ML18057B417
| ML18057B417 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/16/1991 |
| From: | Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9112200175 | |
| Download: ML18057B417 (41) | |
Text
consumers Power ~ B Slade General Manager POWERiNii MICHlliAN'S PRDliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 December 16, 1991 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
PROJECTED VALUES OF RTPrs FOR REACTOR BELTLINE MATERIALS - 10CFR50.61 PRESSURIZED THERMAL SHOCK This letter* submits information required by 10CFR50.61 and responds to the September 3, 1991 NRC request for additional information concerning calculation of the fluence affecting the Palisades reactor vessel.
10CFR50.61 requires each licensee having a pressurized water reactor to submit, by December 16, 1991, projected values of RTPT~ for reactor vessel beltline materials if the value of RTPTS for any material in the beltline is projected to exceed the PTS screening criterion before the expiration date of the operating license. The Palisades operating license expires in March 2007; and, as shown in Enclosure 1 "Projected RT~ 8 Values for the Palisades Reactor Vessel", the circumferential weld is projected to exceed the screening criterion in March 2007 also and the axial welds are projected to exceed the screening criterion in November 2005.
Enclosure 1 provides the bases for the projected RT rs values, assumptions regarding core loading patterns, the copper and nickel contents and the fluence values used in the calculation for each beltline material. The assessment of future RTPrs values is based on chemistry data classified per the requirements of revised 10CFR50.61 as "best estimate" and thus eliminates some data points included in our previous projections which were based on "Generic" data. The enclosed projection is based on "best estimate" data and results in extending the date (compared to previous projections) when the axial welds will exceed the screening criterion and shortening the time before the circumferential welds will exceed the screening criterion. This "best estimate" method of classifying the applicable available data has been di~cussed with our NSSS (Combustion Engineering) and a leading independent consultant who concur with our approach and assumptions.
A CMS ENERGY COMPANY
I, In accordance with the requirements of 10CFR50.6l{b){4), CPCo will submit, before March 16, 1992, an analysis and schedule for implementation of a flux reduction program such as is reasonably practical to avoid exceeding the 10CFRS0.61 PTS screening criterion.
An NRC letter dated September 3, 1991 requested 12 items of additional .
information regarding Palisades use of the DOT 4.3 transport code in our May 17, 1990 and April 17, 1991 submittals which projected the date when the 10CFRS0.61 screening criterion would be exceeded. As stated in our November 11, 1991 submittal, we have withdrawn our May 17, 1990 and April 17, 1991 submittals which relied on in-house DOT 4.3 methodology and are basing our fluence estimates on Westinghouse DOT methodology. Enclosure 2 presents a description of the Westinghouse DOT methodology and addresses items 1-9 of the requested additional information as they pertain to the Westinghouse methodology. The information requested in items 10-12 apply to CPCo's use of the Palisades in-house DOT 4.3 methodology which is no longer used as a basis for our fluence estimates. Therefore, the additional information requested in items 10-12 is no longer relevant and is not addressed.
4/P~
Gerald B Slade General Manager CC Administrator, Region III, USNRC NRC Resident Inspector - Palisades Attachment
ENCLOSURE 1 Consumers Power Company Pa 1i sades Pl ant Docket 50-255 r:ERO~fLcllfDERi PRlA~~~uRE;[s~~L DECEMBER 1991 December 1991 18 Pages
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INTRODUCTION This enclosure provides information in accordance with the requirements of 10CPR50.61.
Specifically, the information forms the bases for the projections of when the Palisades reactor vessel beltline materials will exceed the screening criterion including assumptions regarding core loading patterns, chemistry of the baseplate and vessel welds, and the fluence values used in the calculations.
The major assumption associated with the bases of Palisades RTPTs calculations is the continued use of low l~ge cores. Implementation of a low leakage core concept was initiated with Cycle 8, improved in Cycle 9 and all future cycles will have leakage no higher than that associated with Cycle 9. Pluence is assumed to accumulate at the rate presently occurring in Cycle 9 at a continuous operating capacity factor of 75 %. Although different values for the capacity factor could be substantiated, the resulting variance in the adjusted reference temperature is minimal ~t the accumulative levels of fluence applicable to Palisades.
