ML19312C464
| ML19312C464 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Crystal River  |
| Issue date: | 07/14/1977 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C459 | List: |
| References | |
| NUDOCS 7912130992 | |
| Download: ML19312C464 (12) | |
Text
,
O UNITED STATES j
g*
NUCLEAR REGULATORY COMMISSION g
o, WASHINGTON, D. C. 20666
'o i
\\.'.... /
DUKE POWER COMPANY DOCKET NO. 50-2 OCONEE NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 44 License No. OPR-38 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendments by Duke Power Company (the licensee) dated March 10, 1977, cor' plies wi th the standards and requirements of tne Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regula-tions set forth in 10 CFR Chapter I:
l B.
The facil'ity 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; l
and E.
Af ter weighing the environmental aspects involved, the issuance of this amendment is in accordance with 10 CFR Part 51 of the l
Commission's regulations and all applicable requirements have l
been satisfied.
2.
Accordingly, the license is amended by a change to the Techical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. DPR '38 is hereby amended to read as follows:
1 1
i 7 912130 kM
n
. "( 2 ) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 44, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifica-tions."
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION n
. M m. v F44 /
c A Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Sepcifications Date of Issuance:
July 14,1977
/
UNITE 3 STATES O."
- t NUCLEAR REGULATORY COMMISSION i
3
)
CASHINGTON, D. C. 20666 t
DUKE POWER COMPANY
-1 b DOCKET NO. 50-299-OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 44 License No. DPR-47 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendments by Duke Power Company (the licensee) dated March 10, 1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regula-tions set forth in 10 CFR Chapter I:
B.
The facil.ity will operate in conformity with the application, the provisions of the Act, and tne 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 b:
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.
After weighing the environmental aspects involved, the issuance of this amencment 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 a change to the Techical Specifications as indicated in the attachment to this license
&mendment and paragraph 3.8 of Facility License No. DPR-47 is hereby amended to read as follows:
I l
N
)
"(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 44, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifica-tions."
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION p
A. Schwencer, Chief
$ D Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes. to the Technical Sepci fications
. Date of Issuance:
July 14,1977
l
}
d UNITED STATES p
NUCLEAR REGULATORY COMMIS$10N
^
y" w AsHINGToN. D. C. 20555
,/
DUKE POWER COMPANY 00CKETNO.50-2N OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING L CENSE Amendment No. 4j License No. DPR-55 The Nuclear Regulatory Commission (the Commission) has found that:
1.
A.
The application for amend tents by Duke Power Company (the licensee) cated March 10, 1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regula-tions set forth in 10 CFR Chapter I:
B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized C.
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; The issuance of this amendment will not be inimical to the common D.
defense and security or to the health and safety of the public; and E.
After weighing the environmental aspects involved, 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.
Accordingly, the license is amended by a change to the Techical 2.
Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility License No. DPR-55 is hereby amended to read as follows:
i
. "(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 41, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions."
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION cJ,L.m.-p;..AA.
n A. Schwencer, Chief b Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Sepcifications Date of Issuance: July 14, 1977
)
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 44TO DPR-38 AMENDMENT NO. 44 TO DPR-47 AMENDMENT NO. 41 TO DPR-55 00CKETS NOS. 50-269, 50-270 AND 50-287 Revise Appendix A as follows:
1.
Remove pages 3.la, 3.1-4, 4.2-2, 4.2-3 and replace with identically numbered pages 2.
add page 4.2-4 b
~
P I
(
I
g., e c - Ur.i t 1
~
J All cospen:nts in ths R:ector Coolent Systm cra d2signid to withstand th2 effects of cyclic loads due to syste= te=perature and pressure changes.
These cyclic loads are introduced by nor=al load transients, reactor trips, startup and shutdown operations, and inservice leak and hydrostatic tests.
The various categories of load cycles used for design purposes are provided in Table 4.8 of the FSAR.
The major components of the reactor coolant pressure boundary have been Results of this analysis, analyzed in accordance with Appendix G to 10CFR50.
including the actual pressure-te=perature li=itations of the reactor coolant pressare boundary, are given in BAW-1421(7).
Figures 3.1.2-1A, 3.1.2-2A, and 3.1.2-3 present the' pressure-te=perature limit curves for normal heatup, nor=al cooldown and hydrostatic test respectively. The 11=1: curves are applicable up to the fifth effective These curves are adjusted by 25 psi and 100F full power year of operation.
for possible errors in the pressure and te=perature sensing instruments.
adjusted for the pressure differential between The pressure 11=it is als the point of syste= pressure =easurement and the limiting co=ponent for all operating reactor coolant pump ce=binations.
The pressure-tenperature limit lines shown on Figure 3.1.2-1A for reactor criticality and on Figure 3.1.2-3 for hydrostatic testing have been provided to assure co=pliance with the =ini=u te=perature requirements of Appendix C to 10CFR50 for reactor criticality and for inservice hydrostatic testing.
