ML20115D337
| ML20115D337 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 07/10/1996 |
| From: | Jeffrey Mitchell NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20115D343 | List: |
| References | |
| NUDOCS 9607150123 | |
| Download: ML20115D337 (10) | |
Text
4
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4 UNITED STATES j
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20!M50001
.....,o I
VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 1
VERMONT YANKEE NUCLEAR POWER STATION i
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.147
{
License No. DPR-28 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment filed by the Vermont Yankee Nuclear Power Corporation (the licensee) dated April 4,1996, 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 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 l
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 Specifica-tions as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-28 is hereby amended to read as follows-I 1
9607150123 960710 PDR ADOCK 05000271 P
e
. t Technical Soecifications l
The Technical Specifications contained in Appendix A, as revised through Amendment No.147, are hereby incorporated in the license. The licensee l
shall operate the facility in accordance with the Technical Specifications.
J l
3.
This license amendment is effective as of its date of issuance, and l
shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION ede d NM Jocelyn A. Mitchell, Actirig Director i
Project Directorate I-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: July 10, 1996 i
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J 1
l
ATTACHMENT TO LICENSE AMENDMENT NO.147 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 152 152 155 155 155a 156 156 165 165 165a 169 169 e
i
i e
4 VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION AP is reduced to
<1.7) during required operabi-lity testing of the HPCI system pump, the RCIC system pump, the drywell-suppression chamber vacuum breakers, and the suppression chamber-reactor building vacuum breakers, and SGTS testing.
d.
If the specifica-tions of 3.7.A.9.a cannot be met, and the differential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condition in six (6) hours and a Cold Shutdown condition in the following eighteen (18) hours.
B.
Standby Gas Treatment System B.
Standby Gas Treatment System 1.
a.
Except as specified 1.
At least once per in Specification operating cycle, not to 3.7.B.3.a below, exceed 18 months, the i
whenever the following conditions reactor is in Run shall be demonstrated.
Mode or Startup l
Mode or Hot Shut-a.
Pressure drop
{
down condition, across the combined both circuits of HEPA and charcoal j
the Standby Gas filter banks is Treatment System less than 6 inches shall be operable of water at at all times when 1500 cfm 210%.
secondary contain-ment integrity is b.
Inlet heater input required.
is at least 9 kW.
b.
Except as specified in Specification 3.7.B.3.b below, whenever the reactor is in l
Refuel Mode or Cold Shutdown condition, both circuits of the Standby Gas i
Amendment No. &&,
+3, GG, &+3, 147 152 0
d VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION C.
Secondary Containment System C.
Secondary Containment System 1.
Surveillance of Integrity shall be maintained during the secondary containment following modes or shall be performed as follows:
conditions:
a.
Whenever the a.
A preoperational reactor is in the Run Mode, Startup secondary Mode, or Hot containment Shutdown condition; capability test shall be conducted or after isolating the b.
During movement of Reactor Building irradiated fuel and placing either assemblies or the Standby Gas fuel cask in Treatment System secondary filter train in operation.
Such containment or tests shall c.
During alteration demonstrate the of the Reactor capability to maintain a Core; or 0.15 inch of water d.
During operations vacuum under calm wind with the potential for draining the
( 2 < u < 5 mph )
condition with a reactor vessel.
filter train flow
^
rate of not more than 1500 cfm.
b.
Additional tests shall be performed during the first operating cycle under an adequate
-number of different environmental wind conditions to enable valid extrapolation of the test results, c.
Secondary containment capability to maintain a 0.15 inch of water vacuum under calm wind (2<G<5 mph) conditions with a filter train flow rate of not more than 1,500 cfm, shall be demonstrated at least quarterly and at each refueling outage prior to refueling.
Amendment No. H4, 147 155
~.. ~ _ -..
. _ ~
l VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION f
d.
Oparability testing l
of the Reactor i
Building Automatic Ventilation System isolation valves shall be performed in accordance with Specification 3
4.6.E.
2.
With Secondary 2.
