ML15113A178
| ML15113A178 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/19/1985 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duke Power Co |
| Shared Package | |
| ML15113A179 | List: |
| References | |
| DPR-38-A-142, DPR-47-A-142, DPR-55-A-139 NUDOCS 8510090080 | |
| Download: ML15113A178 (20) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 2065 PUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT NO. I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 142 License No. DPR-38
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Duke Power Company (the licensee) dated May 31, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter 1; 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 (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 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 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 3.8 of Facility Operating License No. DPR-38 is hereby amended to read as follows:
3.R Technical Specifications The Technical Specifications contained in Appendices A and 6, as revised through Amendment No. 142 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
85100900BO 850919 PDR ADOCK 05000269 p
-2
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION ohn F. Stolz, Chief Q eating Reactors Branch #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: September 19, 1985
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.D C 20SSS DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 142 License No. DPP-47
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Duke Power Company (the licensee) dated May 31, 1985, complies with the standards and requi;ements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the 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 common defense and security or to the health and safety of the public; and E. The issuance of thi'.,
-"ment is in accordance with 10 CFR Part 51 of the Commissirn'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 3.8 of Facility Operating License No. DPR-47 is hereby amended to read as follows:
3.9 Technical Specifications The Technical Specifications contained in Appendices A and 6, as revised through Amendment No. 142 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
-2
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATOPY COMMISSION oh F. Stolz, hief ating Reactors Branch #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: Septenber 19, 1985
- e
- o0 UNITED STATES NUCLEAR REGULATORY COMMISSION WwVASHINGTON. D C.
20655 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 139 License No. DPR-55
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Dukp Power Company (the licensee) dated May 31, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter 1;
- 6. 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 (i) that the activities aithorized 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 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 CPR Part 51 of the Commission's regulations and all applicable requirement 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 3.8 of Facility Operating License No. DPP-55 is hereby amended to read as follows:
3.R Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 139 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
-2
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REG'LATORY COMMISSION Jo n F. Stolz, Chief 0
rating Reactors Branch 04 ivision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: September 19, 1985
ATTACHMENTS TO LICENSE AMENDMENTS AMENDMENT NO. 142 TO DPR-38 AMENDMENT NO.
142 TO OPR-47 AMENDMENT NO. 139 TO DPR-55 0OCKETS NOS. 50-269, 50-270 AND 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.
The revised pages are identified by amendment numbers and contain vertical lines indicating the area of change.
Remove Pages Insert Pages 2.1-3c 2.1-3c 2.1-3d 2.1-3d 2.1-9 2.1-9 2.3-3 2.3-3 2.3-10 2.3-10 2.3-13 2.3-13 3.2-1 3.2-1 3.2-2 3.2-2 3.5-17 (3 pages) 3.5-17 (1 page) 3.5-20 (3 pages) 3.5-20 (1 page) 3.5-23 (3 pages) 3.5-23 (1 page) 3.5-26 (2 pages) 3.5-26 (1 page) 3.5-29 (2 pages) 3.5-29 (1 page)
Bases -
Unit 3 The satety limits presented for Oconee Unit 3 hav been generated using the BAW-2 and IWC critical heat flux correlations and the Reactor Coolant System flow rate at 106.5 percent of the design flow (design flow is 131.32 x 106 lbs/hr for four-pump operation). The( fow rate utilized is conservative compared to the actual measured flow rate To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions.
This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature.
The upper boundary of the nucleate boiling regime is termed "departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding failure.
Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of the CHF correlations
, )
The BAW-2 and BWC correlations have been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state oper ation, normal operation transients, and anticipated transients is limited to 1.30 (BAW-2) or 1.18 (BWC). 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 considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits.
The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.
-' n The curve presented in Figure 2.1-IC represents the conditions at which a minimum allowable DNBR is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 139.86 X 106 lbs/hr).
This curve is based on the following nuclear power peaking factors with potential fuel densification and fuel rod bowing effects:
F N = 2.565; F N = 1.71(3)F N = 1.50 q
AlH z
The dcsign peaking combination results in a more conservative DNBR than any other power shape that exists during normal operation.
