. If, 77e W, 47, (Y, W. 123
g Figure 3.2-1c Axial Power Imbalance Limits, 200310 to 330;10 EFPD, Four RC Pumps--Davis-Besse 1, Cycle 5 i
DELETED i
b t
1 i
4 i
1 i
l f
.i.
r i
P i
F t
i i
1 f
4 DAVIS-RESSE, UNIT 1 3/4 2-2b Amendment No. A2, M. 9. #7 79, 123 i
1
,,.,------,,,,....-,....__.-,_,..,..,...n.,.,----
I
s Figure 3.2-1d Axial Power Imbalance Limits, 330 10 to 390 10 EFPD, Four RC 3
2 Pumps, APSRs Vithdravn--
Davis-Besse 1, Cycle 5 l
DELETED i
i l
t I
a 4
I 1
h I
a l
- i
.i l.
1 e
a V
i l
i l
I DAVIS-BESSE, UNIT 1 3/4 2-2c Amendment No. 43, r
fJe $7, SD,123
Figure 3.2.le Deleted i
l I
l 1
l l
1 l
l 1
DAVIS-BESSE. UNIT 1 3/'* a'-2d Mendaent No. J), if, $7, 80
.w l
s figure 3.2 2 AXIAL POWER IMBALANCE Limits, Three RC Pumps -- Davis-Besse 1, Cycle 6
. 113 100 90
(-1C 77) -
- (11.25.77)
(-18.75,69.51 70 e (11.25,69.5) 2
(-21,60.5)'q
$-- 60
>(15,60.5) c.
J h-- 50 w
~ 40 y~
g o (.5.38)
(-21,38) o C
%-- 30 5 RESTRICTED y
ye REGION g
o
- 20 ygg g-G" g-- 10 h 2
2 l
I I
I g
i a d i
I
-50
-40
-30
-20
-10 0
10 20 30 40 50 AX1AL POWER IMBALANCE (%)
DAVIS-EESSE, UNIT 1 2-3 Amendment No. AA. AA, AA.
M. A0,123
3,,
Figure 3.2-2b f
5 Axial Power Imbalance Limits, 25+10/-0 to 200 10 EFFD, Three RC r
Pumps--Davis-Besse 1, Cycle 5 DELETED
~
,4 n
1 4
iL i
i t
i r
I I
I t
L I
f I
4 I
I t
I 4
i l
t 1
I DAVIS-BESSE, UNIT 1 3/4 2-3a Amendment No. II, 71, C, 45, 6I, 80, 123 I
__,_.~_
._,. m,
?.
.c s Figure 3.2-2c Axial Power Imbalance Limits, 200 10 to 3
330 10 EFPD, Three RC Pumps--Davis-lesse le 3
Cycle 5 DELETED DAVIS-BESSE, UNIT 1 3/4 2-3b Amendment No. U,
4*/, tW 6Y, 50, 123
I Figure 3.2-2d
(
Axial Power Imbalance Lisits 33010 to t
390 10 EFPD, Three RC Pumps APSRs 2
Vithdravn--Davis-Besse 1, Cycle 5 l
l t
t DELETED I
i l
i l
h I
i t
i i
t.
\\'
I I
I l
DAVIS-BESSE, UNIT 1 3/4 2-3c Amendment No. Ir7, W. W, W 123
.s POVER DISTRIBtTTION LIMITS QUADRANT POWER TILT LIMITING CONDITION FOR OPERATION 3.2.4 THE QUADR.'.NT POWER TILT shall not exceed the Steady State Limit of Table 3.2-1.
l APPLICABILITY: MODE 1 above 15% of RATED THERMAL POVER.*
ACTION:
Vith the QUADRANT POVER TILT determined to exceed the Steady a.
State Limit but less than or equal to the Transient Limit of Table 3.2-1.
l l.
Vithin 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
a)
Either reduce the QUADRANT POVER TILT to within its Steady State Limit, or b)
Reduce TilERMAL POVER so as not to exceed THERMAL POWER, including power level cutoff, allovable for the reactor coolant pump combination less at least 2%
for each 1% of QUADRANT POVER TILT in excess of the Steady State Limit and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the High Flux Trip Setpoint and the Flux-A Flux-Flov Trip Setpoint at least 2% for each 1% of QUADRANT POVER TILT in excess of the Steady State Limit.
2.
