ML20043E303
| ML20043E303 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/04/1990 |
| From: | Hannon J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20043E304 | List: |
| References | |
| NUDOCS 9006120275 | |
| Download: ML20043E303 (17) | |
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NUCLEAR RECULATORY COMMISSION 4 s MtMile0 TON D. C. Dette
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l TOLEDO EDISON COMPANY l.
THE CLEVELAND ELECTRIC ILLUMINATING COMPANY l
DOCKET NO. 50-346
,D_AV,IS-BESSE NUCLEAR POWER STATION. UNIT N0. 1 A
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 149 License No. NPF-3
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1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Toledo Edison Company and The l
Cleveland Electric Illuminating Company (the licensees) dated December 1, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I I
B.
The facility will operate in conformity with the application, the l
provisions of the Act, and the rules and regulations of the l
Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:
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i (a) Technical Specifications The Technical Specifications contained in Appendix A, as revised i
through Amendment No.149, are hereby incorporated in the license.
The Toledo Edison Company shall operate the facility in accordance i
with the Technical Specifications.
j This license amendment is effecti'e as of its date of issuance and shall 3.
v be implemented not later than 60 days after issuance.
l FOR THE NUCLEAR REGULATORY COMMIS$10N j-
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John N. Hannon, Director Project Directorate III-3 Division of Reactor Projects - !!!, IV, OffIc&SpecialProjects V
e of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 4,1990 t
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ATTACHMENT TO LICENSE AMENDMENT NO.149 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 r
Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment nuder and contain vertical lines indicating tie area of change. The corresponding overleaf pages are also provided to maintain document completeness.
Remove Insert 2-2 2-2 2-5 2-5 B 2-1 B 2-1 B 2-2 B 2-2 B 2-3 8 2-3 B 2-6 B 2-6 8 3/4 2-1 8 3/4 2-1 B 3/4 2-3 8 3/4 2-3 B 3/4 4-1
-B 3/4 4-1
u Figure 2.1-1 Raassar care safety Liant 2500 2M
= RC High Pressure Trip (618,2388)
RC High Temperature Trio 2200 C
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ACCEPTA8LE OPERATION (633.4;;:
2100
=
2129.8s a,
2000
=
(618,1930.8)
Safety Limit i
1900 (616.09.1900)
(621,4,1929.8)
RC Low Pressure Trip RC Pressure Tomo Trip (608.2.1729.8) 1700 e
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0 590 600 610 620 630 640 650 Reacter Outlet Temperature,1 DAVIS-ar.su, twrr 1 2-2 h h t No.ti..#, 4.
H.AP,5,l'M,149
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L9 SAFETT LIMITS AND LIMITING SAFETT SYSTEN SETTINGS I
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l 2.1 SAFETT LIMITS REACTOR CORE 2.1.1 The combination of the reactor coolant more outlet pressure and outlet temperature shall not assood the safety limit shown in Figure 2.1-1.
i APPLicA3fLITT N00Es 1 and 2.
Mt Whenever the point defined by the seabination of reactor coolant eere J','.
outlet pressure and outlet temperature has aseeeded the safety limit, be
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in 307 STAND 8Y vithis one hour.
REACTOR CORE 2.1.2 The combination of reactor TRERNAL POVER and AXIAL POVER INSALANCE shall not exceed the safety limit shown in Figure 2.1 2 for the various combinations of two, three and four reactor coolant pump operation.
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A?PLICASILITY: MODE 1.
n ACTION:
Whenever the point defined by the combination of Reactor Coolant systen flow, AIIAL POVER INEALANCE and TRERMAL POWER has exceeded the appropriate safety limit, be in 507 STANDBY vithin one hour.-
l REACTOR COOLANT SYSTEN PRESSURE L
2.1.3 The Reactor Coolant systen pressure shall not exceed 2750 psig.
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APPLICABILITY:
MODES 1, 2, 3, 4 and 5.
