ML20128G238
| ML20128G238 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/03/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20128G240 | List: |
| References | |
| NUDOCS 9610080402 | |
| Download: ML20128G238 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 3066Hm01
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ENTERGY OPERATIONS INC.
DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 186 License No. DPR-51 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated April 29, 1996, 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; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(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 license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
9610080402 961003 PDR ADOCK 05000313 P
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.- 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. DPR-51 is hereby amended to read as follows:
2.
Technical Snecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 186, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective within 30 days of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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@lu h-Mv Thomas W. Alexion, Projec Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 3, 1996
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l ATTACHMENT TO LICENSE AMENDMENT NO. Ins FACILITY OPERATING LICFNSE NO. DPR-51 l
DOCKET NO. 50-313 l
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.
REMOVE PAGES INSERT PAGES iv iv 7
7 8
8 9
9 9a 9b 9c 13 13 14a 14b 15 15 142 142
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LIST OF FIGURES NumbeE Title EAER
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I 3.1.2-1 REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN LIMITATIONS 20a 3.1.2-2 REACTOR COOLANT SYSTEM NORMAL OPERATION-HEATUP LIMITATIONS 20b 3.1.2-3 REACTOR COOLANT SYSTEM, NORMAL OPERATION COOLDOWN LIMITATIONS 20e 3.1.9-1 LIMITING PRESSURE VS. TEMPERATURE FOR CONTROL ROD DRIVE OPERATION WITH 100 STD CC/ LITER H-O 33 3.2-1 BORIC ACID ADDITION TANK VOLUME AND CONCENTRATION VS. RCS AVERAGE TEMPERATURE 35a 3.5.4-1 INCORE INSTRUMENTATION SPECIFICATION AXIAL IMBALANCE INDICATION 53a 3.5.4.2 INCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX TILT INDICATION 53b 3.5.4-3 INCORE INSTRUMENTATION SPECIFICATION 53c 3.24-1 HYDROGEN LIMITS FOR ANO-1 WASTE GAS SYSTEM 110be 4.4.2-1 NORMALIZED LIFTCFF FORCE - HOOP TENDONS 85b 4.4.2-2 NORMALIZED LIFTOFF FORCE - DOME TENDONS 85c 4.4.2-3 NORMALIZED LIFTOFF FORCE - VERTICAL TENDONS 85d 4.18.1 UPPER TUBE SHEET VIEW OF SPECIAL GROUPS PER SPECIFICATION 110o2 4.18.3.a.3 5.1-1 MAXIMUM AREA 4 UNDARY FOR RADIOACTIVE RELEASE CALCULATION (EXCLUSION AREA) lila
.6.2-1 MANAGEMENT ORGANIZATION CHART 119 6.2-2 FUNCTIONAL ORGANIZATION FOR PLANT OPERATIONS 120 Amendment No. Se,M,99,4GS,H3, iv M7, M9,186
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1 2.,
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE I
i j
Acolicability Applies to reactor thermal power, reactor power imbalance, reactor coolant i
system pressure, coolant temperature, and coolant flow when the reactor is l
I critical.
l Obiective i
To maintain the integrity of the fuel cladding.
l l
Specification i
I Themaximumlocalfgelpincenterlinetemperatureshallbe
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2.1.1 s 5080 - (6.5 x 10" x B.arnup, NWD/MTU)'F) for TACO 2 applications j
j and $4542 - (5.8 x 10-3(x (Burnup, NWD/MTU)*F for TACO 3 applications.
i operation within this limit to ensured by compliance with the Axial Power Imbalance protective limits preserved by Table 2.3-1 " Reactor Protection System Trip setting Limits," as specified in the COLR.
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j 2.1.2 The departure from nucleate boiling ratio shall be maintained
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j greater than the limits of 1.3 for the BAW-2 correlation and I
1.18 for the BWC correlation. Operation within this limit is ensured by compliance with Specification 2.1.3 and with the 1
l Axial Power Imbalance protective limits preserved by Tr.ble 2.3-1 I
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" Reactor Protection System Trip Setting Limits," as specified in the COLR.
l 2.1.3 Reactor Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the Variable Low RCS
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Pressure-Temperature Protective Limits as specified in the COLR.
1 Banaa l
To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding f
under normal operating conditions. This is accomplished by operating j
within the nucleate boiling regime of heat transfer, wherein the heat i
i transfer coefficient is large enough so that the clad surface temperature j
is only slightly greater than the coolant temperature. The upper boundary i
of the nucleate boiling regime is termed departure from nuclea*,e boiling.
