ML20034A818
| ML20034A818 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 04/17/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034A817 | List: |
| References | |
| GL-88-11, NUDOCS 9004240383 | |
| Download: ML20034A818 (4) | |
Text
.
.,8
}[i(
UNITED STATES g
NUCLEAR REGULATORY COMMISSION t
5 WASHINGTON, D. C. 20655 L
L
-*..+
l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION-i
- _n l
SUPPORTING AMENDMENT NO.120 TO FACILITY OPERATING LICENSE N0. DPR-28 VERMONT YANKEE NUCLEAR POWER CORPORATION I
e VERMONT YANKEE NUCLEAR POWER STATION.
DOCKET N0; 50-271-4 INTRODUCTION-By letter dated November 10, 1989, the Vermont Yankee Nuclear Power Corporation
?
(the licensee) requested an amendment to Facility Operating License No. DPR-28 for the Vermont Yankee Nuclear Power Station.. The proposed amendment would 1
revise the Presure-Temperature limit curves.in Technical Specification (TS)
Figure 3.6.1.
This proposed TS change reflects the shift in transition temperature for the-reactorvgsselmaterialsforoperationthroughacumulativeenergyoutputof-4'46 x 10 MWh(t). The change is necessary because the existing curves are 8
limited to a cumulative energy output of 1.79 x 10 MWh(t) which is expected to i
be reached by May 1990.
Previous to their request of November 10, 1989 the licensee responded.to our Generic Letter 88-11 "NRC Position on. Radiation Entrittlement of Reactor Vessel.
Materials" by a letter dated November 10, 1988. This Generic. Letter provided guidance for the calculation of the nil-ductility reference temperature of i_
reactor vessel beltline materials which relates to this Pressure-Temperature I
l limit curve change request as discussed below.
DISCUSSION The licensee has requested permission to revise the pre'ssure/ temperature (P/T) limits in the Vermont Yankee Nuclear Power Station Technical Specifications, Section 3.6.
The request was documented in a letter'from the= licensee dated November 10, 1989. This revision also changes the effectiveness of the P/T limits to 32 effective full power years (EFPY'). The proposed P/T limits.were developed based on Regulatory Guide (RG) 1.99, Revision 2. 'The : proposed revision provides up-to-date P/T limits for the operation of the reactor coolant system.
during heatup, cooldown, criticality, and hydrotest.
To evaluate the P/T limits, the staff used the following NRC regulations and i
guidance: Appendices G and H of 10 CFR Part 50; the ASTM Standards and the ASME Code, which are referenced in Appendices G and H; 10~ CFR 50.36(c)(2); RG 1.99,Rev.2;StandardReviewDlan(SRP)Section5.3.2;andGenericletter
{
88-11.
1 9004240383 900417 PDR ADOCK 05000271 P
PDC l
e The P/T limits are among the limiting conditions of operation in the Technical Spccifications for all comercial nuclear plants. Appendices G and H of 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P/T limits. An acceptable method for constructing the P/T limits is described in SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be i
tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards. These tests. define the extent'of vessel embrittlement at the time of capsule withdrawal in terms of the increase in.
reference temperature. Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter 88-11 requested that licensees use the methods in RG 1.99, Rev.
2, to predict the effect of neutron irradiation on reactor vessel materials.
This. guide defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to accour,t for uncertoir. ties in the prediction method.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program and to periodically withdraw surveillance capsules from the reactor-vessel. Appendix H refers to the ASTM Standards which, in turn, require that 1
the capsules be installed in.the vessel before startup and that.the test specimens made from plate, weld, and heat-affected-zone (HAZ) y.contain H
materials of the rector beltline, l
EVALUATION The staff evaluated the effect of reactor neutron irradiation embrittlement on l
each beltline material in the Vermont Yankee-reactor-vessel. The amount'of i
irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 32 EFPY was plate I-14 with 0.11% copper (Cu), 0.63% nickel (Ni), and an initial RTNDT of 40'F.
j The licensee has removed one surveillance capsule from the Vermont Yankee reactor vessel. The results from this capsule were published in the'Battelle-Columbus Laboratories report BCL-585-84-3. All surveillance capsules contained 1
Charpy impact specimens and tensile specimens made from base metal weld metal and HAZ.