The curves of Figure 2.1-2C are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowin-g:
2.1-3c Amendments Nos. 142
,142
.139
I.
'The combination of the radial peak, axial peak and position of the axial peak that yields no less than the CHF correlation limit.
- 2.
The combination of radial and axial peak that causes central fuel melting of the hot spot.
The limit is 20.15 kw/ft for fuel rod burnup less than or equal to 1,000 MVD/MTJ and 21.2 kw/ft after 1,000 MW'D/MTU.
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 of Figure 2.1-3C correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each.loop, respectively.
The curve of Figure 2.1-IC is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3C.
A B&W topical report discussing the ehanisms and resulting effects of fuel rod bow has been approved by the NRC 4.
The report concludes that the DNBR penalty due to rod bow is insignificant and unnecessary, because the power production capability of the fuel decreases with irradiation.
Therefore, no rod 6ow DNBR penalty needs to be considered for thermal-hydraulic analyses.
The maximum thermal power for three-pump operation is 88.07 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.07 =
79.92 percent power plus the maximum calibration and instrument error.
The maximum thermal power for other coolant pump conditions is produced in a similar manner.
For each curve of Figure 2.1-3C, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than the CHF correlation limit or a local quality at the point of minimum DNBR less than the CKF correlation quality limit for that particular reactor coolant pump situation. The curve of Figure 2.1-IC is the most restrictive of all possible reactor coolant pump maximum thermal power combination shown in Figure 2.1-3C.
References (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March 1970.
(2) Oconee 3, Cycle 3 - Reload Report, BAW-1453, August 1977.
(3) Correlation of 15 x 15 Geometry Zircaloy Grid Rod Bundle C-F Data with the BWC Correlation, BAW-10143P, Part 2, Babcock & Wilcox, Lynchburg, Virginia, August 1981.
(4) Fuel Rod Bowing in Babcock & Wilcox Designs, BAW-10147P-A, Rev. 1, Babcock & Wilcox, May 1983.
2.1-3d Amendments Nos.
142, 142, &139
THERMAL PC"WER LEVEL.%
120
(-32.4, 112.0)
(32.4. 112.0)
Ml= 0.70 M2-0 70 ACCEPTABLE
(-49.5. 100.0)
P4 PUMP
- 100 (49.5. 100 0)
OPERATION
(-32.4 88.07)
(32.4,188.07)
ACCEPTABLE
-80
-49.5. 76.07) f 3 & 4 PUMP (495. 76.07)
OPERATION
(-32.4 60.63)
(32.4, 60.63)
ACCEPTABLE 0
(-49.5, 48.63) 4.2, 3 & 4 PUMP (49 5. 48.63)
OPERATION 4
I-40
-20
-60
-40
-20 20 40 60 REACTOR POWER IMBALANCE CORE PROTECTION SAFETY LIMITS Unit 3 OCONEE NUCLEAR STATION FIGURE 2.1-2C Amendments Nos.
- 141, 142, & 139 2.1-9
1pvel trip and associated reactor power/reactor power-imbalance boundaries by 1.08% - Unit I for 1% flow reduction.
1.07% - Unit 2 1.07% - Unit 3 Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below the minimum allowable value by tripping the reactor due to the loss of reactor coolant pump(s). The circuitry monitoring pump operational status provides redundant trip protection for DNB by tripping the reactor on a signal diverse from that of the power-to-flow ratio.
The pump monitors also restrict the power level for the number of pumps in operation.
Reactor Coolant System Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure setpoint is reached before the nuclear over power trip setpoint. The trip setting limit shown in Figure 2.3-lA - Unit I 2.3-1B -
Unit 2 2.3-IC - Unit 3 for high reactor coolant system pressure (2300 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient.
(1)
The low pressure (1800) psig and variable low pressure (11.14 T
-4706) trip (1800) psig (11.14 1 T out-4706)
(1800) psig (11.14'To -4706) setpoints shown in Figure 2.3-lA have heen established to maintain to DNB 2.3-lB 2.3-1C ratio greater than or al to the minimum allowable value for those design accidents that result
, a pressure reduction.
