ML17264A414
| ML17264A414 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/22/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17264A413 | List: |
| References | |
| NUDOCS 9603250353 | |
| Download: ML17264A414 (14) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SA TY EVA UATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PRESSURI ED THERMAL SHOCK EVALUATION ROCHESTER GAS AND CTRIC CORPORATION R.
GINNA NUCLEAR POWER PLANT DOCKET-NUMB R 50-24 1.0 JIIQTI By letter dated October 11,
- 1995, the Rochester Gas and Electric Corporation (the licensee) submitted for review and approval a pressurized thermal shock (PTS) assessment for the R.E.
Ginna reactor pressure vessel.
The licensee sent additional information in a letter dated December 21, 1995.
The PTS rule, 10 CFR 50.61, adopted on July 23,
- 1985, and revised on Hay 15, 1991, established screening criteria that are a measure of a limiting level of reactor vessel material embrittlement beyond which operation cannot continue without further plant-specific evaluation.
The screening criteria are given in terms of reference temperature RT>>$
The screening criteria are 270 'F for plates and axial welds, and 300 'F for circumferential welds.
The RT>>, is defined as:
RTvrs I + MTprs + H where:
(1) I is the initial reference temperature, (2) MT>>> is the mean value in the adjustment in reference temperature caused by Vrradiation, and (3)
H is the margin to be added to cover uncertainties in the initial reference temperature, copper and nickel content,
- fluence, and calculational procedures.
The initial reference temperature is the measured unirradiated value as defined in the ASIDE Code, Paragraph NB-2331.
If measured values are unavailable for the heat of material of interest, generic values may be used.
The generic values are based on the data for materials of all heats that were made by the same vendor using similar processes.
The generic values of initial reference temperature for welds are defined in the PTS rule.
The MT>>s depends upon the amount of neutron irradiation and the amounts of copper and nickel in the material; it is calculated as the product of a fluence factor and a chemistry factor.
The fluence factor is calculated from the best-estimate neutron fluence at the interface of cladding,
- weld, and metal on the inside surface of the vessel at the location wher e the material receives the highest fluence at the end of the period of evaluation.
The Enclosure
(
9603250353 960322 PDR ADOCK 05000244 PDR
chemistry factor may be determined using credible surveillance data or from the chemistry factor tables in the PTS rule.
The chemistry factors in the tables are dependent upon the best estimate of the copper and nickel content in the material.
The term "best estimate values" is not well defined statistically, but has usually been interpreted as the mean of the measured values.'egulatory Guide (RG) 1.99, Revision 2, contains criteria for determining whether surveillance data are credible.,
RG 1.99 indicates that if there is clear evidence that the copper and nickel content of the surveillance weld differs from that of the vessel weld, the measured values of MTpy$
should be adjusted by multiplying them by the ratio between the chemistry factor for the vessel weld and that for the surveillance weld.
The RG contains a procedure for calculating the vessel weld chemistry factor,from the adjusted or measured values of MT>>$.
In this procedure, the, chemistry factor is calculated by multiplying each adjusted or measured value of MT,7$
by its corresponding fluence factor, summing the products, and dividing by the sum of the squares of the fluence factors.
The resulting chemistry factor will give the relationship of lBTp7$ to fluence that fits the plant surveillance data in such a way as to minimize the sum of'the squares of the errors.
The margin term is intended to account for variability in initial reference temperature and the adjustment in reference temperature caused by irradiation.
The value of the margin term is dependent upon whether the initial reference temperature was a measured or generic value and whether the adjustment in reference temperature was determined from credible surveillance data or from the chemistry factor tables in. the PTS rule.
2.0 EVA UATION The Ginna reactor vessel beltline consists of intermediate shell forging heat number
- 125S255, lower shell forging heat number
- 125P666, and intermediate to lower shell circumferential weld seam SA-847.
The material with the greatest amount of embrittlement is the circumferential weld between the intermediate and lower shell forgings.
The circumferential weld was fabricated by Babcock
& 'Wilcox (B&W) using the submerged arc process with I/8-inch mangamese-molybdenum (Hn-Ho-Ni) weld filler wire (heat number 61782) and Linde 80 flux (lot number 8350).
B&W fabricated the Ginna surveillance weld using the submerged arc process with I/8-inch Hn-Ho-Ni weld filler wire (heat number 61782) and Linde 80 flux (lot number 8436).
Since Linde 80 flux is a neutral flux (neutral fluxes do not add copper or nickel to the weld), the use of a different flux lot between the vessel weld and the surveillance weld is not considered significant.
