ML20203L573
| ML20203L573 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/21/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20203L537 | List: |
| References | |
| NUDOCS 8608280047 | |
| Download: ML20203L573 (6) | |
Text
- Keog 3
UNITED STATES 4
[
g NUCLEAR REGULATORY COMMISSION 5
'~j WASHINGTON, D. C. 20555
%,...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 97 TO PROVISIONAL OPERATING LICENSE NO. DPR-20 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET N0. 50-255
1.0 INTRODUCTION
By application dated March 17, 1986, the Consumers Power Company (the licensee) requested a change to the Palisades Technical Specifications, Sections 3.1.2 and 3.1.3.
The proposed revisions provide new reactor vessel pressure-temperature limits for heat-up, cool-down and the inservice hydrostatic test.
The last surveillance capsule report submitted to the staff by the licensee was Westinghouse Report WCAP-10637, entitled " Analysis of Capsule T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program." This report was submitted to the NRC in a letter from B. D. Johnson to H. R. Denton dated October 31, 1984.
2.0 EVALUATION Pressure-temperature limits must be calculated in accordance with the requirements of Appendix G,10 CFR Part 50, which became effective on July 26, 1983. Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G,10 CFR Part 50, are dependent upon the initial RT
- for the limiting materials in the beltline and closure flange r$hIons of the reactor vessel and the increase of RT r
beltline materNIs.esulting from neutron irradiation damage to the The Palisades reactor vessel was procured to ASME Code requirements that were in effect at the' time of procurement. However, Code requirem,ents at that time did not specify fracture toughness testing to determine the RTHeb,foreachofthematerialsneededtomakethereactorvessel.
the initial RT f r materials in the closure flange and beltline region of th$0feactor vessel could not be determined in accordance with the test requirements of the current ASME Code. Therefore, the initial RT for these materials is estimated from test data from other similar ma@ials used for fabrication of reactor vessels in the nuclear industry. The licensee indicates that the limiting closure flange region
- Formulae contained within this text are defined in the enclosed table.
8608280047 860821 PDR ADOCK 05000255 P
~
, materials were forgings, which were fabricated to ASME Code SA 508 C12 requirements.
The licensee has estimated the RT for these materials inaccordancewithBranchTechnicalPositionMTEb2,"FractureToughness Requirements," which are contained in NUREG-0800, "USNRC Standard Review Plan 5.3.2, Pressure-temperature Limits." This branch technical position provides conservative estimates of RT for reactor vessel materials and uses a RT f 60*F for the closure NInge forgings.
NDT The limiting materials in the reactor vessel ~ beltline are weld metals, which were fabricated by Combustion Engineering using the submerged arc weld process with RACO 3 and MIL B-4 Modified (Mn Mo Ni) weld wires.
However, records clearly indicate that the RAC0 3 weld wire had the addition of Nickel-200 and not the Mil B-4 Modified.
In all submerged arc welds, the flux utilized was Linde 1092. The initial RT for these weldmaterialswasestimatedbythelicenseeas-56*Fwitha"Nandard deviation of 17*F.
These initial RT and standard deviation values wererecommendedbythestaffinCombIsionReportSECY82-465,
" Pressurized Thermal Shock," for welds fabricated by Combustion Engineering using Linde 1092 flux.
The increase in RT resulting from neutron irradiation damage was estimatedbytheIbnseeusingthemethoddocumentedinDraftRegulatory Guide 1.99, Revision 2, " Radiation Damage to Reactor Vessel Materials."
Although this regulatory guide is only a draft, its methodology is considered by the staff to be the most up.-to-date method for predicting neutron irradiation damage.
This method of predicting neutr',n irradiation damage is dependent upon the predicted amount of neutron fluence and the amounts of copper and nickel in the beltline material.
The licensee has conducted a detailed search of vessel and surveillance fabrication records at Combustion Engineering to determine the heats of wire used in their reactor vessel beltline and their surveillance welds.
As a result of this search, the licensee indicates that the surveillance weld was fabricated using heats of wire, which were different from those used in the fabrication of the beltline welds.
The search confirmed that RACO 3 heat numbers W5214 and 348009 and MIL B-4 Mod (Mn Mo Ni) heat number 27204 were utilized to fabricate the reactor vessel beltline.
During fabrication of the reactor vessel, chemical analyses of the beltline welds were not. performed.
However, the licensee in Attachment III to their June 14, 1985 letter to the NRC has established the amounts of cooper and nickel in each of the beltline welds.
The amounts of copper and nickel were estimated from chemical analyses of reactor vessel surveillance welds and other nuclear vessel welds which were fabricated by Combustion Engineering using the same heats of weld wire as the Palisades beltline material.
Since the amount of copper and nickel should be the same within a heat of weld wire and the weld wire is the source of copper and nickel in a weld, the use of chemical analyses from surveillance welds and other nuclear vessel welds fabricated with the same heats of wire as the Palisades beltline weld should provide reliable estimates for the amounts of copper and nickel in the Palisades beltline welds.
The licensee's proposed pressure-temperature limits at 9 Effective Full Power Years (EFPY) have been calculated using a neutron fluence of 1.80 x 1018 n/cm2 (E>1MeV).
The amount of time required to accunulate this neutron fluence incident at the inner diameter of the reactor pressure vessel is dependent upon a radiological evaluation of the core and the reactor vessel.
Report WCAP-10637 contains a description of the radiological analyses performed by Westinghouse on the Palisades core and vessel.
These analyses result in a lead factor of 1.28 between the capsule and the vessel location receiving the highest neutron flux.
The Westinghouse radiological calculation predicts the end of life (2530 MWt for 32 effective full power years) peak neutron fluence to be 6.56 x 1928 n/cm2 (E>1MeV), when the axial peaking factor at the core midplane is 1.20.
