ML19322B316
| ML19322B316 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/25/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19322B304 | List: |
| References | |
| NUDOCS 7912020164 | |
| Download: ML19322B316 (14) | |
Text
b UNITED STATES j )",
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NUCLEAR CEGULATORY COMMISSION g
WASHINGTON. D. C. 20585 e
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0 SAFETY EVALUATION BY THE'0FFICE OF NUCLEAR' REACTOR' REGULATION SUPPORTING AMENDMENT NO. 20 TO FACILITY LICENSE NO. DPR-38 SUPPORTING AMENDMENT NO. 20 TO FACILITY' LICENSE'NO. DPR-47 SUPPORTING AMENDMENT NO. 20 TO FACILITY LICENSE ~NO. DPR-55 DUKE POWER COMPANY _
OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270 AND 50-287 Introduction By letter dated December 1,1975, Duke Power Company (the licensee) requested a change in the Technical Specifications of License No. DPR-38 for the Oconee Nuclear Station, Unit 1.
The proposed amendment is to permit operation of. Unit 1 as reloaded for Cycle 3 operation.
Included in the bases of the analyses performed are the Final Acceptance Criteria (FAC) for Emergency Core Cooling Systems, as required by the Commission's Order for Modification of License dated December 27, 1974.
Discussion The Oconee Unit i reactor core consists of 177 fuel assemblies, each with a 15 x 15 array of fuel rods. The cycle 3 reload will involve the removal of all of the batch 2 fuel (36 assemblies) and 24 of the batch 3 assemblies. The remainder of the batch 3 assemblies and the batch 4 assemblies will be reassigned to new locations for cycle 3 operation.
The fresh batch 5 assemblies will occupy primarily the periphery' of the core and 4 major axes positions slightly interior to the core.
The fuel to be added to the core is not significantly different in design or in operating characteristics from the original fuel it The rearrangement of fuel assemblies in the reloaded core replaces.
will affect core physics and thermal hydraulic calculations, and as a result, appropriate changes to the Technical Specifications have been submitted.
The licensee has provided technical information which includes a general description of the reload core, detailed mechanical design data on the j
reload fuel, nuclear and thermal-hydraulic design data, accident.and transient analyses, fuel rod bow analyses and the loss of coolant accident (LOCA) analysis in support of the reload.
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N 7912020((f
, Evaluation 1.
Fuel and Mechanical Design Creep collapse calculations were performed by the licensee for three-cycle assembly power histories for Oconee Unit I using the Babcock 6 Wilcox (B6W) computer code, CROV, which we approved in our Generic Review of the B6W Cladding Creep Collapse Analysis Topical Report, BAW-10084, issued on August 9, 1974. The calcula-tions included conservative treatment of effects of fission gas (no credit taken), cladding thickness (lower tolerance limit),
initial cladding ovality (upper tolerance limit), and cladding temperature (assembly outlet temperature) on collapse time. The most limiting assembly was found to have a collapse time of more than 26,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> which is greater than the maximum projected cycle 3 life of 21,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> and is therefore acceptable.
Fuel thermal analysis calculations that account for the effects of fuel densification were performed with our approved version of the BGW analytical model TAFY as described in B6W Topical Report BAW-10044 of May 1972. Fuel densification results in increases in stored energy, linear thermal output and the probability of local power spikes from axial gaps. During cycle 3 operation, the highest relative assembly power levels will occur in batches 4 and 5 fuel.
Fuel temperature analysis for batches 1, 2 and 3 fuel is documented in the Oconee 1 Fuel Densification Report, BAW-1388, Revision 1 of July 1973. We agree that this analysis is also applicable to batches 4 and 5 fuel because they have the same linear heat rate capabilities to centerline melt as batches 1, 2 and 3 (20.15 kw/ft).
In view of the above, w'e find 'the licensee's fuel thermal analysis acceptable.
'Ihe batch 5 fuel assemblies.are not new in concept and they do not utilize different component materials. Therefore, on the bases of the analysis presented in the reports referenced, 'we conclude for Oconee Unit 1 cycle 3 that:
(a) The fuel rod mechanical design provides acceptable safety margins for normal operation, and (b) The effects of fuel densification have been adequately accounted for in the fuel design.
