ML20128G984
| ML20128G984 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 10/04/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20128G987 | List: |
| References | |
| NUDOCS 9610090199 | |
| Download: ML20128G984 (7) | |
Text
a rtcu UNITED STATES p
g
.j NUCLEAR REGULATORY COMMISSION S
WASHINGTON, D.C. 3000M001 o
49.....,o VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.150 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 August 9,1996, as supplemented by letter on September 17, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and 4
the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The f# ility 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:
(1) 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 comon 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.
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:
9610090199 961004 PDR ADOCK 05000271 P
d
.g.
J l
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.150, are hereby incorporated in the license. The licensee 1
shall operate the facility in accordance with the Technical Specifications.
i 3.
This license amendment is effective as of its date of issuance, and shall be implemented within 30 days.
I FOR THE NUCLEAR REGULATORY COMMISSION S. Singh Bajwa, Acting Director Project Directorate I-1 i
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation l
Attachment:
Changes to the Technical Specifications Date of Issuance: October 4, 1996 l
l l
j ATTACHMENT TO LICENSE AMENDMENT NO.150 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 l
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 6
6 12 12 l
13 13 142 142 l
227 227
VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 NIIEi_IHTEiEITI Aeolicability:
Acolicability:
Applies to the interrelated Applies to trip setting of the variable associated with fuel instruments and devices which thermal behavior.
are provided to prevent the nuclear system safety limits from being exceeded.
obiective:
Obiective:
To establish limits below which To define the level of the the integrity of the fuel process variable at which cladding is preserved.
automatic protective action is initiated.
Specification:
Specification:
A.
Bundle Sefety Limit (Reactor A.
Trio Settinos Pressure >800 osia and Core Flow >10% of Rated)
The limiting safety system trip settings shall be as 4
a When the reactor pressure is specified below:
>B00 psia and the core flow is greater than 10% of 1.
Neutron Flux Trie rated:
Settinos l
1.
For the Cycle 19 core a.
loading, the existence Settina (Run Mode) of a Minimum Critical l
Power Ratio (MCPR) of When the mode switch j
less than 1.10 (1.12 for is in the RUN Single Loop Operation) position, the APRM shall constitute flux scram trip violation of the Fuel setting shall be as Cladding Integrity shown on Safety Limit (FCISL).
Figure 2.1.1 and Core loadings subsequent shall.bes to Cycle 19 will require recalculation of the Sf,,0. 66 (W-AW) + 544 MCPR.
where:
S = setting in percent of rated
-thermal power 4
(1593 MWt)
W = percent rated two loop drive flow where 2004 rated drive flow is that flow equivalent6 to 48 x 10 lbs/hr core flow Amendment No. H, 47, 64, H, 94, H9,150 s
. _ ~
. - = -
1~
VYNPS BASES:
1.1 TUEL CLADDING INTEGRTTY 4
A.
Refer to General Electric Company Licensing Topical Report, " General Electric Standard Application for Reactor Fuel,' NEDE-240ll-P-A (most recent revision).
The fuel cladding integrity Safety Limit (SL) is set such that no significant fuel damage is calculated to occur if the limit is not violated.
Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur.
Although it is recognized that the onset of
}
transition boiling woyld not result in damage to BWR fuel rods, the critical power at uhich boiling transition f.s calculated to occur has
)
been adopted as a convenient limit. However, the uncertainties in i
monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within i
the core and all uncertainties.
4 The MCPR SL is determined using a statistical model that combines all 1
the uncertainties in operating parameters and tha procedures used to calculate critical power.
The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations.
The MCPR fuel cladding integrity SL is increased for single loop operation in order to account for increased core flow measurement and TIP reading uncertainties, i
B.
Cpre Thermal Power Limit (Reactor Pressure 1 800 esia or Core Flow 110% of Rated) 1 1
At pressures below 800 psia, the core elevation pressure drop l
(0 power, O flow) is greater than 4.56 psi. At low power and all flows this pressure differential is maintained in the bypass re-itr.
2 of the core.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power 1
and all flows will always be greater than 4.56 psi. Analyses show 3 lbs/hr bundle flow, bundle pressure drop that with a flow of 2B x 10 is nearly independent of bundle power and has a value of 3.5 psi.
1 Thus, the bundle flow with a 4.56 psi driving head will be greater 3 lbs/hr irrespective of total core flow and independent than 2B x 10 of bundle power for the range of bundle powers of concern.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.
With the design peaking factors this corresponds to a core thermal power of more than 50%. Thus, a core thermal power limit of 25% for reactor pressures below 800 psia or core flow less than 10% is conservative.
C.
Power Transient Plant safoty analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 1.1.lA or 1.1.lB will not be exceeded. Scram times are checked periodically to assure the insertion timen are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux Amendment No. te, 49, 44, 94, 150 12
1 VYNPS
, BASES:
1.1 (Cont'd) following closure of the main turbine stop valves) does not necessarily cause fuel damnge. However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram af gnals are operable is supported by the extensive plant sefety analysis.
