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Category:FUEL CYCLE RELOAD REPORTS
MONTHYEARML20196K7981999-03-25025 March 1999 Rev 4 to COLR, for Cycle 8 ML20217D3991998-03-31031 March 1998 Cycle 13 Voltage-Based Repair Criteria 90-Day Rept ML20198B9391997-12-22022 December 1997 Cycle 13 Reload & Colr ML20129A8631996-10-10010 October 1996 Cycle 7 Colr ML20108D0411996-05-0101 May 1996 Cycle 12 Colr ML20082T5221995-04-24024 April 1995 Core Operating Limits Rept ML20080R7861995-02-28028 February 1995 Cycle 11 Colr ML20058F1851993-11-24024 November 1993 Cycle 5 Core Operating Limits Rept ML20044H4151993-05-28028 May 1993 Updated Beaver Valley,Unit 1 Cycle 10 Colr. ML20114B0381992-08-0606 August 1992 Cycle 4 Startup Physics Test Rept ML20096C1281992-04-30030 April 1992 Cycle 4 Core Operating Limits Rept ML20077E6661991-04-30030 April 1991 Cycle 9 Core Operating Limits Rept ML20236R8631987-10-31031 October 1987 Reload Safety Evaluation,Beaver Valley Nuclear Plant Unit 1 Cycle 7 ML20098E9221984-07-31031 July 1984 Reload Safety Evaluation,Beaver Valley Nuclear Plant Unit 1 Cycle 5 ML20085L0031983-09-30030 September 1983 Rev 1 to Reload Safety Evaluation,Beaver Valley Nuclear Plant Unit 1,Cycle 4 ML20071J9491983-02-28028 February 1983 Reload Safety Evaluation,Beaver Valley Nuclear Plant Unit 1,Cycle 4 1999-03-25
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARL-99-154, Monthly Operating Repts for Sept 199 for Bvps,Units 1 & 2. with1999-09-30030 September 1999 Monthly Operating Repts for Sept 199 for Bvps,Units 1 & 2. with L-99-139, LER 99-S01-00:on 990813,uncompensated Loss of Ability to Detect within Single Intrusion Security Detection Zone Occurred.Caused by Procedure non-compliance.Involved Personnel Received Counseling Re Event.With1999-09-0202 September 1999 LER 99-S01-00:on 990813,uncompensated Loss of Ability to Detect within Single Intrusion Security Detection Zone Occurred.Caused by Procedure non-compliance.Involved Personnel Received Counseling Re Event.With L-99-140, Monthly Operating Repts for Aug 1999 for Bvps,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Bvps,Units 1 & 2. with L-99-126, Monthly Operating Repts for Jul 1999 for Beaver Valley Power Station,Units 1 & 2.With1999-07-31031 July 1999 Monthly Operating Repts for Jul 1999 for Beaver Valley Power Station,Units 1 & 2.With L-99-107, Monthly Operating Repts for June 1999 for Bvps,Units 1 & 2. with1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Bvps,Units 1 & 2. with ML20209D9531999-06-27027 June 1999 Inservice Insp Ninety-Day Rept Bvps,Unit 2 Outage 7,Year 1999 L-99-096, Monthly Operating Repts for May 1999 for BVPS Units 1 & 2. with1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for BVPS Units 1 & 2. with L-99-078, Special Rept:On 990326,seismic Monitoring Instruments Were Declared Inoperable.Caused by Resolution of Potential TS Compliance Issue & Work Scheduling Issue.Instrumentation Was Returned to Svc Following Calibr & Declared Operable1999-05-0303 May 1999 Special Rept:On 990326,seismic Monitoring Instruments Were Declared Inoperable.Caused by Resolution of Potential TS Compliance Issue & Work Scheduling Issue.Instrumentation Was Returned to Svc Following Calibr & Declared Operable L-99-079, Monthly Operating Repts for Apr 1999 for Beaver Valley Power Station,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Beaver Valley Power Station,Units 1 & 2.With ML20205L0401999-04-0909 April 1999 SER Accepting Util Relief Requests for Inservice Insp Second 10-year Interval for Beaver Valley Power Station, Unit 2 L-99-054, Special Rept:On 990320,meteorological Tower Wind Speed Sensors Were Declared Inoperable.Caused