The surveillance weld is considered to be representative of the vessel circumferential weld between the intermediate and lower shell forgings although they were not necessarily fabricated from the same coil of weld wire.
The filler wire for both the surveillance weld and the beltline weld contained a copper coating.
- 2. 1 Initial Reference Temperature The licensee used an initial reference temperature of -4.8 'F for the limiting circumferential weld.
This is the generic value for welds fabricated using Linde 80 flux. 'he generic data were determined from tests on 34 welds from weld metal procedure qualification test results, nozzle belt weld dropouts, and reactor vessel material surveillance welds that were fabricated by B&W using the submerged arc process with Linde 80 flux.
The data had a minimum value of -50 'F, a maximum value of 33 'F and a standard deviation of 19.7 'F.
The data are documented in B&W Owners Group (BWOG) report BAW-1803, Rev.
1.
The licensee submitted test data from two welds that were fabricated using heat number 61782 weld wire.
One weld had an initial reference temperature of
-1 'F and the other weld had an initial reference temperature of -38 'F.
Since the measured values from the welds fabricated using heat number 61782 weld wire'fall w'ithin the range of the generic data, the use of -4.8 'F as the initial reference temperature of the limiting circumferential weld is acceptable.
2.2 Best-Estimate Chemical'omposition of the Limiting Weld The licensee's best estimate values of the copper and nickel content in the limiting weld are 0.25X and 0.54', respectively.
Linear interpolation of the
, chemistry factors in Table l,of the PTS rule indicates that the chemistry factor is 167.6 'F for welds with such copper and nickel content.
The best estimate values of copper and nickel content are the values recommended in BWOG report BAW-1500 for welds fabricated using heat number 61782 weld wire.
The report contains 12 data points from one weld dropout and 27 from another weld dropout that were fabricated using heat number 61782 weld wire.
These data were taken across the welds.
In addition, the licensee submitted chemical composition data from two Ginna surveillance Char py specimens and a
single data point from a B&W weld qualification test that were fabricated using heat number 61782 weld wire.
The mean'alues of copper and nickel content from all of these weld samples are 0.25X and 0.56X, respectively."
Linear interpolation of the chemistry factors in Table 1 of the PTS rule
, indicates that the chemistry factor is 170.4 'F for welds with these amounts
The NRC staff is concerned that a simple average of the data does not always represent a best estimate of the copper.
The staff's review of other reactor vessels with copper-coated filler wire indicates that there could be large coil-to-coil variability in the amount of copper because the copper coating, varies on the filler wire.
The licensees for Calvert Cliffs and Palisades plants accounted for this variability by determining the best estimate of the copper content from a weighted average of the test results.
In a weighted
- average, the average copper content from samples that represent more'han one coil is weighted according to how many, coils were used to fabricate the weld.
The staff discussed this concern with the licensee for Ginna.
The licensee indicated that it could not accurately determine how many coils were used to fabricate the four welds that represent the data base.
On the basis of the number and location, of the measurements on the two weld dropouts, it was determined that the weld dropouts contain data from many more weld coils than the data from the surveillance 'weld and weld qualification.
Therefore, to account for the coil-to-coil variability in the copper content, the average copper content from the weld dropouts should be given greater weight than the average copper content from the surveillance weld and weld qualification.
The weld dropout with the greater number of measurements has an average copper content of 0.27X and the dropout with the fewer measurements has an average copper content of 0.204X.
The thicker weld dropout has the greater number of measurements; it should have been fabricated with more coils and should provide a best estimate-of copper content for the Ginna beltline weld that conservatively accounts for the coil-to-coil variability in copper content.
Linear interpolation of the chemistry factors in Table 1 of the PTS rule indicates that the chemistry factor is 178.2 'F for welds with 0.27X copper content and 0.56X nickel content.
2.3 Best-Estimate Chemical Composition of the Ginna Surveillance Meld The licensee's best estimates of the copper and nickel content in the Ginna surveillance weld are 0.214X and 0.505X, respectively.
Linear interpolation of the chemistry factors in Table 1 of the PTS rule indicates that the chemistry factor =is 150.9 'F for welds with these percentages of copper and nickel.
B&W fabricated the licensee's surveillance weld using heat number 61782 weld wire and lot number 8436 Linde 80 flux.
The licensee's best estimate of copper and nickel content are the mean values of all the data from welds fabricated using heat number 61782 weld wire and lot number 8436 Linde 80 flux.