L AC" TION:
MODES 1 and 2 -
Whenever the Reactor Coolant Systes pressure hat exceeded 2750 psig, be in NOT STANDBY vith the Reactor Coolant system pressure within its limit within one hour.
MODES 3, 4 Whenever the Reactor Coolant System pressure has and 5 exceeded 2750 psig, reduce the Reactor coolant System pressure to within h s limit withir. 5 minu us.
g DAVIS BESSE, UNIT 1 2-1
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- s Table 2.2-1 IE Reacter Fretection System Instrumentation Trip Setpolets Functiemal unit t
Talp setpelat alleueble values
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1.
Itemmel reacter telp IIet applicable.
Ilot applicable.
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Eigh flux
<104.941 et RATED Tusenus pegER with (104.94E of RATED TERRIIAL regER with leer peeps operettag leur pumpe operstlagt i
<80.6% of RATED TERRAhL POWER with (SO.6E of BATW TEEBIIAL PWWER with i
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ihree pumpe operettag ihree pumpe operettagt
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3.
RC high temperature f618'r f61S*FS I"
4.
Flex - Afles/flewIII
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Four pump trip setpolet not to Four pump elleuchle veless met to i
enceA the llelt line et Figure enesed the lielt line et Figure 2.2-18.
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For three pump sporetten.
For three pump operation, see Figure l
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see Figure 2.2-1 2.2-1 5.
RC low pressureIII 18900 0 psig 11900.0pelge 11900.0 psig**
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RC high pressere
$2.155 psig 9 355.0 psis* S 355.0 psig**
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Sc pressure-temperatureIII ' 1(16.05 T, 'F 7957.5) hig' II'
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- F 7957.'5) pelge
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- s. sigh flea p of ac
-c5.tr et a4Tes Tsennat venus with os.tr et a4Ts3 Tatsmat resen with g:
pumpe en ene pump operating in each leep ene pump operstlag la each leep0 l
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<0.0E of RATED tween 4L PouRE with 4.0E et BATED M FEWER with i
Two pumpe operstlag Se one 1eep and iso pumpe operettag to one leep and I
ne ymmps operating in the other leap me pumpe operettag le the other toept I
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<e.ar.c mas D Tasmaat neuen with e.es.t asene inessino ressa niin pumps operettag er only one pump
-pumpe operstlag er only one pump op-
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g operating erettest 1
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Centalement pressere high f4 psig ;
(4 pelgt i
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Table 2.2-1.
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.h IIITrip may be====m11y bypassed when ac5 pressere $320 psig by acteeting shutdoom hypnos provided that a.
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The high flex trip setpoint is p1 of RATED 15EENAL POWER.
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The shutdown bypass high pressere trip setpoint of <1820 pels is imposed.
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l The shutdown typass is removed when RCS pressere >1820 peig.
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- Allowable value for CEMEIEL FINICTICWAL TEST.
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- Allowable value for CEAIEEEL CALIBRATICII.
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9 Allowable value for CEAIEIEL FWICTICIIAL TEST and CEAINIEL CALIBRATICII.
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2.1 SAFETY LIMITS l
l BASES I
l 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 would 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 tamnerature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime would result in excessive claddin from nucleate boiling (DNB)g temperatures because of the onset of departure and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB using critical heat flux (CHF) correlations. The local l
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 B&W-2 and BWC CHF correlations have been. developed to predict DNB l
for axially uniform and non-unifom heat flux distributions. The B&W-2 correlation applies to Mark-B fuel and the BWC correlation applies to all B&W fuel with zircaloy spacer grids. The minimum value of the DNBR during steady state operation, nomal operational transients, and anticipated transients is limited to 1.30 (B&W-2) and 1.18 (BWC). The value corresponds I
to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating i
conditions.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR equal to or greater than the correlation limit is predicted l
for the maximum possible thermal power 112% when the reactor coolant flow is 380,000 GPM, which is approximately 108% of design flow rate for four operating reactor coolant pumps.
(The minimum required measured flow is 389,500GPM).