Verify that the OUADRANT POVER TILT is within its Steady State Limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the Steady State Limit or reduce THERKAL POVER to less than 60% of THERMAL POWER allovable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the High Flux Trip Setpoint to < 65.5% of THERMAL POWER allovable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4 3.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POVERI subsequent POVER OPERATION above 60% of THERNAL POVER allowable for the reactor coolant pump combination may proceed provided that the QUADRANT POVER TILT is verified within its Steady
[
State Limit at least once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or until i
verified acceptable at 95% or greater RATT.D THERMAL POVER.
l
- See Special Test Exception 3.10.1 I
DAVIS-EESSE. UNIT 1 3/4 2-9 Amendment No.123 i
POVER DISTRIBtTTION LIMITS LIMITING CONLIT3 N FOR OPERATION (Continued) b.
Vith the QUADRANT POVER TILT determined to exceed the Transient Limit but less than the Maximum Limit of Table 3.2-1, due to l
misalignment of either a safety, regulating or axial power shaping rod 1.
Reduce THERMAL POVER at least 2% for each 1% of indicated QUADRANT POVER TILT in excess of the Steady State Limit within 30 minutes.
2.
Verify that the OUADRA!n POVER TILT is within its Transient Limit u.ithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after axceeding the Transient Limit or reduce THERMAL POVER to less than 60%
of THERMAL POVER allovable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the High Flux Trip Setpoint to 3 65.5% of THERMAL POVER allovable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3.
Identify and correct the cause of the out of limit condition prlor to increasing THERMAL POVER: subsequent POVER OPERATION above 60% of THERMAL POVER allovable for the reactor coolant pump combination may proceed provided that the OUADRANT POVER TILT is verified within its Steady State 1.imit at least once per hour for 2.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or until verified acceptable at 95% or greater RATED TSERMAL POVER.
c.
Vith the OUADRAtU POVER TILT determined to exceed the Transient Limit but less than the Maximum Limit of Table 3.2-1, due to l
causes other than the misalignment of either a safety, regulat-ing or axial power shaping rod 1.
Reduce THERMAL POVER to less than 60% of THERMAL POVER allovable for the reactor coolant pump combination within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the High Flux Trip Setpont to S 65.5%
of THERMAL POVER allovable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
2.
Identify and correct the cause nf the out of limit condition prior to increasing THERMAL POVER: subsequent POVER OPERATION above 60% of THERMAL POVER allovable for the reactor coolant pump combination may proceed provided that the QUADRANT POVER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or until verified at 95% or greater RATED THERMAL POVER.
DAVIS-BESSE, UNIT 1 3/4 2-10 Amendmeat No. 123
s POVER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)
ACTIONr (Continued) i d.
Vith the OUADRANT POVER TILT determined to exceed the Maximun Limit of Table 3.2-1, reduce THERMAL POVER to < 15% of RATED l
THERMAL POVER vithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
~
SURVEILLANCE REOUIREMENTS L
4.2.4 The QUADRANI POVER TILT shall be determined to be within the limits at least once every 7 days during operation above 15% of RATED THERMAL POVER except when the QUADRART POVER TILT alarm is inoperable, then the QUADRANT POVER TILT shall be calculated at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i i
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a h
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~
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i DAVIS-BESSE, UNIT 1 3/4 2-11 Amendment No. 123 l
., _ _ ____...__ _ ______ _ _ _ _... _ _ _ _. _. _ _ _ _ _, _ _. ~. _. _ _ ~
[.-
E5 m
Table 3.2-100ADIUurr POTUt TILT Limits E
l m
Steady state Steady state E
limit for limit for U
THEltMAL TREltMAL Transient Maximum POWER 1 50%
POWER > 50%
limit limit QUADItANT POWER TILT as measured by:
Symmetrical incore detector system 6.83 4.12 10.03 20.0 Power range w
channels 4.05 1.%
6.%
20.0 5
u Minimum incore h
detector system 2.80 1.M 4.@
20.0
.g M
g.3
~
F B-
.p
.. ~
~, ---w--
w w---w, wor-m --wvmv,
- v s w w wv - w en -- -r-wmawww
-r-r-s-s------re-
~w--
,,w, v-w
-me-e--
w
s i
POWER DISTRIBUTION LIMITS
[
DNB PARAMETERS l
LIMITING CONDITION FOR OPERATION l
t 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-2.
l a.
