ML20197A602
| ML20197A602 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 02/26/1998 |
| From: | Thomas C NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20197A606 | List: |
| References | |
| NUDOCS 9803090305 | |
| Download: ML20197A602 (12) | |
Text
___--_______ _ ____ _ _ - _ _
)
ja tt
[
t UNITED 6TATES g
}
NUCLEAR REGULATORY COMMISSION WASHINGTON, DO. 30606 4 001
- %...,,./
VERMONT YANKEE NUCLEAR POWER COff_QRATION QQCKET NO 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.152 License No. OPR 28 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applicat!on for the amendment by the Vermont Yankee Nuclear Power Corporation (the licensee) dated July 11,1997, as supplemented November 21, December 22,1997, and February 6,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules arid regulations; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and reguletions of the Commission; C.
There is reasonable assurance (l) that the activities authorized by this amendment can be conducted without endangering the health and safety of the puh?ic, 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.
9803090305 980226 PDR ADOCK 05000271 P
PDR u
o
is i
2 2.
Accordingly, the license is ant nc 3d by changes to the Technical Specifications as indicated in the attachment to this license amedment, and paragraph 3.B of Facility Operating License No. DPR-28 is hereby ame; ded to read as follows:
(B)
Technical Scecifications The Technical Specifications containet h Appendix A, as revised through Amendment No.152, are hereby incrp : rted in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendmet. is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION i
l c$.8 O. O w1M Cecil O. Thomas, Director Project Directorate 1-3 Division of Reactor Projects - t/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specification Date of issuance: February 26, 1998 l
i 4
l
i e
ATTALe i:..a r TO LICENSE AMENDMENT NO.152 FACILITY OPERATING LICENSE NO. DPR 28 ClOCKET NO. 50-271 Replace the following pages of Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment nurrber and contain verticel l nos indicating the areas of change.
Remove jnant 147 147 156 156 157 157 158 158 159 159 160 160 161 161 168 168 279 279 l
l g
'6-
-i',
e VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION e.
Minirum Water Volume -
68,000 cubic feet f.
Maximum Water Vo'.ume -
70,000 cubic feet 2.
The primary containment integrity shell be integrity shall be maintained at all times demonuLrated as required when the reactor is by the Primary critical or when the Containment Leak Rate i
-reactor water Testing Program -
temperature is above (PCLRTP).
2120F and fuel is in the reactor vessel except while performing low power physics tests at atmospheric ptessure at power levels not to exceed 5 Mw(t).
3.
If a portion of a system 3.
(Blank) l thet is considered to be~
an extension of primary containment is to be opened, isolate the affected penetration flow path by use of at least one closed and deactivated automatic valve, closed n.anual valve or_ blind flange.
4.
Whenever primary containment is required, 4.
The leakage from any one l l
t.he leakage from any one main steam line main steam line isolation valve shall-isolation valve shall not exceed 11.5 scf/hr not exceed 15.5 Jef/hr at 24 psig (Pt).
Repair at 44 psig (P ),
and retest shall be conducted to insure compliance.-
i Amendment No. 50,152
- 47 l
J
C, e
e VYNPS 3.7 LIMITING CONDITIG?iS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION 5.
Core spray and LPCI pump 5.
The core spray and LPCI lower compartment door lower compartment openings shall be closed openings shall be at all times except checked closed daily.
during possage or when reactor roolant temperature ia less than 2120F.
D.
Primaty Containment D.
Primary Containment Isolation Valves Isolation Valves 1.
During reactor power 1.
Surveillance of the operatir.g conditions all primary containment l
containment isolation isolation valves should valves and'all be performed as follows:
instrument line flow check valves shall be a.
The operrble operable except as isolatior, valves specified in that are power Specification 3.7.D.2.
operated and automatically initiated shall be tested for automatic initiation and the closure times specified in Table 4.7.2 at least once per operating cycle.
b.
Operability testing of the primary containment isolation valves shall be perforced in accordance with Specification 4.6.E.
c.
At least once per quarter, with the reactor power less thar. 75 percent of rated, trip all main steam isolation valves (one at a time) and verify closure time.
d.
At least cwice per week, the main steam line isolation valves shall be exercised by partial closure and subsequent reopening.
Amendment No. M, M, MB, H4, M, 152 156
8' e,
VYNPS 3.7.-LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION 2.
In the event any 2.
Whenever a containment l
containment isolation isc,1ation valve is-
-l valve becomes inoperable, the position inoperable, reactor.
of at least one other power operation may.
valve in each line continue provided at having an inoperable least one containment valve shall be logged isolation valve in each daily.
line having an
-inoperable valve is in the mode corresponding to the isolated condition.