This curve is based on the following hot channel factors with potential fuel densification and fuel rod bowing effects:
N Fg = 2.83; F3H"I'Ui Z = 1.65 The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal, and form the core DNBR design basis.
DAVIS-BESSE, UNIT 1 B 2-1 Amendment No. JJ,33 N,723,149
b SAFETY LIMITS BASES 1
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 potential fuel rod bow.
j 1.
The DNBR limit produced by a nuclear power peaking factor of l
F0 = 2.83 or the combination of the radial peak, axial peak.
and position of the axial peak that yields no less than the DNBR limit.
2.
The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit ;is 20.5 kw/ft for all fuel in the core.
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 specified flow rates for the two curves of Figure 2.1-2 correspond to the analyzed minimum flow rates with four pumps and three pumps, respectively.
The curve of Figure 2.1-1 is the most restrictive of all possible reactor l
coolant pump-maximum thermal power combinations shown in BASES Figure 2.1.
The curves of BASES 51gure 2.1 represent the conditions at which a minimum DNBR equal to the DNER limit is predicted at the maximum possible thernal l
power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to the corresponding DNB correlationqualitylimit($22%(B&W-2)or+26%(BWC)),whichevercondition is more restrictive.
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DAVIS-BESSE, UNIT 1 8 2-2 AmendmentNo.)),7),f),$),Pp.
111,149
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SAFETY LIMITS 1
BASES For the curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 (B&W-2) i or 1.18 (BWC) and a local quality at the point of minimum DNBR less than
+22%(B&W-2)or+26%(BWC)forthatparticularreactorcoolantpumpsituation.
I The DNBR curve for three pump operation is less restrictive than the four pump curve.
l 2.1.3 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the.
Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer art designed to Section III 1
of the ASME Boiler and Pressure Vessel Code which pemits a maximum transient 1
I pressure of 1105, 2750 psig, of design pressure. The Reactor Coolant System 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.
The Safety Limit of 2750 psig is therefore consistent with 1
l the design criteria and associated code requirements.
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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-DAVIS-BESSE, UNIT 1 B 2-3 Amendment No. JJ,33,#,723,149
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- 2. 2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The reactor protection system instrumentation trip setpoints specified in Table 2.2-1 are the values at which the reactor trips are set for each param-eter. The trip setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.
The shutdown bypass provides for bypassing certain functions of-the reactor-protection system in order to permit control rod drive tests, zero power PHYS-ICS TESTS and certain startup and shutdown procedures. The purpose of the
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shutdown bypass high pressure trip is to prevent nonnal operation.with shut.
down bypass activated. This high pressure trip setpoint is lower. than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The high flux trip setpoint of 5
1 0% prevents any..
significant reactor power from being produced. Sufficient natural circula-tion would be available to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumps were operating.
Manual Reactor Trip The manual reacter trip is a redundant channel to the automatic reactor protec-tion system instrumentation channels and provides manual reactor trip capabil-ity.
High Flux A high flux trip at high power level (neutron flux) provides reactor core pro-tection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.
During nonnal station operation, reactor trip is initiated when the reactor power level reaches 104.94% of rated power.
Due to transient overshoot, heat balance, and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which was used in the safety analysis.
DAVIS-BESSE, UNIT 1 B 2-4 Amendment No f5, 61 a-,.
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LIMITING SAFETY SYSTEM SETTINGS BASES RC High Temperature The RC high temperature. trip < 618'F prevents the reactor outlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients.
Flux -- 6 Flux / Flow The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to acconnodate flow 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-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.
The power level setpoint produced by the power-to-flow' ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum pemissible low flow rate.
Examples of typical power level and low flow rate combinations for the pump situations of Table 2.2-1 that would result in a trip are as follows:
1.
Trip would occur when four reactor coolant pumps are operating if power is 108.0% and reactor coolant flow rate is 100% of full flow rate, or flow rate is 92.59% of full flow rate and power level is 100%.
2.
Trip would occur when three reactor coolant pumps are operating if power is 80.68% and reactor coolant flow rate is 74.7% of full flow 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 l
calculated value of 80.68%.