The licensee has evaluated the previous core peaking factors at Palisades and found them to be 1.15.
An axial peaking factor of 1.15 yields an end of life peak neutron fluence of 6.29 x 1018 n/cm2 (E>1MeV).
Report WCAP-10637 contains the Westinghouse analysis of the dosimetry in Surveillance Capsule W-290.
The calculated peak neutron fluence at the end of life using the results from the Capsule W-290 dosimetry and the predicted load factor of 1.28 is 5.48 x 1018 n/cm2 (E>1MeV).
Since this peak neutron fluence from the Capsule W-290 dosimetry is less than the 6.29 x 1028 n/cm2 (E>1MeV) calculated using the Westinghouse radiological analysis, the value of 6.29 x 1018 n/cm2 (E>1MeV) will conservatively estimate the end of life neutron fluence for the Palisades reactor pressure vessel.
The longitudinal weld is taken as the most limiting material. The amount of copper and nickel in tne longitudinal weld is 0.19% by weight and 1.1% by weight, respectively.
The inside surface fluence is taken as 1.8 x 1019 n/cm2 (E>1MeV). The reactor vessel inside radius is 86 inches, and the outside radius is 94.5 inches which yields reactor vessel wall thicknass and weld thickness of 8.5 inches.
Flaws are postulated on the inside surface and the outside surface of the vessel or weld.
Distance is measured from the inner radius of the vessel outward.
The flaws on the inside surface and outside surface are referred to by location as 1/4 thickness and 3/4 thickness, respectively.
Recently, Draft Regulatory Guide 1.99, Revision 2, was revised to incorporate comments from the public.
No substantive changes in the Regulatory Guide c: curred.
However, the formula for attenuation of neutron fluence th.ough the vessel wall was revised.
Based on Regulatory Guide 1.99, Revision 2, the shift in RT at 1/4 thickness and3/4thicknessforthelongitudinalweldcomputbTby the staff and the licensee differ by a nominal 2% or less.
This difference in ARTNDT is due to the method used to attenuate neutron fluence through the vessel wall.
The staff finds the adjusted RT acceptable.
NDT computed by the licensee
,,-__,,,---,y__
The criteria from Section III, ASME Code, Article G-2000, Vessels, was used to determine the measured temperature during an inservice hydrostatic test (K 1.5K
+K At the flaw on the inside surfaceofthelongNubinaldEld,be). temperatures computed by the staff and the licensee are essentially the same.for the hydrostatic test pressure of 2310 psig.
The minimum criticality temperature of 371*F computed by the licensee is acceptable.
The family of curves for pressure-temperature limits for inservice hydrostatic testing is acceptable to the staff.
The criteria from Section III, ASME Code, Article G-2000, Vessels, was also used to determine the measured temperature during various heat-up and cool-down rates (K 2K K
of the reactor vessel.
For heat-up, eithertheflawinthe90l$gituNn+aldh)dontheinsidesurface(1/4 l
thickness) or the outside surface (3/4 thickness) is limiting.
For cool-down, the flaw in the longitudinal weld on the inside surface (1/4 thickness) is limiting.
The family of heat-up and cool-down curves (O F/hr through 100*F/hr) computed by the licencee either conservatively bound the values calculated by the staff or differ by 2% or less.
The differences in computed temperature values between the licensee and the staff are due to (1) the expressions used to attenuate the neutron fluence, (2) the methodology used to determine the temperature gradient through the vessel wall, and (3) whether K is disregarded when it I
Thus,thektafffindsthe would be conservative to do so.
pressure-temperature limits for the family of heat-up and cool-down curves presented by the licensee to be acceptable.
Figures 3-1 and 3-2 of the licensee's submittal do not include pressure-temperature limits for heat-up and cool-down during core operations; that is, these figures are for the reactor noncritical during heat-up and cool-down.
Pressure-temperature limits for heat-up and cool-down during core operations are obtained by adding 40 F to the temperature values in Figures 3-1 and 3-2.
The resulting temperature must be greater than or equal to the minimum criticality temperature, 371*F.
In the submittal of March 17, 1986 the licensee requested changes to the Technical Specifications, Sections 3.1.2 and 3.1.3.
The staff has l
reviewed each of the proposed changes and finds them all acceptable.
l The staff has used the method of calculating pressure-temperature limits in USNRC Standard Review Plan Section 5.3.2, NUREG-0800, Rev. 1, July 1981, to evaluate the proposed pressure-temperature limits.
The amount of neutron irradiation damage to the beltline materials was estimated using the method documented in Draft Regulatory 1.99, Revision 2.
Amounts of copper and nickel reported in Attachment III to the licensee's letter dated June 14, 1985 and an end of life peak neutron fluence of 6.29 x 1018 n/cm2 (E>1MeV) were used.
Our conclusion is that the proposed pressure-temperature limits meet the safety margins of Appendix G, 10 CFR Part 50, for 9 (EFPY) based on a fluence of 1.8 x 1018 n/cm2 (E>1MeV) and may be incorporated into the plant's technical specifications.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in 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 Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
We have 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 the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: August 21, 1986 Principal Contributor:
Robert J. Wright Enclosure l
. s FORMULAE DEFINITIONS (1)
NDT =~ Nil-Ductility transition (2)
RtNDT = Reference Temperature (*F) as defined in the ASME code.
(3) ARTNDT = Reference Temperature Shift (*F)
'(4)
K, Membrane Hoop Stress Intensity Factor KSI J in
=
g (5)
K
=
IR Reference Stress Intensity Factor KSI J in (6)
K
=
It Thermal Stress Intensity Factor KSI J in t
O a
- t i
e w
3---.
e e----m-3 g
.a..- - - -. - - -. -
--g-.--+y-
-.----. - + -- - - -
y-e--
,