The actual shift in RTET of the beltline region material vill be established
~
periodically during operation by re=oving and evaluating, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core areas, or in test reactors.
The limitation on steam generator pressure and te=perature provide protection At against nonductile failure of the secondary side of the steam generator.
metal temperatures lower than the RT g7 of +600F, the protection against nonductile failure if achieved by limiting the secondary coolant pressure The to 20 percent of the preoperational syste= hydrostatic test pressura.
limitations of 110er and 237 psig are based on the highest esti=ated RTET of +400F and the preoperational syste= hydrostatic test pressure of 1312 psig.
The average =etal te=perature is assu=ed to be equal to or greater than the coolant te=perature. The' limitations include =argins of 25 psi and 100F for Possible instru=ent error.-
The spray ta=perature difference is imposed to =aintain the ther=al stresses at the pressurizer spray line nozzle below the design limit.
,k&
3.1-3a
/c,endments Nos.
44, 44 & 41 c
3%_
s Unien 2 and 3 All reactor coolant systro cerponsnes cro dasigned to withstcnd tha cffcets (1) These of cyclic loads due to system te=perature and pressure changes.
cyclic loads are introduced by unit load transients, reactor trips, and unit bestup and cooldown operations.
The number of ther=al and loading cycles The maximum used for design purposes are shown in Table 4-8 of the FSAR.
unit heatup and cooldown rate of 1000F per hour satisfies stress limits for cyclic operation.
(2) The 237 psig pressure limit for the secondary side of the steam generator at a te=perature less than 1100F satisfies stress levels for temperatures below the DTT.
(3) The reactor vessel plate material and welds have been tested to verify conformity to specified requirements and a maximum NDTT value of 200F has been determined based on Charpy V-Notch The maximum NDTT value obtained for the stea= generator shell material tests.
and welds was 400F.
Figures 3.1.2-1B and 3.1.2-2B contain the limiting reactor coolant system pressure-temperature relationship for operation at DTT(4) and below to assure that stress levels are low enough to preclude brittle fracture.
These stress levels and their bases are defined in Section 4.3.3 of the FSAE.
As a result of fast neutron irradiation in the region of the core, there will be an ir.erease in the NDTT with accu =ulated nuclear operation.
The predicted maxim.m NDTT increase for the 40-year exposure is shown on Figure 4.10. (4)
The actual shift in ND~T will be deter =ined periodically during plant opera-tion by testing of irradiated vessel raterial sa=ples located in this or a similar reactor vessel or in test reactors.(5)
.The results of the. irradiated sample testing will be evaluated and compared to the design curve (Figure 4-11 of FSAR) being used to predict the increase in transition temperature.
The design value for fast neutron (E > 1 MeV) exposure of the reactor vessel is 3.0 x 1010 n/c=2 - s at 2,568 We rated pewer and an integrated exposure of 3.0 x 1019 n/cm2 for 40 years operation.
(6)
The calculated maxi =u=
values are 2.2 x 1010 n/cm2 -- s and 2.2 x 1019 n/cm2 integrated exposure for 40 years operation at 80 percent load.
(4)
Figure 3.1.2-1B is based on the design value which is considerably higher than the calculated value.
The DTT value for Figure 3.1.2.1B is based on the projected NDTT at the end of the first two years of operation.
During these two years, the energy output has been conservatively esti=ated to be 1.7 x 106 thermal megawatt days, which is equivalent to 655 days at 2,568 We core power.
The projected 18 fast neutron exposure of the reactor vessel for the two years is 1.7 x 10 n/cm2 which is based on the 1.7 x 106 thermal megawatt days and the design value.'or fast neutron exposure.
The actual shift in NDTT will be established periodical'ly during plant operation by testing vessel material samples which are irradiated curiulatively by securing them near the inside wall of this or a similar yessel in.the core area or in test reactors.
To compensate for the increases in the SDTT_ caused by-irradiation, the limits on the pressure-temperature relationship.are periodically changed to stay within the established stress limits during heatup and cooldown.
3.1-4 Amendments Nos.
44[44&41 d faTTba
'o**o ah c
e
4.2.3 Tha cert.. ural intsgrity of tha Reactcr u.olcnt System boundcry chall b2 maintain:d at th2 leval required by th2 original cc-captance standards throughout the life of the station.
Any evidence, as a result of the tests outlined in Table IS-261 of Section XI of the code, that defects have developed or grown, shall be investigated, including evaluation of comparable areas of the Reactor Coolant System.
4.2.4 The results of the Inservice Inspections perf ormed pursuant to Specifica tions 4.2.1, 4.2.2, and 4.2.3 shall be reported to the Commission within 90 days of completion.
4.2.5 To assure the structual integrity of the reactor internals through-out the life of the unit, the two sets of main internals bolts (connecting the core barrel to the core support shield and to the lower grid cylinder) shall remain in place and under tension.
This will be verified by visual inspection to determine that the welded bolt locking caps remain in place.