3.
If Specifications 3.7.D.1 and 3.7.D.2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I J
l Amendment No. 152 157
n, VYNPS l
Thir page intentionally left blank.
Amendment No.152 158
e
'N VYNPS l
TABLE 4.7.2 PRIMARY CONTAINMENT ISOLATION VALVES 1
Number of Power i
Operated Valves Maximum Action on l
Isolation Operatir*q Normal Initiating Group (1)_
Valve Identification Inboard Outboard Time (sec)
Position Sicnal 1
Main Steam Line Isolation (2-90A, D&
4 4
5 (Note 2)
Open CC 2-86A, D) 1 Main Steam Line Drain f2-74, 2-77) 1 1
35 Closed SC 1
Recirculation Loop Sample Line (2-39, 2-40) 1 1
5 Closed SC l
2 RHR Discharge to Radwaste (10-57, 10-66) 2 25 Closed SC 2
Drywell Floor Drain (20-82, 20-83) 2 20 Open GC 2
Drywell Equipment Drain f20-94, 20-95) 2 20 Open GC 3
Drywell Air Purge Inlet 19-9) 1 10 Closed SC l
3 Drywell Air Purge Inlet
?)
1 10 Closed SC 3
Drywell Purge & Vent Out, i-19-7A) 1 10 Closed
- SC 3
Drywell Purge & Vent Outlet Bypass 1
10 Closed SC (16-19-6A) 3 Prywell & Suppression Chamber Main Exhaust 1
10 Closed
- SC (16-19-7) 3 Suppression Chamber Purge Supply (16-19-10) 1 10 Cloaed SC 3
Suppression Chamber Purge & Vent Outlet 1
10 Closed SC (16-19-7B) 3 Suppression Charrber Purge & Vent Outlet 1
10 Open GC Bypass (16-19-6B)
Valves 16-19-7 and 16-19-7A shall have stops installed to limit valve opening to 500 or less.
159 Amendment No. 54, M, R, 152
.m
'l TABLE 4.7.2 (Cont'd)
PRIMARY CONTAINMENT ISOLATION VALVES Number of Power Operated Valves Maximum Action on
. Isolation Operating Normal Initiating.
Group (1)
Valve Identification Inboard Outboard Time-(s T)
Position Sianal 3
Exhaust to Standby Gas Treatment System 1
10 Open GC (16-19-6) l 3
Containment Purge Supply (16-19-23) 1 10 Closed SC-l 3
Containment Makeup Supply'(16-20-22A7 1
NA Closed SC 3
Containment Makeup Supply.(16-20-20, 7
5' open GC 16-20-22B) 5 Reactor Cleanup System (12-15, 12-18) 1 1
25 Open GC 6
HPCI (23-15, 23-16)
'l 1
55 Open GC 6
RCIC (13-15, 13-16) 1 1
20 Open GC.
Primary / Secondary Vacuum Relief (16-19-11A, 2
NA Closed SC 16-19-11B)
Primary /Fecondary Vacuum Relief (16-19-12A, 2
ifA Closed Process 16-19-12B) 3 Containment Air Sampling (VG 23, VG 26, 4
5 Open GC 109-76A&B)
Feedwater Check Valves (72-27A, -96A, -28A, NA
'Open
. Process
-28B)
Amendment No. 58, M, M, %, M3,152 160
= - -.
>o
+
,2
~j
. (
TABLE 4.7.2 (Cont'd)
PRIMARY CONTAINMENT ISOLATION VALVES Number of Power Operated Valves Maximum Action on i
Isolation Operating Normal Initiatirg Group (1)
Valve Identificatic3 Inboard Outboard Time (sec)
Position Signal 2
RHR Return to Suppression' Pool (10-39A, B) 2 70 Closed SC 2
RHR Return to Suppression Pool (10-34A, B) 2 120 Closed SC 2
RHR Drywell Spray (10-26A, B & 10-31A, B) 4 70 Closed SC 4
2 RHR Suppression Chamber Spray (10-38A, B) 2 45 Closed SC 3
Containment Air Compressor Suction (72-38A, 2
20 Open GC D) ~
4 RHR Shutdown Cooling Supply (10-18, 10-17) 1 1
28 Closed SC Standby Liquid Control Check Valves (11-16, 1
1 NA Closed Proc.
11-17)
Hydrogen Monitoring (109-75 A, 1-4; 10 NA NA.
NA 109-75 E-D, 1-2)
Sampling Valves - Inlet Hydrogen Monitoring (VG-24, 25, 33, 34) 4 NA NA NA I
1 E
These valves are remote' manual sampling valves which do not receive an isolation signal. Only one valve in f
each line is required to be operable.