The computer provided with Vermont Yankee has a sequence annunciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur.
1 This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.
D.
Reactor Water Level (bhutdownCondition) 1 During periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of decay heat.
4 If reactor water level should drop below the top of the enriched fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding i
1 temperatures and clad perforation. The core can be cooled suf ficiently should the water level be reduced to two-thirds the core height.
Establishment of the safety limit at 12 inches above the top of the enriched fuel provides adequate margin. This level will be continuously monitored.
3 l
i i
J i
Amendment No. 64, 64, 150 13
1 1'
i i
UYNPS 1
i 1
R&jfl:
3.6 and 4.6 (Cont'd) 1 igurities will also be within their normal ranges. The reactor i
cooling samples will also be used to determine the chlorides.
Therefore, the sampling frequency is considered adequate to detect i
i long-term changes in the chloride ion content. -Isotopic analyses j
required by Specification 4.6.B.2 may be performed by a gsama scan and gross beta and alpha determination.
The conductivity of the feedwater is continuous,'y monitored and alarm j
set points consistent with Regulatory requirements given in l
Regulatory Guide 1.56, ' Maintenance of Water Purity in Boiling Water Reactors," have been determined. The results from the conductivity i
monitors on the feedwater can be correlated with the results from the i
conductivity monitors on the reactor coolant water to indicate j
j domineralizer breakthrough and subsequent conductivity levels in the
~
reactor vessel water.
C.
Coolant Leakagg i
i The 5 gpm limit for unidentified leaks was established assuming such i
leakage was coming from the reactor coolt t system. Tests have been conducted which demonstrate that a relation >h y exists between the size of a crack and the probability that the crack will propagate.
These tests suggest that for leakage somewhat greater than the limit specified for unidentified leakage; the probability is sus 11 that imperfections or cracks associated with such leakage would grow rapidly.
Leakage less than the limit specified can be detected within a few hours utilizing the available leakage detection systems.
If the limit is exceeded and the origin cannot be determaned in a reasonably short time the plant should be shutdown to allow further investigation and corrective action.
The 2 gpm increase limit in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period for unidentified leaks was established as an additional requirement to the 5 gym limit by Generic Letter 88-01. 'NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainlt is Steel Piping.'
The removal capacity from the drywell floor drain sug and the equivalent drain sump is 50 gym each. Removal of 50 gym from either of these sumps can h accomplished with considerable margin.
D.
Safety and Relief Valves Parametric evaluations have shown that only three of the four relief valves are required to provide a pressure margin greater than the recommended 25 psi below the safety valve actuation settings as well l
as maintaining the fuel cladding inte2rity safety limit for the limiting overpressure transient below 98% power. Consegmently, 95%
power has been selected as a limiting power level for,three valve operation. For the purposes of this limiting condition a relief valve that is unable to actuate within tolerance of its set pressure is considered to be as inoperable as a mechanically malfimetioning valve.
Experience in safety valve operation shows that a testing of 50% of the safety valves. per refueling outage is adequate to detect failures or deterioration. The tolerance value is specified in section III of the ASME Boiler and Pressure Vessel Code as 21% of design pressure.
An analysis has been performed which shows that with all safety valves set it higher the reactor coolant pressure safety limit of 1375 psig is not exceeded.
Change 16/ March 28,1974, H, 44, MS, H9,150 142
VYNPS 16fl.E:
3.11 FUEL RCDS A.
Aversee Planar Linear Heat Generation Rule (APLHGR)
Refer to the appropriate topical reports listed in Specification 6.7. A.4 for analyses methods.
(Note: All exposure increments in this Technical Specification section are expressed in terms of megawatt-days per short ton.)
I The MAPLEGR reduction factor of 0.83 for single recirculation loop operation is based on the assumption that the coastdown flow from the postulated large break, loop would not be available during a unbroken recirculation in the active recirculation loop, as discussed in NEDO-30060, " Vermont Yankee Nuclear Power Station Single Loop Operation." February 1983.
B.
Linear Heat Generation Rate (LHGR)
Refer to the appropriate topical reports listed in Specification 6.7.A.4 for analyses methods.
l C.
Minimum Critical Power Ratio (MCPR)
Operatine Limit MCPR 1.
The MCPR operating limit is a cycle-dependent parameter which can l
be determined for a number of different combinations of operating modes, initial conditions, and cycle exposures in order to provide reasonable assurance against exceeding the Fuel Cladding Integrity Safety Limit (FCISL) for potential abnormal occurrences. The MCPR operating limits are justified by the analyses, the results of which are presented in the current cycle's Core Performance Analysis Report. Refer to the appropriate topical reports listed in Specification 5.7.A.4 for analysis methods. The increase in MCPh operating limits for single loop operation accounts for increased core flow l
measurement and TIP reading uncertai% ties.
Amendment No. H, u, M H, H, M, MG, H4,150 227