by Calibration Completed by Vendor Did Not Adequately Cover Full Operating Range of Sensors.Removed Sensors & Sent Offsite1999-04-0505 April 1999 Special Rept:On 990320,meteorological Tower Wind Speed Sensors Were Declared Inoperable.Caused by Calibration Completed by Vendor Did Not Adequately Cover Full Operating Range of Sensors.Removed Sensors & Sent Offsite L-99-058, Monthly Operating Repts for Mar 1999 for Bvps,Units 1 & 2. with1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Bvps,Units 1 & 2. with ML20196K7981999-03-25025 March 1999 Rev 4 to COLR, for Cycle 8 L-99-038, Monthly Operating Repts for Feb 1999 for Bvps,Units 1 & 2. with1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Bvps,Units 1 & 2. with ML20203E1181999-02-10010 February 1999 SER Accepting Proposed Revs to Plant,Units 1 & 2 Quality Assurance Program Description L-99-019, Special Rept:On 990120,meteorological Tower Wind Speed Sensors Declared Inoperable.Caused by Processor Card for Sensor Locked Up & Needed to Be Reset.Heater That Fit Around Shaft of Sensor Replaced1999-02-0505 February 1999 Special Rept:On 990120,meteorological Tower Wind Speed Sensors Declared Inoperable.Caused by Processor Card for Sensor Locked Up & Needed to Be Reset.Heater That Fit Around Shaft of Sensor Replaced ML20196F7011999-01-31031 January 1999 BVPS Unit 2 Heatup & Cooldown Limit Curves During Normal Operation at 15 EFPY Using Code Case N-626 ML20203D4811999-01-31031 January 1999 Monthly Operating Repts for Jan 1999 for Bvps,Units 1 & 2, in Accordance with NRC GL 97-02.With ML20207E6631999-01-28028 January 1999 Rev 0 to EMECH-0713-1, Operational Assessment of SG Tubing at Beaver Valley Unit 1,Cycle 13 ML20210G7041999-01-22022 January 1999 BVPS Unit 1 Facility Changes,Tests & Experiments for 980123-990122 ML20207E5861998-12-31031 December 1998 Annual Rept 1998 for Toledo Edison ML20207E5601998-12-31031 December 1998 Annual Rept 1998 for Pennpower ML20198B9021998-12-31031 December 1998 BVPS Unit 1 Simulator Four Yr Certification Rept for 1995-1998 ML20207E5901998-12-31031 December 1998 Dqe 1998 Annual Rept to Shareholders ML20199C9971998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Bvps,Units 1 & 2. with ML20207E5521998-12-31031 December 1998 Annual Rept 1998 for Ohio Edison ML20207E5761998-12-31031 December 1998 Annual Rept 1998 for Illuminating Co ML20204J6751998-12-31031 December 1998 1998 Annual Rept for Dbnps,Unit 1,PNPP,Unit 1 & BVPS Units 1 & 2 ML20199F5341998-12-29029 December 1998 Safety Evaluation Granting Relief Requests 1-TYP-3-B3.140-1, 1-TYP-3-B5.70-1,1-TYP-3-RH-E-1-1,1-TYP-3-B-G-1, 1-TYP-3-APP-I-1,1-TYP-3-UT-1,1-TYP-3-N-509,1-TYP-3-N-521, 1-TYP-3-N-524,1-TYP-3-B3.120-1 & 1-TYP-3-C6.10-1 ML20198K8551998-12-21021 December 1998 SER Granting Relief Request PRR-5 for Third 10-year Inservice Testing for Beaver Valley Power Station,Unit 1 ML20198A1631998-12-0909 December 1998 SER Approving Implementation Program to Resolve USI A-46 at Facility That Has Adequately Addressed Purpose of 10CFR50.54(f) Request L-98-229, Monthly Operating Repts for Nov 1998 for Bvps,Units 1 & 2. with1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Bvps,Units 1 & 2. with ML20195J3131998-11-12012 November 1998 Safety Evaluation Granting First & Second 10-yr Interval Inservice Insp Request for Relief L-98-210, Monthly Operating Repts for Oct 1998 for Bvps,Units 1 & 2. with1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Bvps,Units 1 & 2. with ML20206G0291998-10-31031 October 1998 BVPS Unit 2 Facility Changes,Tests & Experiments for Period 971101-981031 ML20154R9121998-10-20020 October 1998 Safety Evaluation Accepting Proposed Changes to