Reactor Coolant Not Leg Temperature b.
Reactor Coolant Pressure c.
Reactor Coolant Flov Rate E
l APPLICABILIT't: MODE 1
[
ACTION:
l If parameter a or b above exceeds its lia)t. restore the parameter to I
within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POVER to less than 5%
of RATED TRLRMAL POVER vithin the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
If parameter c exceeds its limit, either:
1.
Restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or I
f 2.
Limit THERMAL POWER at least 2% belov RATED THLRMAL POVER for each
[
- 13. parameter e is outside its limit for four pump operation within i
the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or limit THERMAL POVER at least 2% belov 75% of RATED THERMAL POVER for uach 1% parameter e is outside its limit
[
for 3 pump operation within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
l L
SURVEILLANCE REQUIREMENTS I
4.2.5.1 Each of the parameters of Table 3.2-2 shall be verified to t.*
l within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
[
4.2.5.2 The Reactor Coolant System total flov rate shall be determined to be within its limit by measurement at least once per 18 months.
[
i f
DAVIS-BESSE. UNIT 1 3/4 2-13 Amendment No. AI.,123 r
I
f 5:p TABLE 3.2-2 l
N DNS MARCIN E
Required Measured Required Measured U
Parameters with Parameters with Four Reactor Three Reactor Coolan* Pumps Coolant Pumps Parameter Operating Operating Reactor Coolant Hot Leg N
Temperature T 9 10 5610 H
Reactor Coolant Pressure, psig.I I 22062.7 22058.7 II) w II y
Reactor Coolant Flow Rate, gpa 2389,500 2290,957 l
f 1
III
, t Applicable to the loop with 2 Reactor Coolant Pumps Operating.
[*
(2) Limit not applicable during either a THERNAL POWER ramp increase in excess of 5% of RATED TIERNAL
~
POVER per minute or a TEERMAL POWER step increase of greater than 10% of RATED TRERNAL POWER.
D~
-y (3) These minimum required measured flows include a flow rate uncertainty of 2.5%, and are based on an l
g-minimum of 52 lumped burnable poison rod assemblies in place in the core.
I wy I
e 4
6
TABLE 4.3-1 REACTOR PROTECTION SYSTEM INS 11t1 MENTATION SURVEILIANCE REQUIREMDITS s'
CHANNEL MODES IN WHICH y
CHANNEL CHANEEL FUNCTIONAL SURVEILLANCE g
FUNCTIONAL UNIT
_CIIECK CALIBRATION TEST REQUIRED N
J" 1.
Manual Reactor Trip N.A.
N.A.
S/U(1)
N.A.
h 2.
Iligh Flux S
D(2), and Q(7)
M 1, 2 e-a 3.
RC High T:wperature S
R H
1, 2 4.
Flux - AFlux - Flow S(4)
M(3) and Q(7,8)
M 1, 2 5.
RC Low Pressure S
R M
1, 2 l
6.
RC Ifigh Pressure S
R H
1, 2 l
7.
RC Pressure-Temperature S
R M
1, 2 y
8.
Iligh Flux / Number of Reactor Coolant Pumps On S
R M
1, 2
-a 9.
Containment High Pressure S
R M
1, 2 10.
Intermediate Range, Neutron Flux and Rate S
R(7)
S/II(5)(1) 1, 2 and*
k 11.
Source Range, Neutron Flux g,
and Rate S
R(7)
M and S/U(1)(5) 2, 3, 4 and 5 9
h 12.
Control Rod Drive Trip Breakers N.A.
N.A.
M(9) and S/U(1)(9) 1, 2 and*
[
13.
Reactor Trip Module Logic N.A.
N.A.
M 1, 2 and*
14.
Shutdown Bypass High Pressure S
R H
2**,3**,4**,5**
w' 15.
SCR Relays N.A.
N.A.
R 1,2 and
- O 5
)
J
?
TABLE 4.3-1 (Continued)
NOTATION l
(1) -
If not performed in previous 7 days.
i (2) -
Heat balance only, above 15% of RATED THERMAL POWER.
}
(3) -
Vhen THERMAL POVER [TP) is above 50% of RATED THEP. MAL POVER i
[RTP) and at a steady state, compare out-of-core measured AXIAL
{
r POWER IMBALANCE [ API,) to incore measured AXIAL POWER IMBALANCE t
[ API ) as follows:
y RTP [ API, - APl ) = Offset Error y
TP h
Recalibrate if the absolute value of the Offset Error is
) 2.5%.