Intentionally blank.
i Containment Integrity not maintained with the reactor in the Run Mode, Startup Mode,.or Hot Shutdown condition, restore Secondary Containment Integrity within four (4) hours.
3.
If Specification 3.7.C.2 3.
Intentionally blank.
cannot be met, place the reactor in the Hot Shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in the Cold Shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.
With Secondary 4.
Intentionally blank.
Containment Integrity not maintained during movement of irradiated fuel assemblies or the fuel cask in secondary containment, during alteration of the Reactor Core, or during operations with the potential for draining the reactor vessel, immediately perform the following actions:
a.
Suspend movement of irradiated fuel assemblies and the fuel cask in secondary containment; and b.
Suspend alteration of the Reactor Core; and-c.
Initiate action to suspend operations with the potential for draining the reactor vessel.
Amendment No. 147 155a
~
VYNPS 3.7 LIM ~ TING CONDITIONS FOR
[
OrF RATION 4.7 SURVEILLANCE REQUIREMENTS l
5.
Core spray ar" LPCI pump 5.
The core spray and LPCI l
lower compar' rent door lower compartment openings shall be closed l
at all times except checked closed daily.
openings shall be l
during passage or when l
temperature is less than 2120F.
D.
Primary Containment Isolation Valves Isolation valves l
1.
During reactor power 1.
Surveillance of the operating conditions all primary containment l
-isolation valves listed isolation valves should in Table 4.7.2 and all be performed as follows:
instrument line flow check valves shall be a.
The operable operable except as isolation 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.
I b.
Operability testing of the primary containment isolation valves shall be performed in accordance with Specification i
4.6.E.
c.
At least once per quarter, with the reactor power less than 75 percent of rated, trip all main steam isolation valves (one at a time) and verify closure time.
d.
At least twice per week, the main i
steam line isolation valves shall be exercised by partial closure and subsequent reopening.
Amendment No. %, M,
'rM, H4,147 156
4!
i VYNPS 3,AgJp 3.7 (Cont'd)'
The vacuum relief system from the pressure suppression chamber to l
l Reactor Building consists of two 100% vacuum relief breakers (2 parallel sets of 2 valves in series), operation of either system will maintain the pressure differential less than 2 psig; the external design pressure is 2 psig.
service there is no immediate threat With one vacuum breaker out of i
to accident mitigation or primary containment and, therefore, reactor operation can be continued for 7 days while repairs are being made.
The capacity of the ten (10) drywell vacuum relief valves is sized to limit the pressure differential between the suppression chamber and drywell during post-accident drywell cooling operations to the design i
limit of 2 psig. They are sized on the basis ef the Bodega Bay' l
The ASME Boiler and Pressure Vessel
.)
pressure suppression tests.
Code,Section III, Subsection B, for this vessel allows eight i
(8) operable valves, therefore, with two (2) valves secured, containment integrity is not impaired.
Each drywell-suppression chamber vacuum breaker is fitted with a redundant pair of limit switches to provide fail-safe signals to panel mounted indicators in the Reactor Building and alarms in the Control Room when the disks are open more than 0.050* at all points along the seal surface of the disk.
These switches are capable of transmitting the disk closed to open signal with 0.01" movement of the switch plunger.
Continued reactor operation with failed components is justified because of the redundance of components and circuits and, most importantly, the accessibility of the valve lever arm and position reference external to-the valve.
The fail safe feature of the alarm circuits assures operator attention if a line fault occurs.
-The requirement to inert the containment is based on the recommendation of the Advisory Committee on Reactor Saftguards. This recommendation, in turn, is based on the assumption that several percent of the zirconium in the core will undergo a reaction with steam during the loss-of-coolant accident. This reaction would release sufficient hydrogen to result in a flammable concentration in the primary containment building.
The oxygen concentration is therefore kept below 4% to minimize the possibility of hydrogen combustion.
1 General Electric has estimated that less than 0.1% of the zirconium would react with steam following a loss-of-coolant due to operation of emergency core cooling equipment.
