ML20206N242
| ML20206N242 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/24/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Vermont Yankee |
| Shared Package | |
| ML20206N246 | List: |
| References | |
| DPR-28-A-093 NUDOCS 8607010413 | |
| Download: ML20206N242 (8) | |
Text
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
s y,
E WASHINGTON, D. C. 20555
't,.....,o VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 93 License No.
DPR-28 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Vermont Yankee Nuclear Power Corporation (the licensee) dated May 10, 1985, as supplemented November 21, 1985, 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 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; 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 Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. DPR-28 is hereby amended to read as follows:
8607010413 860624 PDR ADOCK 05000271 P
(2) Technical Specifications The Technical Specifications, contained in Appendix A, as revised through Amendment No.
93, 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 UCLEAR REGULATORY COMMISSION
,s' t
e, Daniel R. Mulle, Project Director BWR Project Directorate #2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: June 24, 1986
,,3 y-..--
ATTACHMENT TO LICENSE AMENDMENT N0.93 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Pages 111 111a 111b 116 117 118 i
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Amendment No. 62. W, 93 111
FIGURE 3.6.2 FAST NEUTRON FLUENCE (E* 1 NEV) AS A FUNCTION OF THERMAL ENERGY AND FULL F0WER YEARS iets e
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REFERENCE:
L. M. Lowry et al.
Examination. Testing, and Evaluation of j
Irradiated Fressure Vessel Surveillance Specimens from Vermont Yankee Nuclear Fower Station.
Bate 11e Columbus Laboratories Report #BCL-585-84-3, May 15,1984 1
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VYNPS i
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f 3.6 and 4.6 '.
A.
pressure and Teneerature Limitations All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system l
temperature and pressure changes. These cyclic loads are introduced by nomel load transients, reactor trips,
}
and startup and shutdown operations. The various categories of toed cycles used for design purposes are
]
provided in Section 4.2 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes s
i are limited so that the maximum specifled heatup and cooldown rates are consistent with the design assumptions i
and satisfy the stress limits for cyclic operation.
During heatup, the themel gredients in the reactor vessel well produce thermal stresses which vary from j
compressive at the inner well to tensile at the outer wall. These themal induced compressive stresses tend to alleviate the tensile stresses induced by their internal pressure. Therefore, a pressure-temperature curve 3
based on steady-state conditions (i.e., no themet stresses) represents a lower bound of all similar curves for
}
finite heatup rates when the inner well of the vessel is treated as the governing locations.
l i
The heetup analysis also covers the detemination of pressure-temperature limitations for the case in which the l
outer well cf the vessel becomes the controlling location. The thermal gradients established during heatup i
produce tensile stresses at the outer well of the vessel. These stresses are additive to the pressure induced l
tensile stresses which are already present. The thermal induced stresses at the outer well of the vessel are
}
tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner well cannot be defined. Sut -. r tly, for the j
cases in which the outer well of the vessel becomes the stress controlling location, each heatup rate of j
interest must be analyzed on an individual basis.
In order to prevent undue stress on the vessel nozzles and bottom head region, the recirc'ulation loop
~
temperatures should be within 50*F of each other prior to startup of an idle loop.
I The reactor vessel materials have been tested to determine their initial reference temperature all-ductility 1
transition temperature (RTEDT) of 40*F maximum. Reactor operation and resultant fast neutron (E >l Nev)
I irradiation will cause an increase in the RTypy. Therefore, an adjusted reference temperature can be
{
predicted using current industry practices and Vermont Yankee Surveillance program data.
(Regulatory Guide g
1.99, Revision 2, and Battelle Columbus Laboratory Report BCL 585-84-3, dated May 15, 1984. The pressure / temperature limit curve. Figure 3.6.1, includes predicted adjustments for this shift in RT EDT IDF operation through 1.79x108 leat(t), as well as adjustments for possible errors in the pressure and temperature sensing instnaments, t
i Amendment No. 62, ff, 93 117
.. -....~
-- - a
I The reference temperature of the closure fleese meterial uns determined by meterial testing and Dreach Technical Position NTER 5-2, " Fracture Toughness Requirements for Older plants". The closure flange is leested in a low n.utron fluence -.n. therefore no measursue.T,., suit is.s,eeted - the ufe of me,me.
j l
The actual shift in ETggy of the vessel meterial will be established periodically during operatium by removing l
and evaluaties, in sceordance with ASTN 8185-reactor vessel material irradiation surve111ance specimens installed nose the inside well of the resetor vessel in the core area. Since the neutron spectra at the
]
irradiation semples and vessel inside radius aco essentially identical, the measured transition shift for a i
semple can be applied with confidence to the adjacent section of the reactor vessel. Bette11e Celenhus 1
Laboratory Boyert BCL-585-84-3, dated May 15, 1944, provides this information for the ten-year surveillance
]
capsule. In order to estimate the meterial properties at the 1/4 and 3/4 T positions in the vessel plate, the shift in RTupf is determined in accordance with Begulatory Guide 1,99, Eevision 2.
The hestup and cooldoun f
curves sesst be recalculated when the4ETggr determined from the survelilance capsule is different from the q
calculated &RTygg for the equivalent capsule radiation esposure.
t l
The pressure-tenyerature ilmit lines, shoem on Figure 3.6.1, for reactor criticality and for laservice leek and hydrostatic testing have been provided to assure ceay11ance with the minimum temperature requirements of j
Appendix G to 10CFE50 for reactor criticality and for inservice leak and hydrostatic testing.
i
}
The number of reactor vessel teradiation surveillance specimens and the frequencies for ransving and testing j
these speel==== see provided to assure cosy 11ance with the requirements of Appendix N to CFR part 50.
'i i
Coolant Chemiett7
\\
A steady-state radiciodine concentration limit of 1.1p.ci of I-131 dose equivalent per gram of water in the j
Reactor coolant System can be reached if the gross radionetivity in the gaseous effluents is near the limit, as l
set forth in Specification 3.8.C.la, oc there is a failure or pretenged shutdoun of the cleanup domineraliser.
In the event of a steam line rupture outside the drywell, the BBC staff calculations show the resultant j
radiological dose at the site boundary to be less then 30 tem to the thyroid. This dose was 4
118 Amendment 50. (5, 93
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