2.
Thermal-Hydraulic Analysis The thermal-hydraulic calculations for the Unit l' cycle 3 reload core were made using previously approved models and methods. There were no changes due to mechanical differences since the new fuel assemblies are mechanleally similar and flow resistances are identical to the previously aralyzed cycle 2 core.
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1 As reported in the licensee's letter of August 23, 1973, precision measurement tests of reactor coolant flow were conducted at Oconee Unit 1.
As in the cycle 2 reload, a measured flow value based on the coolant flow measurements, instead of the ' system design flow, is used for the thermal hydraulic analysis for cycle 3.
The coolant flow measurement test results referred to above showed a measured flow rate of 107.8+.82% of design flow. As discussed in the licensee's Startup Report for Unit I dated November 16, 1973, corrections to the test data increased this value to 108.6% f design flow. The value of system flow selected for the-cycle 3 (and cycle 2) thermal hydraulic analysis, 107.6%, is' conservative with' respect to the test results referenced above.
The flux / flow trip setpoint for a two-pump coastdown previously determined for cycle 1 (supplement 17 to the Oconee FSAR) has been reevaluated for the cycle 3 core. The procedure was revised to use the measured flow, 107.6% of design flow, instead of the previously used design flow rate.
Because of higher system flow rates, most of the orifice plugs have been removed from peripheral fuel assemblies.
l This increased the predicted core bypass flow by 2.3% (from 6.04%
to 8.3%) and has resulted in a 5.3% increase in core flow from.the, measured 7.6% excess in system flow rate. The core bypass flow was taken into account in the analyses based on the increased system I
flow rate.
In addition, a 4.6% flow penalty for an assumed stuck open core vent valve was used in the analysis.
Based on the licensee's reevaluation, a flux / flow ratio of 1.07 was determined to give a satisfactory minimum Departure from Nucleate Boiling Ratio (DNBR) of 1.31 under two-pump coastdown conditions, starting from 108% power.
In the reevaluation, the licensee considered the maximum variation from the average value of the reactor coolant flow signal to provide a conservative indication of flow to the Reactor Protective System.
Consequently, the flux / flow trip set point, as proposed for cycle 3 operation,'is more conservatively established as 1.055.
In addition to consideration of the variations in the reactor coolant 1
flow signal, as discussed above, the licensee has also included an allowance for the accuracy of the RPS instrumentation string. This error was accounted for in the flux value used to establish the flux /
flow trip setpoint.
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The present Technical Specifications include monthly and annual surveillance requirements for the flux / flow comparator instrumentation channels. The monthly calibration check verifies the trip setpoint.
using known test signals and the annual requirement includes the calibration of the entire primary flow instrumentation string using an actual differential pressure as input to the system d/p cells.
The accuracy of these checks are on the order of +1%.
To assure continual confidence in the calibration discussed above, a Technical Specification has been included which will require that the reactor coolant system flow be verified to be at least 141.3 x 106 lbs/hr (107.6% design flow) at least once each fuel cycle.
In summary, the licensee has proposed, as in cycle 2, that a reactor coolant flow rate based on measured flow be used in place of design flow in the analyses involving reactor coolant flow.
In conjunction with this, the flux / flow trip setpoint has been reevaluated to meet i
the revised limiting DNBR of 1.3.
In our review of these items, we i
considered the difference between the value of reactor coolant flow used in the calculations (107.6% design flow) and the actual measured i
flow (108.6% design flow), the accuracy of the calibrations performed and the conservative allowances taken by the licensee in the analyses.
In addition, the 4.6% reactor coolant flow penalty imposed for,an assumed stuck open core vent value has been deter' mined to no l'onger be necessary. This has the effect of adding additional conservatism to the analyses performed for the cycle 3 core.
In view of the above, we conclude that the use of measured rather than design flow is acceptable.
Thermal hydraulic design calculations for cycle 3 cperation utilized the same analytical methods previously documented in the Unit 1 FSAR and the Unit 1 Cycle 2 reload submittal. Adjustments to the calcula-tions were made to account for modifications in the use of the BAW-2 Critical Heat Flux (CHF) correlation which was used for the cycle 2 reload. Two modifications to the BAW-2 CHF correlatica have been introduced fer its application to the cycle 3 core. ';'hese are:
(a) An extension downward from 2000 psia to 1750 psia of the pressure range applicable to the correlation, and (b) A reduction in the DNBR from 1.32, representing a 99% confidence level that 95% of the hot rods will not experience DNB, to r
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- 1.30 representing a 95% confidence level that 95% of the rods will not experience DNB.