1 L
(4) -
AXIAL POVER IMBALANCE and loop flov indications only.
(5) -
Verify at least one decade overlap if not verified in previous 7 days.
(7) -
Neutron detectors may be excluded from CRANNEL CALIBRATION.
i (8) -
Flov rate measurement sensors may be excluded from CHANNEL I
CALIBRATION. However, each flow measurement sensor shall be j
calibrated at least once per 18 months.
(9) -
The CHANNEL FUNCTIONAL TEST shall independently verify the I
OPERABILITY of both the undervoltage and shunt trip devices of the Reactor Trip Breakers.
l i
With any control rod drive trip breaker closed.
\\
1 When Shutdovn Bypass is actuated.
l 4
i
\\
b t
e i
i h
i d
i 1
i DAVIS-BESSE, UNIT 1 3/4 3-8 Amendment No. 47, AAE,123
v --
w
-w-s 3/4.4.
REACTOR C001#rf SYSTEM 3/4.4.1.
COOLANT LOOPS AND COOLANT CIRCULATION i
STARTUP AND POVER OPERATION LIMITING CONDITION FOR OPERATION l
3.4.1.1 Roth reactor coolant loops and both reactor coolant pumps in each loop shall be in operation.
r APPLICABILITY:
MODES 1 and 2*.
ACTION:
f f
Vith one reactor coolant pump not in operation, STARTUP and POVER a.
OPERATION may be initiated and may proceed provided TBCRMAL POVER is i
restricted to less than 80.6% of RATED THERMAL POVER and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the setpoints fot the following trips have been reduced in accordance with Specification 2.2.1 for operation with three reactor i
coolant pumps operating:
1.
High Flux 2.
Flux-AFlux-Flov i
I SURVEILLANCE REQUIREMENTS 4
4 4.a.l.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12
(
hours.
I l
4.4.1.2 The Reactor Protection System trip setpoints for the i
instrumentation channels specified in the ACTION statement above shall i
be verified to be in accordance with Specification 2.2.1 for the applicable number of reactor coolant pumps operating either:
t Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> af ter svitching to a three pump combinatio.. if the l
{
a.
switch is made while operating, or b.
Prior to reactor criticality if the switch is made while shutdovn.
f
[
- See Special test Exception 3.10.3.
1 DAVIS-BESSE. UNIT 1 3/4 4-1 Amendment No. #,
c
)).)$e Me 59 123 i
r.--...-.
.. -. - -.yy
-.r-,,,--ve-.w.-
-+ ---. - -,, - - -
3 /t.. /. FIACTOR C00LAS.7 STS EM SHITTDOW AC HOT STANDBT l
gTINC COCITION TOR OPERATION 3.4.1.2 a.
At least two of the coolant loops listed below shall be OPERA 3LE:
1.
Reactor Coolant Loop 1 and its associated staan generator.
2.
Reactor Coolant Loop 2 and its associated staan generator, 3.
Decay Esat Removal Imop 1,*
4.
Decay Heat P.esoval Loop 2.*
b.
At least one of the above coolant loops shall be in operation.**
c.
Not more than one decay heat removal pump may be operated with the sole suction path through DE-11 and DE-12 unless the control power has baan removed from the DE-11*and DE-12 valve operator, or manual valves DE-21 and DE-23 are opened.
d.
The previsions of Speci.fications 3.0.3 and 3.0,4 are not applicable.
APPLICA3ILITY: MODES 3, 4 and 5 AC"'ICN :
a.
With less than the above required coolant loops 0FEIA31.E.
immediataly initiata corrar.tive action to return the required coolant loops to OPERA 3LI status as soon as possible, or be in COLD SEUIDOW within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b.
With none of the above required coolant loops in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediataly initis.te corrective action to return the required coolant loop to operation.
- The nomal or emergency power source may be inoperable in MODE 5.
This loop may not be selected in MODE 3 unless the primary side temperature and l
pressure are within the decaw heat removal system's ocsign conditions.
- The decay heat removal pumps may be de-energiztd for up to I hour provided (1) no operations are permitted that would.:ause dilution of I
the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature, i
DAVIS-BESSE UNIT 1 3/4 4-2 Amendment ho. /.