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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 RCS flow will be greater than or equal to the value in Table 3.2-2.
l DAVIS-BESSE, UNIT 1 B 2-5 Amendment No. 16, 33,A5,61,80,123
LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking. kW/ft limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the power level trip produced by a flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.
RC Pressure - Low. High, and Pressure Temperature The high and low trips are provided to limit the press'ure range in which reactor operation is permitted.
During a slow reactivity insertion startup accident from low power or a slow reactivity-insertion from high power, the RC high pressure setpoint is reached before the high flux trip setpoint. The trip setpoint for RC high pressure, 2355 psig, has been established to maintain the system pressure below the I
safety limit, 2750 psig, for any design transient.
The RC high pressure trip is backed up by the pressurizer code safety valves for RCS over pressure protection..and is therefore set lower than the set pressure for these valves, 1 2525 psig. The RC high pressure trip also backs up the high flux trip.
The RC low pressure,1900.0 psig, and RC pressure-temperature (16.00 Tout
- 7957.5) psig, trip setpoints have been established to maintain the DNB ratio greater than or equal to the minimum allowable DNB ratio for those design accidents that result in a pressure reduction.
It also prevents reactor operation at pressures below the valid range of DNB, correlation limits, protecting against DNB.
High Flux / Number of Reactor Coolant Pumps On In conjunction with the flux - aflux/ flow trip the high flux / number of reactor coolant' pumps on trip prevents the minimum core DNBR from decreasing below the f
minimum allowable DNB ratio hy tripping the reactor due to the loss of reactor coolantpump(s). The pump monitors also restrict the power level for the number of pumps in operation.
DAY!S-BESSE, UNIT 1 B 2-6 Amendment No. 23,f 5,69,$1,149
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3/4.2 POWER DISTRIBUTION LIMITS i
BASES The specifications of this section provide assurance of fuel integrity during Condition I (normal operation) and II (incidents of moderate frequency) events by:
(a) maintaining the minimum DNBR in the core > the minimum allowable DNB ratio during nomal operation and during short term transients, (b) maintaining j
the peak linear power density < 18.4 kW/f t during normal operation, and (c) maintaining the peak power density less than the limits given in the bases to specification 2.1 during short term transients.
In addition, the above criteria must be met in order to meet the assumptions used for the loss-of-coolant accidents.
The power imbalance envelope and the insertion limit curves defined in the CORE OPERATING LIMITS REPORT are based on LOCA analyses which have defined the -
maximum linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance Criteria of 2200'F following a LOCA. Operation outside of the. power imbalance envelope alone does not constitute a situation that -
would:cause the Final Acceptance Criteria to be exceeded should a LOCA occur.
The power imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the insertion limits, as defined in the CORE OPERATING LIMITS REPORT and if the steady-state limit QUADRANT POWER TILT' exists.
Additional conservatism is introduced by application of:
a.
Nuclear. uncertainty factors, b.
Thermal calibration uncertainty.
c.
Fuel densification effects, d.
Hot rod manufacturing tolerance factors.
e.
Potential fuel rod bow effects.
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The ACTION statements which permit limited variations fran the basic require-ments are accompanied by additional restrictions which t;nsures that the original criteria are met.
The definitions of the design limit nuclear power peakdng factors as used in these specifications are as follows:
F Nuclear heat flux hot channel factor, is defined as the maximum local g
fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and rod dimensions, l
l DAVIS-BESSE, UNIT 1 B 3/4 2-1 Amendment No. JJ,33,45 JAA,149 l
^
POWER DISTRIBUTION LIMITS BASES N
F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio 3g of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power.
It has been determined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking and on minimum DNBR at full power are met, provided:
N g i.g3; F3g i 1.71 F
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Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking.
It has been detemined that the above hot channel factor lim-its will be met provided the following conditions are maintained.
1.
Control rods in a single group move together with no individual rod in-sertion differing by more than 16.55 (indicated position) from the group average. height.