All locking caps will be inspected af ter hot functional testing and whenever the internals are removed from the vessel during a refueling or maintenance s hutdown.
The core barrel to core support shield caps will be inspected each refueling shutdown.
4.2.6 Sufficient records of each inspection shall be kept to allow com-parison and evaluation el future inspections, i
4.2.7 The inservice inspection program shall be reviewed at the end of five years to consider incorporation of new inspection techniques and equipment which have been proved practical and the conclusions of this review and evaluation shall be discussed with the NRC, 4.2.8 At approximately three-year intervals, the bore and keyway of each reactor coolant pump flywheel shall be subjected to an in-place, volumetric examination. Whenever maintenance or repair activities necessitate flywheel removal, a surface examination of exposed surfaces and a complete volumetric examination shall be performed, if the interval measured from the previous such inspection is greater than 6 2/3 years.
4.2.9 The reactor vessel natcrial irradiation surveillance specimens removed from Units 1, 2 and 3 reactor vessels in 1976 shall be installed, irradiated in and withdrawn from the Crystal River j
Unit 3 reactor vessel in accordance with the schedule shown in Table 4.2-1.
Following withdrawal of each capsule listed in Table 4.2-1, Duke power Company shall be responsiole for testing 1
the specimens in those capsules and submitting a report of test results in accordance with 10 CFR 50, Appendix H.
D " " lD D '9~ Y f
.o A o
XX n
4.2 2 Amendments Nos. 44, 44 & 41 m
A
The licensee shall submit a report or application for license 4.2.10 amendment to the NRC within 90 days after the occurrence of the following: After March 13, 1978, any time that Crystal River Unit No. 3 fails to maintain a cumulative reactor utilization factor of greater than 45%.
The report shall provide justification for continued operation of Oconee Nuclear Station Units 1, 2 and 3 with the reactor vessel surveillance program conducted at Crystal River Unit No. 3 or the application for license amendment shall propose an alternative program for conduct of the reactor vessel surveillance program.
4.2.11 During the first two refueling periods, two reactor coolant system piping elbows shall be ultrasonically inspeeted along their long-itudinal welds (4 inches beyond each side) for clad bonding and for cracks in both the clad and base metal. The elbows to be inspected are identified in B5W Report 1364 dated December 1970.
.2.12 To assure that reactor internals vent valves are not opening during operation, all vent valves will be inspected during each refueling outage to confim that no vent vr.1ve is stuck open and that each valve operates freely.
Bases The surveillance program'has been developed to co= ply with Sec': ion II of the ASME Roiler and Pressure Vessel Code. Inservice Inspection of Nuclear Reactor Coolant Systems, 1970, including 1970 winter addenda, edition.
The progrc::.
places major emphasis on the area of highest stress concentrations and on areas.where f as t neutron irradiation might be suf ficient to change material properties.
)
n e number of reaccor vossci specimens and the frequencies for removing and testing these specimens are provided to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
For the purpose of Technical Specification 4.2.10.
Cumulative reactor utili:ation factor is defined as: [(Cumulative thermal mejawatt hours since attainment of commerical operation at 100% power) x 100] +
.(licensed thermal power) x (cumulative hours since attainment of consercial operation at 100% power)].
The definition of Regulatory Guide 1.16, Revision 4 (August 1975) applies for the term l
"commerical operation".
Early inspection of Reactor Coolant System piping elbows is considered desirable in order to reconfirm the integrity of the carbon steel base metal when explosively clad with sensitized stainless steel.
If no degradation is observed during the two annual inspections, surveillance requirements will revert to Section XI of the ASME Boiler and Pressure Vessel Code.
OP"}D *0 ~3~ jg J.k m W Ju a
Amendments Nos. 44, 44 & 41
~
H TABLE 4.2 - 1 OCONEE NUCLEAR STATION CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT CRYSTAL RIVER UNIT NO. 3 Capsule Desicnation Insertion Withdrawal OCI-A End of 1st Cycle End of 7th Cycle OCI-B End of 7th Cycle End of 16th Cycle OCI-C End of 2nd Cycle End of lith Cycle OCI-D End of 9th Cycle End of 18th Cycle OCII-A End of 1st Cycle End of 2nd Cycle OCII-B End of 4th Cycle End of 9th Cycle OCII-D End of 9th Cycle End of 18th Cycle OCII-E End of ist Cycle End of 9th Cycle OCII-F End of 9th Cycle End of 18th Cycle OCIII-B End of ist Cycle End of 2nd Cycle OCIII-C End of 2nd Cycle End of 7th Cycle OCIII-D End of ist Cycle End of 9th Cycle OCIII-E End of 5th Cycle End of 18th Cycle OCIII-F End of lith Cycle End of 20th Cycle Note:
OCI __ Capsules are from Unit No.1 OCII _ Capsules are from Unit No. 2 OCIII-Capsules are from Unit No. 3 4.2-4 Amendments Nos. 44, 44 & 41