}
4 i
i
?
Amendment No. 58, f4, 4-15, 24,15 2 161 P
4
-_ 7,_
o sg.
VYNPS BASES:
4.7 (Cont'd)
The maximum allowable test leak rate at the peak accident pressure of 44 psig (La) is 0.80 weight t per day.
The maximum allowable-test leak rate at the retest pressure of 24 psig (Lt) has been conservatively determined to be 0.59 weight percent per day. This value will be verified to be conservative by actual primary containment leak rate measurements at both 44 psig and 24 psig upon completion of the containment structure.
1 As most leakage and detsrioration of integrity is expected to occur through penetrations, especially those with resilient seals, a periodic leak rate test program of such penetration is conducted at the peak accident pressure of 44 psig to insure not only thet the leakage remains acceptably low but also that the sealing materials can withstand the accident pressure.
The Primary Containment Leak Rate Testing Program is based on Option B to 10CFR50, Appendix J, for development of leak rate testing and surveillance schedules for reactor containment vessels.
Surveillance of the suppression Chamber-Reactor Building vacuum breakers consists of operability checks and leakage tests (conducted as part of the containment leak-tight
'7 tests). These vacuum I
breakers are normally in the closed pos:-ton and open only during tests or an accident condition.
OperabiAity testing is performed in conjunction with Specificatior 4.6.E.
Inspections and calibrations are performed during the refui.ing outages this frequency being l
based on equipment quality, experience, and engineering judgment.
The ten (10) drywell-suppression vac.2m relief valves are designed to open to the full open position (the position that curtain area is equivalent to valve bore) with a force equivalent to a 0.5 psi differential acting on the suppression chamber face of the valve disk. This oraning specification assures that the design limit of 2.0 pr.id between the drywell and external environment is not exceeded. Once each refueling outage each valve is tested to assure that it will open fully in response to a force less than that specified. Also it is inspected to assure that it closes freely and operates properly.
The containment design has been examined to establish the allowable bypass area between the drywell and suppression chamber as 0.12 ft,
2 This is equivalent to one vacuum breaker open by three-eighths of an inch (3/8") as measured at all points aros'd the circumference of the disk or three-fourths of an inch (3/4") as mearured at the bottom of the disk when the top of the disk is on the seat. Since these valves open in a manner that is purely neither mode, a conservative allowance of one-half inch (1/2") has been selected as the maximum permissible valve opening. Assuming that permissible valve opening could be evenly divided among all ten vacuum breakers at once, valve open position assumed to indication for an individual valve must be activated less than fifty-thousandths of an inch (0.050") at all points along the seal surface of the disk. Valve closure within this limit may be determined by light indication from two independent position detection and indication systems. Either system provides a control room alarm for a nonseated valve.
Amendment No. GO, 444,152 168
i 9i VYNPS 4.
An evaluation of the change, which shows the predicted releases of radioactive materials in Z iquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; 5.
An evaluation of the change, which shows the expected maximum exposures to member (s) of the public at the site boundary and to the general population that differ from those previously estimated in the license application and amendments thereto 6.
A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; 7.
An estimate of the exposure to plant operating personnel as a result of the change; and 8.
Documentation of the fact that the change was reviewed and found acceptable by PORC.
B.
Shall become effective upon review and acceptance by PORC and approval by the Plant Manager.
6.15 Primary Containment Leak Rate Testino Procram A program shall be established to implement the leak rate testing of the primary containment as required by 10CFR50.54(o) and 10CFR50, Appendix J, Option B as modified by approved exemptions.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, entitled ' Performance Based Containment Leak-Test Program," dated September 1995.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 44 psig.
The aximum allowable primary containment leak rate, La, at Pa, shall be 0.8% of primary containment air weight per day.
Leak rate acceptance criteria are:
1.
Primary containment leak rate acceptance criterion i 1.0 La.
2.
The as-left primary containment integrated leak rate test (Type A test) acceptance criterion is 1 0.75 La.
3.
The combined local leak rate test (Type B and C tests) acceptance criterion is 1 0.60 La, calculated on a maximum pathway basis, prior to entering a mode of operation where containment integrity is required.
4.
The combined local leak rate tect (Type B and C tests) acceptance criterion is 1 0.60 La, calculated on a minimum pathway basis, at all times when primary containment integrity is required.
5.
Airlock overall leak rate acceptance criterion is s 0.10 La when tested at > Pa.
The provision of the Defirition (1.0.Y) for Surveillance Frequency does not apply to the test frequencies specified in the Primary Containment Leak Rate Testing Program.
Amendment No. sa, 95, 152 279
_