(2,3)
Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (11.14 T 4746) out (11.14 T
- 4746) out (11.14 T 4746) out Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (618 0 F) shown in Figure 2.3-lA has been established to prevent excessive core coolant 2.3-1B 2.3-IC temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of 620 0 F.
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 loss-of coolant accidsnt, even in the absence of a low reactor coolant system pressure trip.
2.3-3 Amendments tios. 142,
- 142,
& 139
THERMAL POWER LEVEL,%
(-17.0. 107.0)
(17.0, 107.0)
M10.944 ACCEPTAB E 4 M2=-0.9A4 ACCEPTAB3 E 2,.4 PUMP OPEF ATION I
100
(-35.0. 90.0)
(35.0 90.)
I I
II
(-17.0 79.9)
(17.0 79.9)
RE ACCEPTAB E 3 & 4 PEPUMP OPEF ATION UNTI
(-35.0,62.9)
(35.0, 62.9)
I-60
(-17.0. 52.4)
(17.0, 52.4)
ACCEPTA 3E 2,3, & 4 PUMP OPEF ATION
(-35.0, 35.4) aN3.,
54 I
-20I INT PROTECIE SYLEST IO FIGURE 2.3-2C 2.3-10 Amendments flos.
4 3
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wt 4o 4
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C6 F0c Amendments~~*
Nos 14U 4"
&19231
3.2.
HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS Applicability Applies to the high pressure injection and the chemical addition systems.
Objective To provide for adequate boration under all operating conditions to assure ability to bring the reactor to a cold shutdown condition.
Specification The reactor shall not be critical unless the following conditions are met:
3.2.1 Two high pressure injection pumps per unit are operable except as
,specified in 3.3.
3.2.2 One source per unit of concentrated soluble boric acid in addition to the borated water storage tank is available and operable.
This source will be the concentrated boric ac d storage tank containing at least the equivalent of 1020 ft of 11,000 ppm boron as boric acid solution with a temperature at least 100F above the crystallization temperature. System piping and valves necessary to establish a flow path from the tank to the high pressure injection system shall be operable and shall have the same temperature requirement as the concentrated boric acid storage tank. At least one channel of heat tracing capable of meeting the above temperature requirement shall be in operation. One associated boric acid pump shall be operable.
If the concentrated boric acid storage tank with its associated flowpath is unavailable, but the borated water storage tank is available and operable, the concentrated.boric acid storage tank shall be restored to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the reactor shall be placed in a hot shutdown condition margin equivalent to 1% Ak/k at 200oF within the next twelve hours; if the concentrated boric acid storage tank has not been restored to operability within the next 7 days the reactor shall be placed in a cold shutdown condition within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
If the concentrated boric acid storage tank is available but the borated water storage tank is neither available nor operable, the borated water storage tank shall be restored to operability within one hour or the reactor shall be placed in a hot shutdown condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in a cold shutdown condition within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.2-1 Amendments Nos. 142 142
&139
Bases The high pressure injection system and chemical addition system provide control of the reactor coolant system boron concentration.(1) This is normally accomplished by using any of the three high pressure injection pumps in series with a boric acid pump associated with either the boric acid mix tank or the concentrated boric acid storage tank. An alternate method of boration will be the use of the high pressure injection pumps taking suction directly from the borated water storage tank.(2)
The quantity of boric acid in storage in the concentrated boric acid storage tank or the borated water storage tank is sufficient to borate the reactor coolant system to a 1% dk/k subcritical margin at cold conditions (700 F) with the maximum worth stuck rod and no credit for xenon at the worst time in core life. The current cycles for each unit were analyzed with the most limiting case selected as the basis for all three units. Since only the present cycles were analyzed, the specifications will be re-evaluated with each reload. A minimum of 1020 ft3 of 11,000 ppm boric acid in the concentrated boric acid storage tank, or a minimum of 350,000 gallons of 1835 ppm boric acid in the borated water storage tank (3) will satisfy the requirements. The volume requirements include.a 10% margin and, in addition, allow for a deviation of 10 EFPD in the cycle length. The specification assures that two supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold condition. The required amount of boric acid can be added in several ways. Using only one 10 gpm boric acid pump taking suction from the concentrated boric acid storage tank would require approximately 12.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to inject the required boron. An alternate method of addition is to inject boric acid from the borated water storage tank using the makeup pumps.