QA Program Description in Chapter 17.2 of BVPS-2 Ufsar.Proposed Changes Would Modify QA Organization to Allow Warehouse QC Inspectors to Report to Manager of Nuclear Procurement Dept ML20154P7491998-10-19019 October 1998 SE Accepting Second ten-year Interval Inservice Insp Request for Relief RR-1-TYP-2-B5.40-1,Rev 0,for Plant, Unit 1 ML20198F7611998-10-0606 October 1998 Duquesne Light Co,Beaver Valley Power Station 1998 Emergency Preparedness Ingestion Zone Exercise, Conducted on 981006 ML20154C6711998-10-0101 October 1998 Safety Evaluation Concluding That Revised Model Identified in Dl Submittal Was Appropriate for Analysis of Installed Conduit Ampacity Limits.Determined That There Are No Outstanding Safety Concerns with Respect to Ampacity ML20154D5001998-09-30030 September 1998 Special Rept on Overview of BVPS-1 & BVPS-2 TS Compliance Issues & Corrective Action Taken L-98-197, Monthly Operating Repts for Sept 1998 for Beaver Valley Power Station,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Beaver Valley Power Station,Units 1 & 2.With ML20154E2171998-09-28028 September 1998 Follow-up Part 21 Rept Re Defect with 1200AC & 1200BC Recorders Built Under Westronics 10CFR50 App B Program. Westronics Has Notified Bvps,Ano & RBS & Is Currently Making Arrangements to Implement Design Mods L-98-188, Special Rept:During 1998,Unit 2 SG Eddy Current exam,26 Tubes Were Improperly Encoded in SG 2RCS-SG21C During Previous Outage.Use of Independent Databases to Track New Indications Being Implemented as Preventive Measure1998-09-21021 September 1998 Special Rept:During 1998,Unit 2 SG Eddy Current exam,26 Tubes Were Improperly Encoded in SG 2RCS-SG21C During Previous Outage.Use of Independent Databases to Track New Indications Being Implemented as Preventive Measure L-98-178, Monthly Operating Repts for Aug 1998 for Bvps,Units 1 & 2. with1998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Bvps,Units 1 & 2. with ML20155B5871998-08-28028 August 1998 Non-proprietary Rev 1 to 51-5001925-01, Risk Assessment for Installation of Electrosleeves at BVPS & Callaway Plant ML20236X2351998-08-0505 August 1998 Part 21 Rept Re Defect Associated W/Westronics 1200AC & 1200BC Recorders Built Under Westronics 10CFR50,App B Program.Beaver Valley,Arkansas Nuclear One & River Bend Station Notified.Design Mod Is Being Developed L-98-168, Monthly Operating Repts for July 1998 for Bvps,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Bvps,Units 1 & 2 L-98-157, Special Rept:On 980423,inoperability of Seismic Monitoring Instrument Noted.Caused by Obsolescence of Instrument & Inability to Obtain Necessary Spare Parts.Design Change Is Being Pursued to Obtain Replacement Product1998-07-29029 July 1998 Special Rept:On 980423,inoperability of Seismic Monitoring Instrument Noted.Caused by Obsolescence of Instrument & Inability to Obtain Necessary Spare Parts.Design Change Is Being Pursued to Obtain Replacement Product L-98-139, Monthly Operating Repts for June 1998 for Bvps,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Bvps,Units 1 & 2 L-98-119, Monthly Operating Repts for Bvps,Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for Bvps,Units 1 & 2 1999-09-30
[Table view] |
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INPF-73 BEAVER VALLEY UNIT 2 CYCLE 6 l CORE OPERATING LIMITS REPORT I
This Core Operating Limits Report provides the cycle specific ,
parameter limits developed in accordance with the NRC approved I methodologies specified in Technical Specification Administrative l Control 6.9.1.14. I l
Soecification 3.1.3.5 Shutdown Rod Insertion Limits l The shutdown rods shall be withdrawn to at least 225 steps.