- 8. (8, la. 92
m EMERGENCY CORE C0OLING SYSTEMS ECCS SUBSYSTEMS - 7;V_, < 280*F LIMITING CONDITION FOR OPERATION 3.5.3 As a minissa, one ECC3 subsystem comprised of the following shall be OPERA 8LE:
a.
One OPERABLE decay heat (DH) pump, b.
One OPERABLE DH cooler, and c.
An OPERABLE flow path capable of taking suction from the borated water storage tank (BNST) and manually transferring suction to the r
containment emergency sump during the recirculation phase of operation.
APPLICABILITY: MDE 4.
ACTION:
a.
With no ECCS subsystem OPERABLE because of the inoperability of the DH pump, the DH cooler or the flow path from the BWST, restore at least one ECCS subsystem to OPERABLE status within one h0ur or uaintain the Reactor Coolart System Tav0 less than 280'F by use of alternate heat removal methods.
b.
In the event the ECCS is actuated and injects water into the l
reactor coolant system, a Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date, t
l SURVEILLANCE REOUTREMENT3 4.5.3 The ECCS subsystems shall be de enstrated OPERAILE per the applicable Surveillance Requirements of 4.5.2.,
DAVIS-BESSE, UNIT 1 3/4 5-6 Amendment No. E, 57 i
i i
t EMERGENCY CORE COOLING SYSTEMS BORATED VATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The borated water storage tank (BUST) shall be OPERABLE with:
a.
An available borated water volume of between 482,778 l
and 550,000 gallons, b.
Between 1800 and 2200 ppa of boron, and c.
A minteum vater temperature of 35't.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
Vith the borated water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least 50T STANDBY vithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHVfDOVN vithin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.5.4 The BVST shall be demonstrated OPERABLE:
a.
At least once per 7 days by:
1.
Verifying the available borated water volume in the tank, 2.
Verifying the boron concentration of the water.
b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the water temperature when outside air temperature <35'F.
DAVIS-BESSE, UNIT 1 3/4 5-7 Amendment No. 75, 123
b 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 80 RATION CONTROL 3/4.1.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
During Modes 1 end 2 the SHUTDOWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the insertion limits.
SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion RCS boron concentration and RCS T"Na.
The mest restrictive condition occurs at EOL, with T at no d operating temperature.
The SHUTOOWN MARGIN required U 9 consistent with FSAR safety analysis assumptions.
i 3/4.1.1.2 BORON DILUTION A minimum flow rate of at least 2800 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual through the Reactor Coolant System in the core during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 2800 GPM will circulate at equivalent Reactor Coolant System volume of 12.110 cubic feet in approximately 30 minutes.
The reactivity change rate associated with boren concentration reduction will be within i
the capability for operator recognition and control.
3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT l
The limitations on moderator temperature coefficient (MTC) are i
provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance require-ment for measurement of the MTC each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to a
the reduction in RCS boron concentration associated with fuel burnup.
The confirmation that the measured MTC value is within its limit provides assurance that the coefficient will be maintained within acceptable values throughout each fuel cycle.
l DAVIS BESSE, UNIT 1 B 3/4 1-1 i
I REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MINIMUM TEMFERATURE FOR CRITICA1.ITY This specification ensures that the reactor vill not be made critical with the reactor coolant system average temperature less than 525'F.
This limitation is required to ensure (1) the moderator temperature coefficient is within its analysed temperature range, (2) the protective instrumentation is within its normal operating range. (3) the pressuriser is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor pressure vessel is above its minimus RT temperature.
g 3/4.1.2.
50RATIONSYSTE'i{
The boron injection system ensures th4t negative reactivity contrcl is available during each snde of facility operation.
The components required to perform thir function include (1) borated water sources, (2) makeup or DHR pumps, (3) saps ste flow paths (4) boric acid pumps, (5) associated heat tracing systems, and (6) an emergency power supply from operable emergency busses.
Vith the RCS average temperature above 200'F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or co:;rective action may be completed without undue risk to overall facility safety from injection system failure.- during the repair period.
The boration capability of either system is sufficient to provide a SgUTDOVN MARGIN from all operating conditions of 1.0% Ak/k after menon decay and cooldovn to 200'F. The maximum boration capability requirement occurs from full power equilibrium xenon conditions and requires the equivalent of either 7373 gallons of 8742 ppe borated water from the boric acid storage tanks or 52,726 gallons of 1800 ppe borated water from the borated water storage tank.