The required boric acid can be injected in less than six hours using only one of the makeup pumps.
The concentration of boron in the concentrated boric acid storage tank may be higher than the concentration which would crystallize at ambient conditions.
For this reason, and to assure a flow of boric acid is available when needed, these tanks and their associated piping will be kept at least 10aF above the crystallization temperature for the concentration present. The boric acid concentration of 11,000 ppm in the concentrated boric acid storage tank corresponds to a crystallization temperature of 880F and therefore a temperature requirement of 980 F. Once in the high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures assure boric acid solubility.
REFEREN CES (1) FSAR, Sections 9.3.1, and 9.3.2 (2) FSAR, Figure 6.0.2 (3) Technical Specification 3.3 3.2-2 Amendments Nos. 142
, 142
, & 139
POWER LEVEL CUTOFF*100% FP (2 2.
192260.102)
(30.102)
(220.102)..X6'1 100-RESTRICTED (e0 OPE RAT ION (240.80)
UNACCEPTABLE OPERATION 60 (160,50)
(200.50) 0 040 SHUTDOWN ACCEPTABLE MARGIN OPERATION 20 (90.15)
(0,5)o 0
0 50 100 150 200 250 300 ROD INOEx,%WD 0
25 50 75 100 I
I BANK 5 0
25 60 75 100 I
Il II BANK 6 25 50 75 i00 I
11I BANK 7 ROD POSITION LIMITS FOR FOUR PLrP OPERA.xTION FROM 0 EFPD TO EOC UNIT 3 00KlPOal OCONEE NUCLEAR STATION
- Figure 3.5.2-3 (1 of 1) 3.5-17 Amendments Nos. 142
$ 142,
& 139 J~J
100 RESTRICTED OPERATION (210.77 (236.771 (300 77),
UNACCEPTABLE((160.50 0
S D
UA LACCEPTABLE G
TOPERATION LIMIT 20 (90. 15)
(0.51 0
0 50 100 150 200 250 300 ROD INDEK.%WD 0
26 so 75 100 I
11 I
I BANK 5 0
26 50 76 100 1
I I
BANK 6 0
25 50 75 100 II1 BANK 7 ROD POSITION LIMITS FOR THREE PUMP OPERATION FROM 0 EFPD TO LOC UNIT 3 ou[aoI6 OCONEE NUCLEAR STATION Figure 3.5.2-6 (1 of
- 1) 3.5-20 Amendments Nos.
142, 142
, & 139
100 RESTRICTED OPERATION (170.52) 005.2 2
UNACCEPTABLE (200.50) o OPERATION SHUTDOWN ACCEPTABLE MARGIN OPERATION LIMIT 20 (110.15 (0
50 100 150 200 2 50300 ROD INDEX.%WD 0
25 50 75 100 il I
BANK5 0
25 50 75 100 SI I
I' BANKO6 0
25 50 75 100 I
II II BANK 7 ROD POSITION LIMITS FOR TWO PUMP OPERATION FROM 0 EFPD TO EOC Quno P
I Cb E NUCLEAR STATION Figure 3.5.2-9 (1 of 1) 3.5-23 Amendments Nos.
142., 142, & 139
REACTOR POWER,%FP
(-20.0,102.0) 100 (30.0,102.0) 100 ACCEPfABLE
(-30.0,90.0)
OPER TION 80 60 RESTRICTED OPERATION RESTRICTED OPERATION 40 20
-100
-80
-60
-40
-20 0
20 40 60 80 100 IMBALANCE,%
OPERATIONAL POWER IMBALANCE ENVELOP!
,U(yl OCONEE NUCLEAR STATION FIGURE 3.5.2-12 (1 of 1) 3.5-26 Amendments Nos. 142,
142, & 139
Figure 3.5.2-15 (Deleted)
(Note that no rod position limits exist for Unit 3 axial power shaping rods.)
3.5-29 Amendments Nos.
142, 142, & 139 P7,,MI