We recently completed a re-evaluation of the BAW-2 CHF correlation' to verify its continued suitability in relation to available rod bundle DNB data. We determined that the BAW-2 correlation ~ continues to be an acceptable correlation over the pressure, quality, mass flux, rod diameter and rod spacing range of its original data base.
In coi.iunction with our reevaluation of the BAW-2 CHF correlation we also reviewed the licensee's proposed modifications to the correlation for the cycle 3 core. The original data base for the correlation covered the pressure range 2000-2450 psia and resulted in a 1.32 minimum allowable DNB ratio to ensure with 99% confidence that 95%
of the hot rods did not experience DNB. As an attachment to their letter of February 3,1976, B5W provided information which compared the BAW-2 CHP correlation with data in the low pressure range from five different test bundles. The mean measured-to-predicted ratio for all data was 1.05 and the minimum allowable DNBR was 1.29 for a 95% confidence that 95% of the hot rods at the DNBR would not experience DNB.
The 1.32 minimum DNB ratio used by B6W is based upon 95% of the hot rods at that DNBR not experiencing DNB, with a 99% confidence. If the confidence icyc1 is changed to 95%, which is consistent with the standard review plan and industry practice, the minimum allowable DNBR becomes 1.30.
Based on the above, we find both the extension of the BAW-2 CHF correlation to pressures down to 1750 psia and the change to a minimum DNBR of 1.30 to be acceptable. The BAW-2 CHF correlation has been shown to be conservative in the low pressure region and the 1
change.to a 1.30 minimum DNBR is consistent with the requirements of Standard Review Plan 4.4.
In addition, the proposed reduction is +.he reactor coolant low pressure trip (1800 psig from 1985 psig) it consistent with the extension of BAW-2 CHF correlation downward to 1750 psig and is therefore also acceptable.
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Nuclear Analysis The licensee has provided values for core physics parameters for the Unit I cycle 3 core which reflect minor differences when compared to those for cycle 2.
These differences are attributable to the fact that the core has not yet reached an equilibrium cycle and such differences are to be expected. We have concluded that no significant changes exist in the' core design between cycles 2 and 3.
In addition, the same calculational methods and design information were used to obtain the important nuclear design parameters. Based'on the above and the fact that startup tests (to be conducted prior to power l
operation) will verify that the critical aspects of core performance are within the assumptions of the safety analysis, we find the licensee's nuclear analysis for cycle 3 to be acceptable.
4.
Transient and Accident Analysis s
i The licensee has provided the results of examinations conducted of each FSAR accident analysis with respect to changes in cycle 3 paramete,rs to determine the effects of the reload and to ensure that thermal performance during hypothetical transients !s not degraded.
We hgve reviewed the licensee's submittal and agree that in most cases the consequences of transients are less severe and in no case are they more severe.
j 5.
Rod Bow Penalty I
1 By letter dated February 27, 1976, the licensee provided information to supplement its December 1, 1975 cycle 3 reload submittal which would revise the Technical Specifications to account for the effect of rod bow on core parameters.
In conjunction with these revisions, i
the licensee is also proposing changes to quadrant tilt specifications, applicable to all three Oconee units, which would specify the limit on actual quadrant power tilt, using as a frame of reference the real core power ratio instead of the power ratio measured by just the out-of-core detector system, as is presently done.
In the analysis supporting the proposed Technical Specification changes for Unit I the licensee indicated that:
(a) The rod bow effect on the flow area oi-the hot channel is-adequately compensated for by the flow area reduction factor, (b) The power spike caused by the rod bow effect away from the hot channel, when added to the hot rod in the area of the minimum DNBR, shows that the Unit I cycle 3 DNBR limit (1.30) conservatively accounts for the effects of rod bowing, and i
(c) The power spike due to rod bow, when added to the other factors l
affecting the power imbalance limit for the Reactor Protection i
System (RDS), necessitates a reduction in the core safety and l
RPS imbal :1ce limits. These limits exist to preclude exceeding the centra. fuel melt criteria which is more limiting than DNBR for cycle 3.