(DNB). At this point there is a sharp reduction of the heat transfer coefficient, which could result in high cladding temperatures and the possibility of eladding failure. Although DNB is not an observable parameter during reactor operation _, the observable parameters of neutron a
l power, reactor coolant flow, temperature, and pressure can be related to DNS through the use of a critical heat flux (CHF) correlation. The l
BAW-2(1) and BWC(2) correlations have been developed to predict DNB and the i
location of DNS for axially uniform and non-uniform heat flux f.
distributions. The BAW-2 correlation applies to Mark-B fuel and the BWC correlation applies to Mark-BE fuel. The local DNS ratio (DNBR), defined as the ratio of the heat flux that would cause DNS at a particular core j
location to the actual heat flux, is indicative of the margin to DNB.
The
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minimum value of the DNBR, during steady-state operation, normal j
operational transients, and anticipated transients is limited to 1.30 j
(BAW-2) and 1.18 (BWC).
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Amendment No. M,4M,446,186 7
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l A DNBR of 1.30 (BAW-2) or 1.18 (BWC) corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this
. is consideted a conservative margin to DNB for all operating conditions.
The difference between the actual core outlet pressure and the indicated reactor coolant system pressure for the allowable RC pump combination has l
been considered in determining the Variable Low RCS Pressure-Temperature l
Protective Limits.
l The variable Low RCS Pressure-Temperature Protective Limits presented in the l
COLR represent the conditions at which the DNBR is greater than or equal to the l
minimum allowable DNBR for the limiting combination of thermal power and number l
of operating reactor coolant pumps which is based on the nuclear power peaking factors (3) as specified in the COLR with potential fuel densification effects.
l 3
i The Axial Power Imbalance Protective Limits in the COLR are based on the
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i more restrictive of two thermal limits and include the effects of potential fuel densification:
1.
The DNBR limit produced by the limiting combination of the radial peak, axial peak, and position of the axial peak.
2.
The combination of radial and axial peak that prevents central fuel melting at the hot spot as given in the COLR.
Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalanen produced by the power peaking.
The flow rates for the Variable Low RCS Pressure-Temperature Protective Limits specified in the COLR correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop.
The Variable Low RCS Pressure-Temperature Protective Limit fort four reactor coolant pumps operating is the most restrictive of all possible reactor coolant pump maximum thermal power combinations as specified in the COLR.
The Variable l
Low RCS Pressure-Temperature Protective Limits in the COLR represent the conditions at which the DNBR limit is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation.
If the actual pressure / temperature point is below and to the right of the pressure / temperature line, the variable Low RCS Pressure-Temperature Protective Limit is exceeded.
l The local quality at the point of minimum DNBR is less than 22 percent (BAW-2)(1) or 26 percent (BWC)(2).
Amendment No. M,W,M,H4,He,186 8
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Using a local quality limit of 22 percent (BAW-2) or 26 percent (BWC) at 1
the point of minimum DNBR as a basis for less than four reactor coolant pumps operating of the Variable Low RCS Pressure-Temperature Protective Limits specified in the 00LR is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.
The DNBR as calculated by the BAW-2 or the BWC correlation continually increases from point of minimum DNBR, so that the exit DNBR is always l
higher and is a function of the pressure.
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The maximum thermal power, as a function of reactor coolant pump operation is limited by the power level trip produced by the flux-flow ratio (percent flow x flux-flow ratio), plus the appropriate calibration and instrumentation errors.
For each combination of operating reactor coolant pumps of the Variable Low RCS Pressure-Temperature Protective Limits specified in the COLR, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 (BAW-2) or 1,18 (BWC) or a local quality at the point of minimum DNBR less than 22 percent (BAW-2) or 26 percent (BWC) for that particular reactor coolant pump combination.
The Variable Low RCS Pressure-Temperature Protective Limit for four reactor coolant pumps operating is the most restrictive because any pressure-temperatura point above and to the left of this curve will be above and to the left of the other curves.
REFERENCES (1)
Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-lOOOOA, May, 1976.
(2)
BWC Correlation of Critical Heat Flux, BAW-10143P-A, April, 1985.
(3)
FSAR, Section 3.2.3.1.1.c.
AmendmentNo.yg,M,H,M,W,4M, 9
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1 pumps (s).. The pump monitors also restrict the power level for I
the number of pumps in operation.
C.