These data include the 12 data points from one of the weld dropouts, the 2 data points from the Ginna surveillance Charpy specimens, and the 1 data point from the BSW weld qualification test.
However, the mean values of all the data from welds fabricated using heat number 61782 weld wire and lot number 8436 Linde 80 flux should not be used to determine the best estimate of the copper and nickel content in the surveillance weld because the surveillance weld and the other welds were probably not fabricated using the same coil of weld wire.
The staff believes that the best estimate of the copper and nickel content in a surveillance weld is the mean value of the measurements of these elements from the surveillance weld itself.
Therefore, in this case, the best estimate of the copper and nickel content in the surveill,ance weld is the mean values from the two data points from the Ginna surveillance Charpy specimens.
- The, mean values of copper and nickel content from the Ginna surveillance Charpy specimens are 0.225X and 0.53X, respectively.
Linear interpolation of the, chemistry factors in Table 1 of the PTS rule indicates that the chemistry factor is 157.6& 'F for welds containing these amounts of copper and nickel.
2.4 Evaluation of Surveillance Data The licensee determined the chemistry factor for its vessel weld using:
(1) the Ginna surveillance weld data, (2) the ratio procedure that is recommended in RG 1.99, Rev.
2 when the chemistry of the surveillance weld differs from the vessel
- weld, and (3) the calculational procedures that are recommended in F
J'
RG 1.99, Rev. 2.
The licensee estimated that the ratio of the chemistry factor of the vessel weld to the chemistry factor of the surveillance weld was
- 1. 11.
The licensee calculated the chemistry factor 164.8, 'F.
The staff determined the chemistry factor for the vessel weld using its best-estimate chemistry for the vessel weld (0.27X copper and '0.56X nickel) and the surveillance weld that was discussed above.
The ratio of the chemistry factor of the vessel weld to the chemistry factor of the surveillance weld was
- 1. 13.
The staff calculated the chemistry factor as 167.8 'F.
The licensee determined that its surveillance data met the credibility criterion in RG 1.99, Rev. 2.
Criterion 3 in the RG indicates that the scatter of the measured MT>>, about the best-fit line as described in Regulatory Position 2.1 generally should be less than 28 'F for welds and 17
'F for base metal.
In evaluating this criterion, the licensee did not evaluate the data relative to the best-fit line described in Regulatory Position 2. 1.
That line requires that the linear interpolation of the data should result in a line with 0 MT at 0 neutron fluence.
The licensee's best-fit line was a linear interpoPation of the data, but the MT>>, at 0
neutron fluence was permitted to have positive or negative values.
The licensee's best-fit line does not meet the recommendations of Regulatory Position 2. 1.
However, the staff independently 'evaluated the scatter of the-measured AT>>
about the best-fit line as described in Regulatory Position 2.1 of RG 1.90, Rev. 2, and determined that the surveillance data satisfied criterion 3 in RG 1.99, Rev.
2.
- Hence, the surveillance data are credible and should be used to determine the chemistry factor for the vessel weld.
2.5 Hargin Value The licensee calculated the margin value in accordance with the methodology recommended in RG 1.99, Rev.
2, using the standard deviation for the initial reference temperature from the generic Linde 80 data and reducing by half the standard deviation for the increase in reference.
(RG 1.99, Rev.
2 recommends that the standard deviation for the increase in reference temperature be reduced in half, if the surveillance data are credible).
The licensee calculated a margin value of 48.3 'F.
This value is acceptable since it was calculated in accordance with the methodology in RG 1.99, Rev.
2.
2.6 Projected RT>>~ Value at Expiration of License The RT>>, value calculated by the licensee at the expiration of its license EOL) is 265 'F.
The RT>>s value calculated by the staff for Ginna at the expiration of its license is 268 'F.
The staff calculated its value using:
(1) a generic value of initial reference temperature, (2) staff best estimates of copper and nickel content for the vessel and surveillance welds, (3) a chemistry factor calculated from Ginna surveillance weld data and adjusted to account for the difference between the best-estimate chemistry of the vessel and surveillance welds, (4) an end of life neutron fluence of 3.68E19 n/cm, and (5) a margin value of 48.3'F.
The difference between the staff and licensee RT>>
values results from the difference in chemistry factors (167.8
'F calculated by the staff and 164.8 'F calculated by the licensee).
The NRC staff assessed the impact of increased variability in chemistry on the RT>>$ value of the vessels of pressurized-water reactors.