In view of the considerations identified in (c) ' bove, the licensee a
is proposing that a rod bow spike penalty of 2.15% be absorbed by reducing the quadrant tilt limit for Unit 1, from 4% to 2.77%.
These values would be the limit when the out-of-core detectors are used for quadrant tilt measurements. To improve clarity and. provide
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a quadrant tilt limit which would be independent of the measurement system used (out-of-core or in-core detector system) the licensee is proposing to also revise the quadrant tilt specifications to refer to actual quadrant tilt and to use this method in the operation of l
all three Oconee Units. The equivalent peaking increase for unit 1 would then be revised from 7.36% to 5.10%, to account for rod bow effects.
4 1
In addition to the power spike penalty associated with the rod bowing phenomenon there has been determined to be a DNB penalty resulting from displaced coolant flow. This penalty, however, is essentially compensated for by allowances made in the design.
B6W has not yet formally submitted l
a rod bow model for our review. The model we have utilized is appropriately j
conservative, however, due to the uncertainties involved and the lack of sufficient supportive data, we have imposed an additional 2% DNB penalty.
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ne licensee's proposed reduction in the quadrant tilt limit to accommodate the rod bow spike penalty is more limiting than the 2% DNB penalty we have imposed and is therefore more conservative.
Based on the above, we find the proposed Technical Specifications for Units 1, 2 and 3 to be acceptable.
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6.
ECCS Analysis On December 27, 1974, the Atomic Energy Commission issued an Order I
for Modification of License implementing the requirements of 10 CFR 50.46, " Acceptance Criteria and Emergency Core Cooling Systems for Light Water Nuclear Power Reactors." One of the requiremer.ts of the Order was that the licensee shall submit a re-evaluation of ECCS l
cooling perfomance calculated in accordance with an acceptable evaluation model which conforms with the provisions of 10 CFR 50.46. The Order also required that the evaluation shall be accompanied by such proposed changes in Technical Specifications or license amendment as may be i
necessary to implement the evaluation results. As required by our Order of December 27, 1974, the licensee, by letter acted July 9, 1975 and as s upplemented August 1,1975, submitted en ECCS reevaluation and related Technical Specifications.
Included in the reload application of December 1, 1975, the licensee has submitted the related Technical Specifications for Unit 1, cycle 3.
The reevaluation and Technical Specifications were submitted using the B6W ECCS evaluation model as described in BAW-10104 of May 1975.
The background of the staff review of the B6W ECCS evaluation model i
and its application to Oconee is described in the staff SER for this facility dated December 27, 1974, issued in connection with the Order for Modification of License. The bases for acceptance of the principal portions of the evaluation model are set forth in the staff's Status Report of October 1974 and the Supplement to the Status Report of November 1974 which are referenced in the December 27, 1974 SER. De December 27, 1974 SER also describes the various changes i
required in the earlier version of the B6W model. Together, the December 27, 1974 SER and the Status Report and its Supplement describe an acceptable ECCS evaluation model and the basis for the staff's acceptance of the model. The Oconee 1 ECCS evaluation which is covered by this safety evaluation report properly conforms to the accepted model. The licensee's July 9, 1975 submittal contains documentation by reference to B6W Topical Reports of the revised ECCS
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. model (with the modifications described in our Decembe and a generic break spectrum appropriate to Oconee 1; BAW 1 r 27, 1974 SER) 1975 and BAW-10103, June 1975, respectively
- 0104, May Company included in this July 9th submittal a sepaIn addition, Duke Po worst break for Oconee Unit 1, using the following pla t rate analysis of the parameters:
n -specific (a)
Power level = 1.02 x 2568 Mwt.
used 1.02 x 2772 Mwt.
The generic analyses in BAW-10103 (b)
Initial average fuel temperature assumed reflects th The generic analyses used T = 30500F. core (T = 30300F e reload
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(c)
Different pin dimensions were employed to reflect f uel changes.
