ML19309F612
| ML19309F612 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/07/1980 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309F609 | List: |
| References | |
| NUDOCS 8004300168 | |
| Download: ML19309F612 (11) | |
Text
800 coa /d W
O e naos 8['s Io UNITED STATES
+
g NUCLEAR REGULATORY COMMISSION a
WASHINGTON, D. C. 20565 g
g
\\
/
DUQUESNE LIGHT COMPANY OHIO EDIS0N COMPANY PENNSYLVANIA POWER COMPANY DOCKET N0. 50-334 BEAVER VALLEY POWER STATION, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 27 License No. DPR-66 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Duquesne Light Conpany, Ohio Edison Company, and Pennsylvania Power Company (the licensees) dated February 15, 1980 as supplemented February 21, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the 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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.
l
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-66 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
27, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION W &'
A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: April.7, 1980 l
l l
i
ATTACHMENT TO LICENSE AMENDMENT NO.27
~
FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Remove Pages Insert Pages V
V 3/4 4-28 3/4 4-28 3/4 4-29 3/4 4-29 3/4 4-30a 3/4 4-30b 3/4 4-30c 3/4 4-30d 3/4 4-30e 3/4 4-30f 3/4 4-30g 3/4 4-30h B 3/4 4-10 B 3/4 4-10 B 3/4 4-11 L
i
INDEX LIMITING CONDITIONS FOR OPER' 10N AND SURVEILLANCE REQUIREMENTS Page SECTION 3/4.4.2 SAFETY VALVES - SHUTD0WN...............................
3/4 4-5 3/4.4.3 SAF ET Y VALV ES - 0 PE RATI NG..............................
3/4 4-6 3/4 4-7 3/4.4.4 PRESSURIZER............................................
3/4.4.5 STEAM GENERATORS.......................................
3/4 4-8 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4 4-11 Leakage Detection Systems..............................
Operational Leakage....................................
3/4 4-13 3/4.4.7 CHEMISTRY..............................................
3/4 4-15 3/4.4.8 SPECIFIC ACTIVITY......................................
3/4 4-18 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................
3/4 4-22 Pressurizer............................................
3/4 4-27 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components..................
3/4 4-28 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4. 5.1 ACCUMULATORS...........................................
3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T
> 350 F.........................
3/4 5-3 avg --
ECCS SUBSYSTEMS - T,yg < 350 F...'......................
3/4 5-6 3/4.5.3 3/4.5.4 BORON INJECTION SYSTEM Baron I nj ec ti on Ta nk...................................
3/4 5-7 Heat Tracing...........................................
3/4 5-8 3/4.5.5 REFUEllhG WATER STORAGE TANK...........................
3/4 5-9 BEAVER VALLEY - UNIT 1 V
Anendment No. 27 l
INDEX LIMITING CONDITIONS FOR OP 1ATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity...................................
3/4 6-1 Containment Leakage.....................................
3/4 6-2 Co n ta i nme n t Ai r Loc ks...................................
3/4 6-5 Internal Pressure.............
3/4 6-6 Air Temperature.........................................
3/4 6-8 Containment Structural Integrity........................
3/4 6-10 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System.........................
3/4 6-11 Containment Reci rcul ation Spray System.................. 3/4 6-13 Chemical Addition System................................
3/4 6-15 3/4.6.3 CONTAINMENT ISOLATION VALVES............................
3/4 6-17 3/4.6.4 COM8USTIBLE GAS CONTROL Hydrogen Analyzers......................................
3/4 6-20 i
Electric Hydrogen Recombiners...........................
3/4 6-21 i
Hydrogen Purge System...................................
3/4 6-23 l
l 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM S team Je t Ai r Ej ector...................................
3/4 6-25 Mechanical Vacuum Pumps.................................
3/4 6-26 BEAVER VALLEY - UNIT 1 VI
REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
A maximum heatup of 100'F in any one hour period, a.
b.
A maximum cooldown of 200*F in any one hour period, and A maximum spray water temperature differential of 320'F.
c.
APPLICABILITY: At all times.
ACTION:
With the pressurizer tenperature limits in excess of any of the above limits, restore the tenperature to within the limits within 30 minutes; perform an analysis to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.
l The spray water temperature differential shall be determined to be I
within the limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during steady state operation.
l l
l 17 BEAVER VALLEY - UNIT 1 3/4 4-27 Amendment No.
REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.
APPLICABILITY: All MODES ACTION:
With the structural integrity of any ASME Code Class 1 component (s) a.
not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations.
b.
With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requiremants, restore the structural integrity of the affected compone-ti, to within its limit or isolate the affected component (>
etior to increasing the Reactor Coolant System temperature above 200*F.
With the structural integrity of any ASME Code Class 3 component (s) c.
not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.
l d.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.10 Each ASME Code Class 1, 2, and 3 component shall be demon-strated OPERABLE in accordance with Specification 4.0.5 and by performing the following augmented inservice inspection program:
BEAVER VALLEY - UNIT 1 3/4 4-28 Amendment No. 27 s
~+
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) a.
At the first refueling outage, dye penetrant inspections shall be performed on all six reactor vessel nozzles and volumetric inspections shall be performed on at least two nozzles in the same manner as that conducted in the baseline inspection, b.
At the second and third refueling outages dye penetrant inspections shall be performed on all six nozzles and volumetric inspections shall be performed on at least two nozzles in the same manner as the baseline inspections so that all six nozzles shall be volumetrically inspected during the first three refueling outages.
c.
Defects found during these inspections shall be evaluated to determine their significance and reported to the Commission pursuant to Technical Specification 6.9.1.8.c.
BEAVER VALLEY - UNIT 1 3/4 4-29 Amendment No. 27 f
.a
l i
THIS PAGE LEFT INTENTIONALLY BLANK.
l l
l I
i BEAVER VALLEY - UNIT 1 3/4 4-30
TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS E
50 FT-LB/35 MIN. UPPER SHELF RT DT FT-LB 9
MATERIAL CU P
NDTT MIL TEMP F N
COMPONENT TYPE
- F LONG TRANS F
LONG TRANS y
CL. HD Dome A533,8,C1.1
-40 40 77*
17 115 75**
CL. HD. SEG.
A533,8,C1.1
-20 18 59*
-1 118 77**
i E
.HD. FLG.
A508,C1.2 10
-25
-5*
10 100 65**
[
VESSEL FLG.
A508,C1.2 60 21 56 60 151 98**
INLET N0ZZLE A508,C1.2 60$ 105 200*
140 82 53**
INLET N0ZZLE A503,C1.2 60$
58 125*
65 94 61**
INLET N0ZZLE A508,C1.2 60$
72 148*
88 94 61**
OUTLET N0ZZLE A508,C1.2 60$
12 67*
60 95 62**
g
[
OUTLET N0ZZLE A508,C1.2 60$
98 107*
60 112 73**
a OUTLET N0ZZLE A508,C1.2 60$
55 112*
60 103 67**
UPPER SHELL A508,C1.2 40 10 37*
40 150 98**
INTER. SHELL A533,8,C1.1 0.14 0.015 -10 57 103 43 123 81 INTER. SHELL A533,8,C1.1 0.14 0.015 -10 60 83 23 125 81 SURV 75 87 27 LOWER SHELL A533,8,C1.1 0.20 0.010 -50 LOWER SHELL A533,8,C1.1 0.14 0.015 -20 33 80 20 127 75 TRANS. RING A508,C1.1 30 30 60*
30 139 90**
BT. HD. SEG.
A533,8,C1.1
-30 11 46*
-14 122 79**
BT. HD. DOME A533,8,C1.1
-50 15 48*
-12 122 79**
WELD WELD 0.26 0.0.8 -60
-16
-60 103 128 20
-40 HAZ HAZ
-40 Estimated (77 FT-Lb/54 mil Temp. for Longi. Data) 65% of Longitudinal Upper Shelf)is less)
- Estimated 60 F or 100 FT-lb Temp. or what
$ Estimated SURV - Mater al for surveillance program according to ASTM-185-70.
=._-.
REACTOR COOLANT SYSTEM BASES i
l vessel inside radius are essentially identical, the measured transition I
shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ART DT determined from the surveillance capsule is different for the equivalent capsule radiation exposure.
from the calculated 4RTNDT The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature require-ments of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-3 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
L The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the i
pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
I 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Comnission pursuant to 10 CFR Part 50.55a(g)(6)(1).
\\
i BEAVER VALLEY - UNIT 1 B 3/4 4-10 Amendment No. 27 m
r p
u.
--p7 e-,