The requirement for a minimum available volume of 482,778 gallons of l
borated water in the borated water storage tank ensures the capability for borating the RCS to the desired level. The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.41 therefore, the larger volume of borated water is specified.
Vith the RCS temperature belov 200'F, one injection system is acceptable without single failure consideration on the basis of the DAVIS-BES!?, UNIT 1 B 3/4 1-2 Amendment No.A1, 33, Ah, M, A1,123
t i
REACTIVITY CONTRnL SYSTEMS BASES 3/4.1.2 BORATION SYSTEMS (Continued) stable reactivity condition of the reactor and the additional l
restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.
The boron :apability required below 200*F is sufficient to provide a i
SRWDOW MARGIN of 1% ok/k af ter xenon decey and cooldown from 200'F to i
70'F.
This condition requires either 600 gallons of 7875 ppa borated water from the boric acid storage system or 3,000 gallons of 1800 ppa borated water from the borated vater storage tank.
The botton 4 inches of the borated water storage tank are not available, i
and the instrumentation is calibrated to reflect the available volume.
All boric acid tank volume is available. The limits on water volume, and boron concentration ensure a pH value of between 7.0 and 11.0 of the solution recirculated within containment after a design basis accident.
The pH band minimises the evolution of iodine and minimises the effect of chloride and caustic stress corrosion cracking on mechanical systems and components.
The OFERABILITY of one boron injection system during REFUELING ensures I
that this system is available for reactivity control while in MODE 6.
3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section (1) ensure that acceptable power distribution limits are mai.itained (2) ensure that the minimum SRWD0VN MARGIN is maintained, and (3) limit the potential effects of a rod ejection accident.
OPERABILITY of thq control rod position indicators j
is required to determine control rod positions and*thereby ensure compliance with the control rod alignment snd insertion limits.
l I
The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met.
For example, misalignment of a i
safety or regulating rod requires a restriction in THERMAL POVER. The reactivity vorth of a misa11gned rod is limited for the remainder of the i
fuel cycle to prevent exceeding the assumptions used in the safety t
analysis.
[
l The position of a rod declared inoperable due to misalignment should not i
be included in computing the average group position for determining the OPERABILITY of rods with lesser misalignments.
l I
DAVIS-BESSE, UNIT 1 B 3/4 1-3 Amendment No. 123
o REAC*IV!"Y CCNTROL SYSTIMS BASES 3/4.1.3.
MOVABLE CCNTROL ASSD#.3 LIES (Continued)
The maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analyses. Nessurement with Tavs 2 525'T and with re-actor coolant pumps operating ensures that the ressured drop times will be representative of insertion times experienced during a reactor trip at operat-ing conditions.
Control rod positions and CPERABILITY of the rod position indicators are re-quired to be verified en a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with frequent verifications required if an automatic acnitoring channel is inoperable.
These variitcation frequencies are adequate f or assuring that the applicable LCO's are satisfied.
' Technical Specification 3.1.3.8 provides the ability to prevent excessive power peaking by transient xenon at RATED THERMAL PORER. Operating restrie-tiens resulting frem transient xenon power peakAnn, including xenon-f ree startup, are inherently included in the limits of Sectiens 3.1.3.6 (Regulat-ing Rod Insertion Limits), 3.1.3.9 (Axial Power Shaping Rod Insertion Limits),
and 3.2.1 (Axia\\ Power Imbalance) fer transient peaking behavior bounded by the following factors.
For the period of cycle operation where regulating red groups 6 and 1 are allowed to be inserted at RATED THERMAL POWER, an 8 peaking increase is applied at or above 92: TP.
An 18 increase is applied belov 92: TP.
For operation where caly regulating red group 7 is allowed to be inserted at RATED TRERMAL POWER, a 51 peaking increase is applied at or above 92: TP and a 13 increase is applied belov 92: TP.
If these values, checked every cycle, conservatively bound the peaking sf fects of all transient xen6n, then the need for any hold at a power level cutof f be-lov RATED THEPy.AL POWER is precluded. If not, either the power level at which the requirenants of Section 3.1.3.8 must be satisfied or the above-listed f ac-tors vill be suitably adjusted to preserve the LOCA linear heat rate limits.
The li=1tation on axial ' power ' shaping red insertion is necessary to ensure that.pover peaking limits are not exceeded.