Specification 3.1.3.6 Control Rod Insertion Limits .,
Control Banks A and B shall be withdrawn to at least 225 steps. !
Control Banks C and D shall be limited in physical insertion as shown I in Figure 1.
Specification 3.2.1 Axial Flux Difference l NOTE: The target band is 17% about the target flux from 0% to 100% RATED THERMAL POWER.
The indicated Axial Flux Difference: j
- a. Above 90% RATED THERMAL POWER shall be maintained within the 7% target band about the target flux difference.
4
- b. Between 50% and 90% RATED THERMAL POWER is within the limits shown on Figure 2.
- c. Below 50% RATED THERMAL POWER may deviate outside the target band.
Specification 3.2.2 Fg(Z) and Fxy Limits Fg (Z) $ CEQ
, for P > 0.5 P
FQ(Z) $ CEQ
0.5 l Where: CFQ = 2.40 P= THERMAL POWER l RATED THERMAL POWER K(Z) = the function obtained from Figure 3.
BEAVER VALLEY - UNIT 2 1 OF 6 COLR 6 i 9505040065 950424 PDR ADDCK 05000412 i p PDR !
2 ; !NPF-73 ,
l I
The Fxy limits [Fxy(L)] for RATED THERMAL POWER within specific core planes shall be:
Fxy(L) =
Fxy(RTP) (1 + PFXY * (1-P))
Where: For all core planes containing D-BANK: ,
l Fxy(RTP) $ 1.71 For unrodded core planes:
1 Fxy(RTP) 5 1.75 from 1.8 ft. elevation to 5.7 ft. elevation i Fxy(RTP) $ 1.80 from 5.7 ft. elevation to 8.7 ft. elevation i
Fxy(RTP) 5 1,77 from 8.7 ft. elevation to 9.7 ft. elevation l
Fxy(RTP) 5 1.72 from 9.7 ft. elevation to 10.2 ft. elevation l PFXY = 0.2 P= THERMAL POWER RATED THERMAL POWER ]
l Figure 4 provides the maximum total peaking factor times relative power (Fg T *prel) as a function of axial core height during normal core operation.
Specification 3.2.3 FNDH i FNDH $ CFDH *(1 + PFDH *(1-P)) {
l Where: CFDH = 1.62 l PFDH = 0.3 P= THERMAL POWER RATED THERMAL POWER 1
l I
BEAVER VALLEY - UNIT 2 2 OF 6 COLR 6 l J
I
l :
NPF-73 220 l l(54.53, 225)j 200 ,
/ l(100,187 3: 180 f BANK C /
7 j
160 7
/
{140 ,
E z 120 /
/ ,/
/
!(0, 114)l BANK D /
h g 100 j
7 2 /
00 60 #
/
0 '
CC 40 20 '
/
f[(8,0)l 0
0 10 20 30 40 50 60 70 80 90 100 RELATIVE POWER (Percent)
FIGURE 1 l
CONTROL ROD INSERTION LIMITS AS A FUNCTION OF POWER LEVEL BEAVER VALLEY - UNIT 2 3 of 6 COLR 6 l 1
NPF-73 100
@ gg _ UNACCEPTABLE l'II* *0) III* *0)
UNACCEPTABLE _
3 OPERATION / \ OPERATION o
/ \
m 70 '
/ \ \
60 0 f ACCEPTABLE OPERATION
\
& 50 ' '
{O 40 ~'(-31, 50) (31, 50)
$ 30 5
0- 20 10 0 l
-50 -40 -30 -20 -10 0 10 20 30 40 50 l
1 FLUX DIFFERENCE ( AI) %
l i
FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER l
BEAVER VALLEY - UNIT 2 4 of 6 COLR 6 l
t a
NPF-73 1.2 (0,1.00) (6,1.00) 1.0 ~ (10.8, .94) ~
.s0 \
\ '
(12,.647) g .60 g
.40
.20 I
0.00 O 2 4 6 8 10 12 CORE HEIGHT (Feet)
FIGURE 3 Fa' NORMAllZED OPERATING ENVELOPE, K(2)
BEAVER VALLEY - UNIT 2 5 of 6 COLR 6 I
I /
NPF-73 I
2.6 I i l l '
l H 0.0, 2 . 4 0 }- '6.0, 2.40] l l 44 ,, !, AgadbAAAA4&Ai &d6A, '
l & "A&ay j *Angggg,,gagggAAA &d AAll 3
34 l ! I A k '
i
^
4 2.0 1 !