(d) as-built value for Oconee Unit 1 (6.5 versus 7 75 e ect the analyses).
n generic (c)
System enthalpics and steam generator heat loads to reflect the lower power level of 2568 Mwt were changed (f)
Initial pin pressures and oxide layer thickness s to reflect the different fuel in Oconee 1 e
were changed The generic analysis in BAW-10103 identified the w as the 8.55 ft2 double-ended cold leg break at the pump dischar orst break size with a CD = 1.0.
The table below summarizes the results of the LOCA limit analyses which determine the allowabl ge limits as a function of elevation in the core foe linear heat rate 1
r Oconee Unit 1:
Elevation LOCA (ft)
Limit Peak Cladding Temperature (OF)
Max. Local Time of (kw/ft)
_ Ruptured Oxidation Rupture
!)nruptured ~
(%)
Node Node (sec)
Oconee 1 2
16.0 4
1882 17.5 1930 6
1975
'3.40 10.90 18.0 1978 8
2066 3.17 12.39 17.0 2146 10*
1743 5.46 15.55 16.0 2110 1642 5.19 15.01 1931 2.93 39.20
- See discussion below.
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l 9-1 The maximum core-wide metal-water reactor for Oconee 1 was calculated to be 0.557 percent, a value which is below the allowable limit of 1 percent.
As shown in the tabulation, the calculated values for the peak clad temperature and local metal-water reactor were below the allowable limits specified in 10 CFR 50.46 of 22000F and 17 percent, respectively.
BAW-10103 has also shown that the core geometry remains amenable to cooling and that long-term core cooling can be established.
Tne staff noted during its review of BAW-10103 that the LOCA limit calculation at the 10-foot elevation in the core showed reflood rates below 1 inch /second at 251 seconds into the accident (Section 7.2.5).
Appendix K to 10 CFR 50.46 requires that when reflood rates are less than 1 inch /second, heat transfer calculations shall be based on the assumption that cooling is only by steam, and shall take into account any flow blockage calculated to occur as a result of cladding swelling or rupture as such blockage might affect both local steam flow and heat transfer. As indicated by the staff in the Status Report of October 1974 and supplement of November 1974, a steam cooling model.' ' reflood rates less than 1 inch /second was not submitted by B6W for staff review.
The steam cooling model submitted by B4W in BAW-10103 is therefore considered to be a proposed model change requiring further staff review and ACRS consideration. Accordingly, B6W was informed that until the proposed steam cooling model is reviewed, the heat transfer calculation at the 10-foot elevation during the period of steam cooling specified in BAW-10103 must be further justified.
In lieu,cWE using their proposed steam cooling model, B6W has submitted the results of calculations at the 10-foot elevation using adiabatic heatup during the steam cooling 1
period, where this pericd is defined by B6W as the time when the reflood rate'first goes below 1 inch /second to the time that REFLOOD predicts the 10-foot elevation is covered by solid water. The new calculated peak cladding temperature, local metal-water reaction and core-wide metal-water reaction at the 10-foot elevation are 1946oF, 3.02%, and
.647%, respectively. These values remain below the allowable limits of 10 CFR 50.46 and are acceptable to the staff. Until a steam cooling model has been accepted by the staff, these values will serve as the LOCA results for Oconee 1 at the 10-foot elevation.
As indicated above, Duke Power Company elected to provide a plant-specific calculation for Jconee Unit I utilizing selected as-built data. We have reviewed the input changes used (relative to BAW-10103) and believe them appropriate for Oconee Unit 1.
We have reviewed the Technical Specifications proposed by the licensee in the July 9, 1975 submittal, and as revised October.31, 1975, to assure that operation of Oconee Unit I will be within the limits imposed by the Final Acceptance Criteria (FAC) #or ECCS system performance.
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. These criteria permit an increase in the allowabic heat generation rate from 15 to 16 Kw/ft at the 10-foot elevation, as compared to the Interim Acceptance Criteria. For Unit 1, the LOCA-related heat generation limits (maximum of 18.0 Kw/ft) occur in the Cycle 2 reload fuel (batch 4).
We have concluded that the proposed Technical Specifications, as submitted for Unit I cycle 2 operation, meet the necessary criteria and are acceptable. Since Oconee Unit 1 is currently undergoing refueling for Cycle 3 operation we have also reviewed the proposed Technical Specifications for Cycle 3 operation to assure that they also meet the SAC. We have determined that the LOCA related heat generation limits, as for cycle 2, occur in the batch 4 fuel. The maximum LOCA related heat generation rate is therefore unchanged at 18.0 Kw/ft.