The copper and nickel chemistry of the reactor vessel baseplate and welds is based on the results of chemical analysis of Palisades' actual reactor vessel material for baseplate material and on the best estimate mean of measured values of welds made with the same weld wire heat number used in the Palisades axial and circumferential welds. Specific details supporting the
.determination of Palisades' vessel chemistry is provided later in this enclosure.
The data base utilized for fluence calculations consists of power distribution information developed from core monitoring system data, fuel vendor supplied core models, plant-specific dimensional data and material composition. The modeling of the vessel and fluence analysis was performed by Westinghouse using the DOT ill W discrete ordinates transport code and the SAILOR code cross-section library. The Westinghouse calculational methodology has been benchmarked by calculating the ORNL PCA benchmark and the Mol Belgium VENUS benchmark.
The initial reference temperature used for the axial and circumferental welds is -56°P which is specified for use with LINDE 1092 and LINDE 124 flux while a measured value of -5°P was used for the base metal. The margin term used for the axial and circumferential welds is 66 °P and 34 °P is used for base metal. These terms correspond to the initial reference temperature determination associated with the welds and base metal.
Application of best-estimate chemistry as defined in 10CPR50.61 and Westinghouse fluence calculations have been used to determine the Palisades reactor vessel's projected capability to meet the PTS screening criteria. The results indicate that the baseplate material will meet the screening criterion but the axial weld will exceed the screening criterion in November of 2005.
The circumferential weld reaches its screening criterion at the Plant license expiration date of March 2007.
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Background
On June 14, 1985, Consumers Power Company (CPCo) submitted to the Nuclear Regulatory Commission (NRC) a proposed Technical Specifications Change Request for new Palisades Plant reactor pressure vessel pressure and temperature limits. Included in that submittal was a description of the Palisades Plant reactor vessel materials. In Attachment III of that letter, CPCo stated "It would appear that the Palisades RACO 3 vessel welds are best represented by other RACO 3 welds made for other vessels and for other surveillance programs at the time of manufacture of the Palisades reactor vessel." The summary then went on to conclude that the mean values for RACO 3 welds of.0.19 wt% Cu and 1.10 wt% Ni extracted from EPRI Report EPRI NP-3573-SR1 were representative of Palisades' axial welds. The chemistry for the Mil B-4 modified heat 27204 circumferential weld was concluded to contain .21 wt% Cu and .98 wt% Ni. This information was obtained from a Pacific Gas and Electric Company record search at the CE Chattanooga facility for the Diablo Canyon Unit 1 reactor vessel/surveillance program fabrication data. The Palisades reactor vessel plate material contains .25 wt% copper, and .54 wt% nickel. This information is based on chemical analysis of specimens which represents Palisades actual reactor vessel material.
On August 21, 1985, the NRC issued the revised pressure and temperature limits for the Palisades Nuclear Plant (PNP). The Safety Evaluation referenced CPCo's June 14, 1985 submittal and stated "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 PNP beltline material. Since the amount of copper and nickel should be consistent within a heat of weld wire and the 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 from the same heats of wire as the PNP beltline weld should provide reliable estimates for the amounts of copper and nickel in the PNP beltline welds."
This statement appears to be incorrect when applied to the data submitted in 1985 since that data contains heats other than those used in the Palisades' reactor vessel. However, this statement is consistent with the 10CFR50.61 classification of "best estimate" data and correctly applies to the data utilized in this assessment.
On January 23, 1986, CPCo responded to the original PTS rule issued July 23, 1985 as 10CFR50.61. This response includes the statement "Consumers Power Company letter dated June 14, 1985 clearly established our intent to employ generic chemistry for both circumferential and axial welds metal". The January 23 letter then lists the. same chemistries as described above.