' S i i I i\
1.8 \
m I I i l l i \ 1 g &
l l 2. 0, 1.553{
b w
I I i j t-1.4 l 1.2 H
1.0 BASIS FXY 0.8 -
1.75 FROM 1.8 FT. UP TO 5.7 FT.
0.6 1.80 FROM 5.7 FT. UP TO 8.7 FT.
. 1,77 FROM 8.7 FT. UP TO 9.7 FT.
0.4 1.72 FROM 9.7 FT. UP TO 10.2 FT.
n ,,
0.2 0.0 s 0 2 4 6 8 10 12 l CORE HEIGHT (Feet)
FIGURE 4 MAXIMUM (FJ* Pm) VS. AXIAL CORE HEIGHT DURING NORMAL CORE OPERATION l l
i BEAVER VALLEY- UNIT 2 6 of 6 COLR 6 l l 1
l i
l
> ?
4 ATTACHMENT A l
Beaver Valley Power Station, Unit No. 2 Cycle 6 Reload and Core Operating Limits Report Technical Specification Bases Change
- . .\
This change modifies Bases 2.1.1, Reactor Core, to incorporate the Improved Standard Technical Specification wording from NUREG-1431 to address the change in peaking factors provided in the Cycle 6 COLR.
Remove Insert l 1
B 2-1 B 2-1 B 2-2 B 2-2
+ l l
l l
l l
l l
r
~
NPF-73 2.1 SAFETY LIMITS BASES L1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel 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 temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) 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 through the WRB-1 correlation. The WRB-1 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 DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting fuel rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for the WRB-1 correlation).
Incorporating the peaking factor uncertainties in the correlation limit results in a DNBR design limit value of 1.21.
This DNBR value must be met in plant safety analyses using nominal values of the input parameters that were included in the DNBR uncertainty evaluation. In addition, margin has been maintained in the design by meeting a safety analysis DNBR limit of 1.33 in performing safety analyses.
The curve of Figure 2.1-1 shows the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
The curves are based on enthalpy hot channel factor limits prov ded in the Core Operating Limits Report (COLR).
i BEAVER VALLEY - UNIT 2 B 2-1 Revised by NRC letter dated
1 o y NPF-73 SAFETY LIMITS BASES _
2.1.1 REACTOR CORE, continued These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(AI) function of the overtemperature AT trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the overtemperature AT trip will reduce the setpoint to provide protection consistent with core safety limits.
2.1.2 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, pressurizer, and the RCS piping, valves and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirenents.
The entire Reactor Coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a reactor trip system or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allovable Values for the reactor trip setpoints have been specified in Table 2.2-1. Operation with setpoints less conservative chan the Trip Setpoint but within the Allowable Value is BEAVER VALLEY - UNIT 2 B 2-2 Revised by NRC letter dated 1 i