Based on the above, we find that the proposed Technical Specifications for cycle 3 operation also meet the FAC of ECCS performance and are therefore acceptable.
Our review of other plant-specific assumptions discussed in the following paragraphs regarding the Oconee 1 analyses addressed the areas of single failure criterion, long-term boron concentration, potential submerged equipment, partial loop operation, ECCS valve interlocks, and the containment pressure calculation.
Single Failure Criterion Appendix K to 10 CFR 50 of the Commission's regulations requires that the combination of ECCS subsystems to be assumed operative shall be those available after the most damaging single failure of ECCS equipment has occurred. The licensee has assumed all containment cooling systems operating to minimize containment pressure and has separately assumed the loss of a 4160 Volt Feeder Bus resulting in the operation of only one LPI and one HPI pump to minimize ECCS cooling.
A review of Oconee 1 piping and instrumentation diagrams indicated that the spurious actuation of certain motor-operated valves could affect the appropriate singic failure assumptions. A spurious actuation of core flooding tank (CFT) vent valves CF-5 or CF-6 would result in a decrease in CFT pressure. The rate at which this decrease occurs is controlled by a preset needle throttling valve (CF-16 or CF-18) downstream of the electrically-operated valve. The predetermined position i
of the needle valve is provided by manually turning the local handwheel such an amount as to limit the rate at which a depressurization of the CFT could take place. A recent test at Oconee indicated that the tested valve setting allowed 17 minutes for the CFT pressure to decay from 625 psi to the low pressure alarm, 580 psi, when the electrically-operated valves were opened. Since it is clear that CFT pressure is important to mitigating the consequences of a LOCA, a Technical Specification is included which will require that the normally closed motor-operated valves CF-5 and CF-6 have their breakers locked open and tagged except when adjusting core flood tank pressure.
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- A review was also conducted of the electrical schematics for ECCS motor-operated valves.
It was determined that a single failure of valve interlocks could not affect the appropriate single failure assumptions.
To further minimize the potential for a water hammer due to the discharge of ECC water into a dry line, we will require that valves LP-21 and LP-22 be Icft in the open position during normal operation.
maintains the LPI lines filled with a continual supply of water from This the BWST due to the available static head built into the system Such a configuration will also eliminate the need for one automatic of these valves to provide water to the LPI pumps. safety a The normal value lineup in HPI system provides a similar supply of water to the IfPI pumps In addition, a Technical Specification is included to require the me.nthly venting of ECCS (llPI and LPI) pump casings to ensure that na air pockets have formed.
performed prior to any ECCS flow tests.
Such venting will also be Containment Pressure The ECCS containment pressure calculations for Oconee Class plants were performed generically by BSW for reactors of this type as described in BAN-10103 of June 1975.
Our review of BGW's evaluation model was published in the Status Report of October 1974 and supplement of November 1974.
We concluded that B6W's containment pressure model was acceptable for ECCS evaluations.
input parameters used in the containment analyses be submitted review of each plant.
Oconee I was submitted in the licensee's submittal of July 9A conta ur j
, 1975.
t Justification for the containment input data was submitted for Oconee Unit I by letter dated October 10, 1975.
This justification allows comparison of the actual containment parameters for Unit I with those assumed in the July 9,1975 submittal and BAN 10103 of June 197S.
heat sinks, and operation of the containment heat-remov regard to the conservatism for the ECCS analysis.
based on as-built design information.
This evaluation was The containment heat removal systems were assumed to operate at their maximum capacities, and minimum operation values for the spray water and service water temperatures were assumed.
be conservative for Oconee Unit 1.The containment pressure analysis was demon i
a
We have concluded that the plant-dependent information used for the ECCS :ontainment pressure analysis for Oconee 1 is reasonably conservative and, therefore, the calculated containment pressures are in accordance with Appendix K to 10 CFR 50 of the Commission's regulations.
Long-Term Boron Concentration We have reviewed the proposed procedures and the systems designed for preventing excessive boric acid buildups in the reactor vessel during the long-term cooling period after a LOCA.