RCS Pressure I
During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip is reached before the nuclear overpower trip setpoint. The trip setting limit shown in Table 2.3-1 for high reactor coolant l
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system pressure (2355 poig) has been established to maintain 1
the system pressure below the safety limit (2750 psig) for any design transient.(2) l 1
The low pressure (1800 poig) and variable low pressure (COLR) trip setpoint shown in Table 2.3-1 have been established to maintain the i
DNB ratio greater than or equal to the minimum allowable DNB ratio for those design accidents that result in a pressure reduction.(2,3)
To account for the calibration and instrumentation errors, the accident analysis used the protective limit specified in the COLR.
l D.
Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit
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(618F) shown in Table 2.3-1 has been established to prevent l
excessive core coolant temperatures in the operating range.
Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of 620 F.
E.
Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
F.
Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system.
The reactor protection system segments which can be bypassed are shown in Table 2.3-1.
Two conditions are Lmposed when the bypass is used:
1.
A nuclear overpower trip set point of 55.0 percent of rated power is automatically imposed during reactor +<utdown.
2.
A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.
Amendment 9,M,49,M,494,H4, E 13
e.
f Table 2.3-1 Reactor Protection System Trio Settino Limits One Reactor Coolant Pu Four Reactor coolant Pumps Three Reactor Coolant Puraps Operating in Each Loop ( )
l operating (Nominal Operating (Nominal (Nominal Operating Shutdown i
Operatino Power - 10011 Operatino Power. 75%l Power. 49%)
Bvoass Nuclear power, % of 104.9 104.9 104.9 5.O(*)
rated, max Nuclear Power based on Protection System Maximum Protection System Maximum Protection System Maximum Bypassed flow (b) and imbalance, Allowable Setpoints for Allowable Setpoints for Allowable Setpoints for 4 of rated, max Axial Power Imbalance Axial Power Imbalance Axial Power Imbalance envelope in COLR envelope in COLR envelope in COLR Nuclear Power based on NA NA 55 Bypassed pump monitors, % of rated, max (C)
High RC system 2355 2355 2355 1720(*)
pres s'are, poig; max Low RC system 1800 1800 1800 Bypassed pressure, psiq. min variable low RC Specified in RCS Specified in RCS Specified in RCS Bypassed system pressure, Pressure-Temperature Pressure-Temperature Pressure-Temperature poig, min Protective Maximum Protective Maximum Protective Maximum Allowable Setpoints Allowable Setpoints Allowable Setpcints figure in COLR figure in COLR figure in COLR RC temp, F, max 618 618 618 618 i
High reactor building 4(18.7 psia) 4(18.7 paia) 4(18.7 psia) 4(18.7 pressure, poig, max psia)
(*) Automatically set when other segments of the RPS (as specified) are bypassed.
(b) Reactor coolant system flow, %
(c)The pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two-pump operation.
(d) Operation with one Reactor Coolant Pump operating in each loop is limited to 24 hrs. with the reactor critical.
Amendment No. 9,M,44,49,W,W,99,494,m,-M8,186 15
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6.12.3 CORE CPERATING LIMITS REPORT a
6.12.3.1 The core operating limits shall be established and documented in the OORE OPERATING LIMITS REPORT prior to each reload cycle or prior to any remaining part of a reload cycle for the following Specifications:
2.1 Safety Limits, Reactor Core - Axial Power Imbalance protective limits and Variable Low RCS Pressure-Temperature Protective Limits 2.3.1 Reactor Protection System trip setting limits -
Protection System Maximum Allowable Setpoints for Axial Power Imbalance and Variable low RC system pressure l
t 3.1.8.3 Minimum Shutdown Margin for Low Power Physics Testing 3.5.2.1 Allowable Shutdown Margin limit during Power Operation 3.5.2.2 Allowable Shutdown Margin limit during Power Operation with inoperable control rods i
3.5.2.4 Quadrant power Tilt limit 3.5.2.5 Control Rod and APSR position limits 3.5.2.6 Reactor Power Imbalance limits 6.12.3.2 The analytical methods used to determine the core operating limits addressed by the individual Technical Specification shall be those previously reviewed and approved by the NRC in Babcock
& Wilcox Topical Report BAW-10179P-A, " Safety Criteria and Methodology for Acceptable Cycle Reload Analyses" (the approved revision at the time the reload analyses are performed). The approved revision number shall be identified in the CORE OPERATING LIMITS REPORT.
6.12.3.3 The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core tt:ermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
6.12.3.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance j
for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
Amendment No. 94, 99, 66, tie, 449, 142 (next page le 146) t49,tge,186
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