Ginna was identified as the plant most affected by the staff's generic assessment.
Using the methodology described in the assessment, Ginna was projected to reach the PTS screening criterion in about 7 effective full power years from January 1,
1995.
In this generic assessment, the best estimate of copper content for vessel welds was the mean value of copper content in all BWOG nozzle dropouts and in surveillance welds that were fabricated by B&W.
The mean copper content in these B&W fabricated welds is 0 '87X ~
The best estimate of copper content for welds fabricated using heat number 61782 weld wire was determined from test results from four sample welds (two weld dropouts, a surveillance weld, and a weld qualification) ~
The staff is concerned that the use of the mean copper content from this small amount of data could under-estimate the amount of embrittlement.
- Hence, the staff calculated the RT ~
value for Ginna using the mean copper content from all B&W fabricated
- weeds, which is 0.287X.
The projected value of RT,, is 276 'F using a copper content of 0.287X and all of the other parameters that were discussed in the preceding paragraphs'ence, using Ginna surveillance data and the copper content from the staff's generic assessment indicates that the
'inna reactor vessel would be below the PTS screening criteria at the expiration of its license.
The reason for the difference between this evaluation and the generic evaluation referenced above is that the generic assessment did not consider the evaluation of plant-specific surveillance data in accordance with 10 CFR 50.61.
1)
The licensee's method for determining the credibility of the Ginna surveillance data did not conform to the guidance in RG 1.99, Rev.
2.
However, staff evaluation of the data indicates that the Ginna surveillance data comply with the credibility criteria of RG 1.99, Rev.
2.
2)
Since the Ginna surveillance data comply with the credibility criteria of RG 1.99, Rev. '2, the surveillance data should be used to determine the chemistry factor for the limiting Ginna vessel weld 3)
The licensee's and the staff's calculated values of RT 7$ at the expiration of the Ginna license are within 3 'F (265 '( and 268 'F) and are well below the 300 'F screening criterion specified in 10 CFR 50.61 for circumferential welds.
Since this conclusion is dependent upon the available chemistry data and surveillance data, it is subject to change when new data become available.
Principal Contributor:
Barry Elliot Date:
~pc/
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March 22, 1996 Dr. Robert C. Mecr Vice President, Nuc r Operations
~Rochester Gas
& Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
R.E.
GINNA NUCLEAR POWER PLANT PRESSURIZED THERMAL SHOCK EVALUATION (TAC NO. M93827)
Dear Dr. Mecredy:
By letter dated October ll, 1995, you submitted an updated pressurized thermal shock (PTS) evaluation for the R.E.
Ginna Nuclear Power Plant (Ginna).
The NRC staff has completed its review of your submittal.
In the enclosed safety evaluation (SE), the staff concludes that the Ginna reactor vessel is projected to be below the PTS screening criteria at the expiration of its license.
Since this conclusion is dependent upon the'vailable chemistry and surveillance data, it is subject to change when new data become available.
Additionally, the Ginna reactor coolant system pressure and temperature limits report, as referenced in Ginna Technical Specification 5.6.6, should be updated to include the material properties discussed in the enclosed SE.
The update should cover the use of the ratioing procedure in Regulatory Guide 1.99, Revision 2.
Also with regard to,'your,,specific methodology for determining low-temperature overpressure protection setpoints and its application of American Society of Mechanical Engjneers Boiler and Pressure Vessel Code (ASME Code)
Case N-514; you should request an exemption from 10 CFR 50.55a to use the ASME Code'ddenda that incorporates Code Case N-514,Section XI, Division', nLow'Temperature Overpressure Protection,"
approved February 12, '1992.
i ~
ll Sincerely,
\\
I Original signed by:
,l E
V Allen R. Johnson, Pro'ject Manager Project Directorate I-1 Division of Reactor Projects
, I/II,
Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosure:
As stated cc:
See next page DISTRIBUTION:
See next page DOCUMENT NAME:
G'iGINNAi93827. LTR
- See Previous Concurrence
- See previous concurrence To receive a copy of this document, indicate in the box:
"C" - Copy without attachment enclosure "E" - Co wit
.attachment enclosure "N" = No co OFFICE LA:PDI-1 E
PII: PD I-1 EHCB*
OTSB*
D:
NAME SLittle AJohnson/rs JStrosnider CGrimes Ssh DATE 03/
96 03/
/96 03/21/96 Officia Record Copy 03/22/96 03/
96
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Dr. Robert C.