Duke Power Company has agreed to implement procedures for Unit I which would allow adequate boron dilution during the long-term and which will comply with the single failure criterion. These procedures will employ a hot leg drain network similar to the concept described in BAW-10103. To employ a singic failure proof mode, Duke Power Company will make modifications to the existing Decay Heat Removal (DHR) design during the cycle 3 refueling outage.
The proposal consists of the addition of two drain lines from the decay heat drop line to the sump. One line (installed upstream of the DiiR isolation valves) will include two qualified motor-operated valves. The other line (installed downstream of the DliR isolation valves) will include one qualified motor-operated valve. By letter dated February 24, 1976, the licensee indicated its intention to test the design and installation of the drain lines by conducting a preoperational test prior to reactor startup.
In addition, by letter dated March 4, 1976, the licensee committed to the installation, prior to cycle 4 operation, of equipment to provide positive indication of flow in the drain lines, i
We have concluded that the licensee's proposal to prevent long-term boron concentration is acceptable and that the preoperational test to confirm proper installation and functioning will provide adequate assurance during Cycle 3 operation that the system will function under post-LOCA conditions.
Submerged Valves The applicant has conducted a review of equipment arrangement to determine if any valve motors inside the containment will become submerged following a LOCA.
Based on this review, no valves were identified which would be flooded and which would affect short-term or long-term ECCS functions or containment isolation.
Partial Loop Analyses To allow an operating configuration with less than four reactor coolant pumps on the line (partial loop), the staff requires an analysis of the predicted consequences of a LOCA occurring during i
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( the proposed partial loop operating mode (s).
By letter dated August 1, 1975, the licensee submitted an analysis for partial loop operation with one idle reactor coolant pump (three pumps operating). Using a reduced power level of 77% of rated power, B6W performed this analysis assuming the worst-case break (8.55 ft2 DE, Cp = 1) and maximum Linear lleat Generation Rate (UlGR) (18.0 kw/ft) from the 4-pump analysis discussed above. The worst break selected'was located in the active t
leg of the partially idle loop. Placing the break at the discharge of the pump in an active cold leg of the partially idle loop (instead of at the discharge of the pump in an active cold leg of the fully active loop) yields the most degraded positive flow through the cote during the first half of the blowdown and results in higher cladding temperatures.
i The maximum cladding temperature for the one-idle-pump mode of operation A staff review of all input assumptions and cont.lusions
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was 17660F.
resulted in a set of inquiries which were answered by the licensee's t
Ictter of October 31,1975 and B6W's letter of October 10, 1975.
The results of a new analysis were submitted to reflect a more appropriate value of initial pin pressure.
The original partial loop analysis contained in the licensee's letter of August 1, 1975, used an initial pin pressure of 1600 psi. As was demonstrated in the time-in-life sensitivity study, submitted by letter dated August 1, 1975, the worst pin pressure for this analysis should have been 760 psi, The maximum cladding temperature for the re-analysis is 17840F, a value which is within the criterion of 10 CFR 50.46.
Therefore, this analysis may be used to support Duke Power Company's proposed operation with one idle reactor coolant pump.
Since an analysis of ECCS cooling performance with one idle reactor coolant pump in each loop has not been submitted, power operation in this configuration will be limited by Technical Specifications to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Single loop operation (i.e., operation with two idle pumps in one loop)-
will be prohibited, by Technical Specifications, without notifying the Commission.
We have completed the review of the Oconee 1 ECCS performance re-analysis and have concluded:
(a)
The proposed Technical Specifications are based on a 1.0CA analysis performed in accordance with Appendix K to 10 CFR 50.
(b)
The ECCS minimum containment pressure calculations were performed in accordance with Appendix K to 10 CFR 50.
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. l The single failure criterion will be satisfied provided that (c) the modifications as specified above are implemented.
The proposed procedures for long-term cooling after a LOCA are (d)
'Ihe implementation of these procedures during the acceptable.
cycle 3 refueling outage is required to provide assurance that the ECCS can be operated in a manner which would prevent excessive A commitment by the boric acid concentration from occurring.
licensee to install the positive indication to show that the hot Icg drain network is working during post-LOCA conditions is i
required and has been received by letter dated March 4, 1976.