2.
Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6.
3.
The regulating rod insertion limits of Specification 3.1.3.6 are main-i tained.
4.
AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE is a measure of the difference in power between the top and bottom halves of the core. Calculations of core average axial pc: king factors for many plants and measurements from operating plants under a variety of operat-ing conditions have been correlated with AXIAL POWER IMBALANCE. The cor-relation shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCE-is maintained between the limits specified in Specifica-tion 3.2.1.
The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum al-lowable control rod insertion and are the core DNBR design basis.
Therefore, for operation at a fraction of RATED THERMAL POWER, the design limits are o
met. Whenusgngincoredetectorstomakepowerdistributionmapstodeter-mine F and F g
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The measurement of total peaking factor F eas,.shall be increased by 1.4 a.
percent to account for manufacturing tolehances and further increased by 7.5 percent to account for measurement error.
i DAVIS-BESSE, UNIT 1 B 3/4 2-2 Amendment No. JJ 61 I
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d POWER DISTRIBUTION LIMITS BASES N
b.
The measurement of enthalpy rise hot channel factor. F increasedby5percenttoaccountformeasurementerrofg'shallbe For Condition II events, the core is protected from exceeding the values given in the bases to specification 2.1 locally, and from going below the minimum allowable DNB ratio by automatic protection on power. AXIAL POWER IMBALANCE pressure and temperature.
Only conditions 1 through 3. above, are mandatory since the AXIAL POWER IMBALANCE is an explicit input to the reactor protection system.
The QUADRANT POWER TILT limit assures that the radial power distribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation.
The QUADRANT POWER TILT limit at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.
In the event the tilt is not corrected, the margin for uncertainty on Fo is reinstated by reducing the power by 2 percent for each percent of tift in excess of the limit.
3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the nonnal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the FSAR initial assumptions and have been analytically demonstrated adequate to main-tain a minimum DNBR greater than the minimum allowable DNB ratio throughout each analyzed transient.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument read-out is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate using delta P instrumenta-tion is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
I DAVIS-BESSE UNIT 1 B 3/4 2-3 Amendment No. 33.4.149 j
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..n 3/4.4 REACTOR COOLANT SYSTEM BASES l
3/4.4.1 REACTOR COOLANT LOOPS 1
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The plant is Gsigned to operate with both reactor coolant loops in operation, and maintain DNBR above the minimum allowable DNB ratio during l
all normal operations and anticipated transients.
With one reactor I
coolant pump ut in operation in one loop. THERMAL POWER is restricted by i
the Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE, ensuring l
that the DNBR will be maintained above the minimum allowable DNB ratio at l
the maximum possible THERMAL POWER for the number of reactor coolant pumps l
in operation or the local quality at the point of minimum DNBR equal to the DNB correlation quality limit, whichever is more restrictive.
In MODE 3 when RCS pressure or temperature is higher than the decay heat removal system's design condition (i.e. 330 psig and 350'F), a single reactor coolant loop provides sufficient heat removal capability. The I
remainder of MODE 3 as well as in MODES 4 and 5 eitner a single reactor coolant loop or a DHR loop will be sufficient for decay heat removal; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor cool 6nt loops are not OPERABLE, this specification requires two DHR loops to be OPERABLE.
Natural circulation flow or the operation of one DHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduc-tion will, therefore, be within the capacity of operator recognition and control.
3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psig.
Each safety valve is designed to relieve 336,000 lbs per hour of saturated steam at the valve's setpoint.
The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating DHR loop, connected to the RCS, provides overpressure relief capability anc' will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient.
The relief capacity of the decay heat removal system relief valve is adequate to relieve any overpressure condition which could occur during shutdown.
In the 3 vent that this relief valve is not OPERABLE, reactor coolant system pressure, pressurizer level and make up water inventory is I
limited and the capability of the high pressure injection system to DAVIS-BESSE, UNIT 1 B 3/4 4-1 Amendment No. 33,38,57,72.128,149 1
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