This quantity of zirconium would not liberate enough hydrogen to form a combustible mixture.
B. and C. Standbv Gas Treatment System and Secondary containment System The secondary containment is designed to minimize any ground level release of radioactive materials which might result from a serious accident. The Reactor Building provides secondary containment during reactor operation, when the drywell is sealed and in service; the Reactor Building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling.
Because the secondary containment is an integral part of the complete containment system, secondary containment is required at all times that primary i
containment is required except, however, for initial fuel loading and low power physics testing.
j In the Cold Shutdown condition or the Refuel Mode,the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these conditions. Therefore, maintaining Secondary Containment Integrity is not required in the Cold Shutdown I
condition or the Refuel Mode, except for other situations for which Amendment No. 44, C;;;; Ch;;;;, 110, 147 165
VYNPS R.Aggg,:
3.7 (Cont'd)
A significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel, during alteration of the Reactor Core, or during movement of irradiated fuel assemblies or the fuel cask in the secondary containment.
With the reactor in the Run Mode, the Startup Mode, or the Hot Shutdown condition, if Secondary Containment Integrity is not maintained, Secondary Containment Integrity must be restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during the Run Mode, the Startup Mode, and the Hot Shutdown condition. This time period also ensures that the probability of an accident (requiring Secondary Containment Integrity) occurring during periods where Secondary Containment Integrity is not maintained, is minimal.
If Secondary Containment Integrity cannot be restored within the required time period, the plant must be brought to a mode or condition in which the LCO does not apply.
Movement of irradiated fuel assemblies or the fael cask in the secondary containment, alteration of the Reactor Core, and operations with the potential for draining the reactor vessel can be postulated to cause fission product release to the secondary containment.
In such cases, the secondary containment is the only barrier to release of fission products to the environment. Alteration of the Reactor Core and movement of irradiated fuel assemblies and the fuel cask must be immediately suspended if Secondary Containment Integrity is not maintained.
Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position.
Also, action must be immediately initiated to suspend operations with the potential for draining the reactor vessel to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until operations with the potential for draining the reactor vessel are suspended.
Amendment No.147 165a
()
VYNPS JAffd:
4.7.(Cont'd)
At the end of each refueling cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by at least 1 psi with respect to the suppression chamber pressure and held constant. The 2 psig set point will not be exceeded. The subsequent suppression chamber pressure transient (if l
any) will be monitored with a sensitive pressure gauge.
If the drywell pressure cannot be increased by 1 psi over the suppression chamber pressure it would be because a significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed.
If the drywell pressure can be increased by 1 psi over the suppression chamber the rate of change of the suppression chamber pressure must not exceed a rate equivalent to the rate of leakage from the drywell through a 1-inch orifice.
In the event the rate of change exceeds this value then the source of leakage will be identified and eliminated before power operation is resumed.
The drywell-suppression chamber vacuum breakers are exercised in accordance with Specification 4.6.E and immediately following termination of discharge of steam into the suppression chambar. This monitoring of valve operability is intended to assure that valve operability and position indication system performance does not degrade between refueling inspections. When a vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights are designed to function as followc:
Full Closed 2 White - On (Closed to 10.050" open)
Open 2 White - Off
(>0.050" open to full open)
During each refueling outage, two drywell-suppression chamber vacuum breakers will be inspected to assure sealing surfaces and components have not deteriorated.
Since valve internals are designed for a 40-year lifetime, an inspection program which cycles through all valves in one-eighth of the design lifetime is extremely conservative.
Experience has shown that a weekly measurement of the oxygen concentration in the primary containment assures adequate surveillance of the primary containment atmosphere.
B. and C.
Standby Gas Treatment System and Secondary Containment System Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 0.15 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leakage tightness of the reactor building, and performance of the standby gas treatment system.
The testing of reactor building automatic ventilation system isolation valves in accordance with Technical Specification 4.6.E demonstrates the operability of these valves.
In addition, functional testing of initiating sensors and associated trip channels demonstrates the capability for automatic actuation.
Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system performance capability.
Amendment No. M,147 169