The proposed mode of reactor operation with one idle reactor (e) coolant pump is supported by a LOCA analysis perfomed in accordance with Appendix K to 10 CFR 50. Operation with one idle pump in each loop is restricted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Requests for single loop operation will be reviewed on a case-by-case basis.
We have completed our evaluation of the licensee's Unit I cycle 3 reload application and conclude that the licensee has performed the required analyses and has shown that operation of the cycle 3 core will be within In addition, we conclude applicable fuel design and perfomance criteria.
that the licensee's proposed Technical Specification changes meet the Final Acceptance Criteria based on an acceptable ECCS model conforming to the requirements of 10 CFR 50.46 and that the restrictions imposed 27, 1974 Order for Modification on the facility by the Commission's December of License should be terminated and replaced by the limitations established in accordance with 10 CFR 50.46.
We have determined that the amendment does not authorize a change in i
effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this
'l determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmentri impact and pursuant to 10 CFR 951.5(d)(4) that an environmental statement, negative declaration, or environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
. Conclusion We have concluded, basd on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangeral by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
\\
Dated:
March 25, 1976 Y
t
.n UNITED STATES NUCLEAR REGULATORY C05NISSION_
50-269, 50-270 AND S0-287_
DOCKET NOS.
DUKE POWER COMPANY NCTTICE OF ISSUANCE OF AMENDMENS TO FACILITY OPERATING LICENSES Notice is hereby given that the U. S. Nuclear Regulatory Commission to Facility (the Commission) has issued Amendments No. 20,20 and 17 d
Operating Licenses No. DPR-38, DPR-47 and DPR-55, respectively, issu i
for operation to Duke Power Company which revised Technical Specificat ons
- County, of the Oconce Nuclear Station, Units 1, 2 and 3, located in Oconce as of their date of issuance.
The amendments are effective South Carolina.
ih These amendments (1) revise the Technical Specifications to establ s t ble operating limits for Unit I cycic 3 operation based upon an accep a i
Emergency Core Cooling System evaluation model conforming to' the requ imposed ments of 10 CFR 50.46, (2) terminate the operating restrictions 27, 1974 Order for Modification of on Unit 1 by the Commission's December License and (3) revise the Technical Specifications to specify quadrant power tilt limits for Units 1, 2 and 3 independent of the measurement system used.
The application for the amendments complies with the standards an
), and requirements of the Atomic Energy Act of 1954, as amended (the Act The Commission has made appropriate the Commission's rules and regulations.
findings as required by the Act and the Commission's rules and regu in 10 CFR Chapter I, which are set forth in the license amendments.
79/1M cC2 2
t
. <.. Notice of Proposed Issuance of Amendment to Facility Operating License No. DPR-38 in connection with Unit 1 Cycle 3 reload was published in the FEDERAL REGISTER on February 5,1976 (41 F.R. 5354).
No request for a hearing'or petition for leave to intervene was filed following notice of the proposed action.
The Commission has determined that the issuance of these amendment will not result in any significant environmental impact and that pursuant to 10 CFR 551.5(d?(4) an environmental statement, negative declaration t
or environmental impact appraisal need not be prepared in connection with issuance of these amendments.
For further details with respect to this action, see (1) the application for amendment dated December 1, 1975, as supplemented February 24 and 27, 1976, (2) Amendments No. 20, 20, and 17 to Licenses No. DPR-38, DPR-47, and DPR-55, (3) the Commission's related Safety Evaluation.
All of these items are available for public inspection at the Commission's Public Document Room, 1717 H Street, N
. W., Washington, D. C, and at the Oconce County Library, 201 South Spring Street
, Walhalla, South Carolina 29691.
A copy of items (2) and (3) may be obtained upon request addressed
'c the U. S. Nuclear Regulatory Commission, Washington, D. C.
- 20555, Attention:
Director, Division of Operating Reactors.
5
i e g g.
Dated at Bethesda, Maryland, this 25th day of March 1976.
FOR 'IllE NUCLEAR REGULATORY COSNISSION
/
,, w s4 s449' (
T Robert A. Purple, Chie Operating Reactors Branch #1
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Division of Operating Reactors t
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