For the limiting beltline material, plate I-14, the staff calculated the. ART to be 63.1 F at 1/4T (T = reactor vessel beltline thickness). and 55.5* F for-2 3/ATat33EFPY. The staff used a neutron fluence of 1.6E17 n/cm at 1/4T and 9E16 n/cm at 3/4T. The ART was determined using Section 1 of RG 1.99, Rev.
2, as only one surveillance capsule has been removed from Vermont Yankee.
The licensee used a more conservative safety factor then the one in RG 1.99, Rev. 2, to calculate an ART of 89 F EFPY at 1/4T and 73' F at 3/4T at 32 EFPY
oM s
i
/
t #
1 for a plate material in the beltline. The staff believes that an ART of 63.1' F l
is sufficient to protect the reactor vessel from embrittlenent. Substituting the ART of 63.1* F_into the equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and hydrotest meet the belt material requirement in Appendix G of 10 CFR Part 50.
5 In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials. Paragraph IV.A.2 of Appendix G states that when the pressure i
exceeds 20% of the preservice system hydrostatic test pressure, the
+
temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120' F for normal operation and by 90' F for hydrostatic pressure tests and leak tests.
Paragraph IV.A.3 of Appendix G states "an exception may be made for boiling water reactor vessels when water level is withinLthe normal range for power operation and the pressure is less than 20 percent of the pre-service system hydrostatic test pressure.
In this-case the minimum permissible temperature is 60' F (33' C) above-the reference temperature of the closure flange regions that are highly stressed by the bolt preload."'
Based on the flange reference temperature of 20' F, the staff has determined.
that the proposed P/T limits satisfy paragraph IV.A.2 of Appendix G.
Section IV.B of Appendix G requires that the. predicted Charpy USE at end of life be above 50 ft-lb. No USE data were available for plates 1-15. 1-16, and 1-17. However, Charpy impact data at 40* F were available. The lowest individual reading was for plate 1-17--ie., 65 ft-lb. Using-this'for a USE value and Figure 2 of RG 1.99, Rev. 2, it was predicted that the EOL USE would be 59.5 ft-lb. This is greater than 50 ft-lb and, therefore, is acceptable.
SUMMARY
The staff concludes that the proposed P/T limits for the reactor coolant o
l system for heatup, cooldown, leak test, and criticality are valid through 32
~
EFPY as the limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The licensee's submittal also satisfies Generic Letter '88-11 guidance as the licensee used the methods of RG 1.99, Rev. 2 to calculate the ART. Therefore, the proposed P/T limits are acceptable for incorporation into the Vermont Yankee Technical Specification.
P
.i y*
e ENVIRONMENTAL CONSIDERATION t
This amendment involves s' change in a requirement with respect to the installation or use of a facility component located within the restricted area as defined. in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of-any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously published a proposed finding that the amendment J
involves no significant hazards consideration and there has been no public-coment on such finding. Accordingly, this amendment meets the eligibility!
criteria for categor.ical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant i
to 10 CFR 51.22(b), no environmental impact statement or environmental assessment -
need be prepared in connection with the issuance of this amendment.
CONCLUSION The Comission made a proposed determination that the amendment involves no significant hazards consideration which was published-in the Federal
.l Register (55 FR 2449 ) on January 24, 1990 and consulted with the State of Vermont. No public coments were received and the State of Vermont did i
not have any comments.
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance.that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's-regulations, and (3)-the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
REFERENCES l
1.
Regulatory Guide-1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2, Pressure-Temperature Limits 3.
L. M. Lowery et al., " Final Report on Examination, Testing,'and Evaluation-of Irradiated Pressure Vescel Surveillance Specimens from the Vermont Yankee Nuclear Power Station to Yankee Atomic Electric Company,"
BCL-585-84-3, Battelle-Columbus Laboratories, May 15, 1984 4.
R. W. Capstick (VYNPC) to USNRC Document Control Desk,
Subject:
' Vermont Yankee Response to NRC Generic letter 88-11, November 10, 1988.
5.
W. P. Murphy (VYNPC) to USNRC Document Control Desk,
Subject:
Proposed Change to Revise the Reactor Vessel Pressure-Temperature Curves in-the Vermont Yankee Technical Specification (Generic Letter 88-11),
November 11, 1988.
Principal Contributor: John Tsao Dated:
APR 171990
--