DAVIS-3 ESSE. UNIT 1 B 3/4 1-4 Amendment No. JWP,45
1 i
I 3/4. 5 EMERGENCY CORE COOLING SYSTEMS (ECCS) l BASES 3/4.5.1 CORE FLOODING TANKS i
The OPERABILITY of each core flooding tank ensures that a sufficient volume of borated water will be imediately forced into the reactor vessel in the event the RCS pressure falls below the pressure of the tanks.
This initial surge of water into the vessel provides the initial cooling mechanism during large RCS pipe ruptures.
The limits on volume, boron concentration and pressure ensure that the assumptions used for core flooding tank injection in the safety analysis are met.
The tank power o "operating bypasses" perated isolation valves are considered to be in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.
In addition, as these tank isolation valves fail to meet single failure criteria, removal of power to the valves is required.
The limits for operation with a core flooding tank inoperable for
'iny reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional tank which may result in unacceptable peak cladding tempera-tures.
If a closed isolation valve cannot be imediately opened, the full capability of one tank is not available and prompt action is required to place the reactor in a mode where this espability is not required.
3/4.5.2 and 3/4.5.3 'ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems with RCS average temperature > 280'F ensures that sufficient emergency core cooling capability wT11 be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.
Either subsystem operating in conjunction with the core flooding tanks is capable of supplying. sufficient core cooling to maintain the peak cladding tempera-tures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward.
In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery 4eriod.
DAVIS-BESSE, UNIT 1 B 3/4 5-1 A, endment No. 2 0
e EMERGENCY CORE COOLING SYSTEMS BASES Vith the RCS temperature belov 280'F, one OPERAELE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.
The Surveillance Requirements provided to ensure OPERABILITY of each component ensures, that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.
The decay heat removal system leak rate surveillance requirements assure that the leakage rates assumed for the system during the recirculation phase of the lov pressure injection vill not be exceeded.
Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flovs vill be main-tained in the event of a LOCA.
Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to (1) prevent total pump flov from exceeding runout conditions when the system is in its minimum resistance configuratinn, (2) provide the proper flov split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flov to all injection points equal to or above that assumed in the ECCS-LOCA analyses.
3/4.5.4 BORATED VATER STORAGE TANK The OPERABILITY of the borated water storage tank (BVST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on BVST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flov to the core, and 2) the reactor vill remain suberitical in the cold condition following mixing of the BVST and the RCS vater volumes with all control rods inserted except for the most reactive control assembly.
These assumptions are consistent with the LOCA analyses.
The bottom 4 inches of the borated water storage tank are not available, and the instrumentation is calibrated to reflect the available volume.
The limits on water volume, and boron concentration ensure a pH value of between 7.0 and 11.0 of the solution sprayed within the containment after a design basis accident. The pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion cracking on mechanical systems and components.
DAVIS-BESSE, UNIT 1 B 3/4 5-2 Amendment No. Abi I23 l
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DAVIS-BESSE NUCLEAR POWER STnTION
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LOW POPULATION ZONE FIGURE 5. I-2 l
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DAVIS-BESSE, UNIT 1 5-3 l
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I DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 40 psig and a temperature of 264'F.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 177 fuel assemblies with each fuel assembly containing 208 fuel rods clad with Zircaloy -4.
Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 2500 grams uranium.
The initial core losding shall have a maximum enrichment of 3.0 veight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.3 veight percent U-235.
CONTROL RODS 5.3.2 The reactor core shall contain 53 safety and regulating and 8 axial power shaping (APSR) control rods.
The safety and regulating control rods shall contain a nominal 134 inches of absorber material.
The nominal values of absorber material shall be 80 percent Silver, 15 percent Indium and 5 percent Cadmium. All control rods shall be clad with stainless steel tubing. The APSRs shall contain a nominal 63 inches of absorber material at their lover ends. The absorber material for the APSRs shall be 100 percent Incenel-600.
5.4 REACTOR COOLAtTT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
In accordance with the code requirements specified in Section a.
5.2 of the FSAR, with allowance for normal degradation pursuant to applicable Surveillance Requirements, b.
For a pressure of 2500 psis, and For a temperature of 650'F, except for the pressurizer and c.
prercuri:ct :urgc line which is 670'F.
DAVIS-BESSE, UNIT 1 5-4 Amendment No. TY.
19' 123
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