Hec Vice President; Nu ar Operations Rochester Gas
& Electric Corporation 89 East Avenue Rochester, NY 14649 V
SUBJECT:
R.E.
GINNA NUCLEAR POWER PLANT PRESSURIZED THERMAL SHOCK EVALUATION (TAC NO. H93827)
Dear Dr. Hecredy:
By letter dated October 11,
- 1995, you submitted updated pressurized thermal shock (PTS) evaluation for the R.E.
Gi a Nuclear Power Plant (Ginna).
The NRC staff has completed its r Tew of your submittal.
In the enclosed safety evaluation (SE), the aff concludes that the Ginna reactor vessel is projected to be below he PTS screening criteria at the expiration of its license.
Since thi conclusion is dependent upon the available chemistry and surveillanc data, it is subject to change when new data become available.
Additionally, the Ginna react coolant system pressure and temperature
'imits
- report, as reference in Ginna Technical Specification 5.6.6, should be updated to inclu e the material properties discussed in the enclosed SE.
The update hould cover the use of the rationing procedure in Regulatory Guide 1.9
, Revision 2.
Also with regard to your specific methodology for deter ning low-temperature overpressure protection setpoints and its ap ication of American Society of Mechanical Engineers Boiler and Pressur Vessel Code (ASME Code)
Case N-514, you should request an exempt'on from 10 CFR 50.55a to use the ASHE Code Addendum that incorporat Code Case N-514,Section XI, Division 1, "Low Temperature Ov pressure Protection,"
approved February 12, 1992.
Sincerely, ocket No. 50-244 Allen R. Johnson, Project Manager Project Directorate I-l Division of Reactor Projects
-. I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc:
See next page DISTRIBUTION:
See next page DOCUMENT NAME:
G: iGINNAi93827. LTR To receive attachment enclosure "E" = Co with attachment e
sure "N" = No co a copy of this document, indicate in the box:
"C" = Copy without P
EMCB OFFICE LA:PDI 1
E PH:PDI-1 NANE SLittie AJohnson/rsi JStrosnide OTSB CGrimes D:PDI-1 SShenkmen DATE 03/eL /96 03/
96 03/
i /96 Official Reco d Copy 03/
/96 03/
/96
N I ~
Dr. Robert C.
Mec Vice President, Nu ar Operations
. Rochester Gas L Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
R.E.
GINNA NUCLEAR POWER PLANT - PRESSURIZED THERMA SHOCK EVALUATION (TAC NO. M93827)
Dear Dr. Mecredy:
By letter dated October 11,
- 1995, you submitted an upd ed pressurized thermal shock (PTS) evaluation for the R.E.
Ginna Nu ear Power Plant (Ginna).
The NRC staff has completed its review of your submittal.
In the enclosed safety evaluation (SE), the staff co eludes that the Ginna reactor vessel is projected to be below the PTS creening criteria at the expiration of its license.
Since this conclu on is dependent upon the available chemistry and surveillance daty, 's subject to change when new data become available.
Additionally, the Ginna reactor coolan system pressure and temperature limits report, as referenced in Ginn echnical Specification 5.6.6, should be updated to include the ma rial properties discussed in the enclosed SE.
The update should c
er the use of the ratioing procedure in Regulatory Guide 1.99, Revisi n 2.
Also with regard to your specific methodology for determining lo -temperature overpressure protection setpoints and its applicatio of American Society of Mechanical Engineers Boiler and Pressure Vessel ode (ASME Code)
Case N-514, you should request an exemption fro 0
CFR 50.55a to use the ASME Code Addendum that incorporates Code g se N-514,Section XI, Division 1, "Low Temperature Overpress Pe Protection,"
approved February 12, 1992.
Sincerely, Docke No. 50-244 En osure:
As stated Allen R. Johnson, Project Manager Project Directorate I-l Division of Reactor Projects I/II Office of Nuclear Reactor Regulation c
See next page DISTRIBUTION:
See next page OFFICE LA:PDI-E PM:PDI-1 EMCB*
0TSD ~
NAME SLittl AJohnson/rs l JStrosnider CGrimes DATE 03/
/96 03/
/96 03/21/96 OffTcial Record Copy 03/
96 DOCUHE T NAME:
G:iGINNAi93827.LTR
- See previous concurrence To receive a copy of this document, indicate in the box:
"C" Copy without attachment enclosure "E" = Co wit attachment enclosure "N" = No co 0:PDI-1 SShankman 03/
/96
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