1 Marston, T.U. et al, "Robinson 2 Reactor Vessel: Pressurized Thermal Shock Analysis for a Small Break LOCA", EPRI NP-3573-SR, August, 1984.
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In their May 6, 1986 response to the January 23, 1986 submittal, the NRC requested that CPCo define the phrase "generic chemistry". CPCo's response on August 7, 1986 states "The term
'generic chemistry' as employed in Consumers Power Company's January 23, 1986 submittal indicated our intention to use the average of the copper and nickel values as determined from the search of fabrication records of other reactor vessels constructed by Combustion Engineering during the same period that the Palisades reactor vessel was being fabricated."
Palisades' use of available chemistry data and classification of the data as generic was established prior to the issue of Regulatory Guide 1.99 Rev 2 which provided specific guidance to determine the content of copper and nickel in vessel welds based on the source of the data.
While Palisades~ chemistry data did not exactly meet the specific classification guidance of Regulatory Guide 1. 99 Rev 2 these values were considered to be based on an acceptable alternative.
Therefore, CPCo has used the above chemistry as input to all subsequent PTS submittals concerning Palisades accumulated fluence and efforts at flux reduction through fuel management.
Previously submitted data for the axial and circumferential welds and plate material are provided in (attached) Tables 1, 2, 3 and 4 respectively. This data was used to calculate the average chemistry values labeled "previously submitted" in Table 9.
Current Situation While reviewing the new revision to 10CFR50.6l(b) to assess the need to resubmit our previous PTS assessment, CPCo determined that our presently approved vessel chemistry does not precisely fit within tqe categories described in 10CFR50.61(b)(2)(iv). Previously submitted axial weld chemistry used a combination of data taken from weld deposits made from the same RACO 3 weld wire heat as was the Palisades reactor vessel, as well as data taken from other heats of RACO 3 welds and RACO 3 welds with no identified specific heat. As earlier explained, CPCo termed that submitted chemistry to be generic. The 10CFR50.61 definition of generic is described in a footnote in paragraph (b)(2)(iv) stating "Data from reactor vessels fabricated to the same material specification in the same shop as the vessel in question and in the same time period is an example of 'generic data.'" A review of Palisades submitted axial weld data (Table
- 1) indicates that greater than half of the data came from welds made from the same weld wire heat as the Palisades vessel. This heat specific data is considered to be significantly more representative of the Palisades welds as explained below.
Since a significant length of time has passed since CPCo's original determination of weld chemistry, a data search was conducted to ensure a best effort was made to consider all available chemistry data. This search involved several organizations and included a search of weld chemistry records maintained by NSSS vendors Westinghouse and Combustion Engineering.
Also included was a search of NRC docketed PTS submittals conducted by Bechtel and a search of weld chemistry records maintained by ATI Consulting (William Server). This search confirmed the applicability of the existing heat specific weld data and found additional heat specific data applicable to the Palisades reactor vessel welds.
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Based on engineering judgement, confidence in weld chemistry data is reflected in the following logic:
- 1. The best chemistry is actual samples taken from the actual welds in the vessel.
- 2. The next best is the chemistry of that unit's surveillance material if the surveillance weld used the same heat of wire as the weld in question and was made with a similar weld procedure.
- 3. The next best data is that from other welds made with the same heats of wire, with the same type of flux, in the same shop, with similar procedures, and in the same time frame as the actual vessel welds. Original records should be researched to assure that the chemistry is for actual welds and is not for weld wire.
- 4. The next best alternative is to use welds made with wire purchased to the same specification from the same supplier in the same size, welded with the same weld procedure and parameters in the same shop and in the same area as the actual vessel welds.
- 5. If none of the above are possible then specification maximums or default values which
- bound expected chemistry results can be used.
Applying this logic to the chemistry classifications specified in the revision to 10CFR 50.6l(b)
(2)(iv):
- Items 1, 2 and 3 are considered "best estimate" values since they represent actual plant specific weld material or welds made from the same weld wire heat number that matches the critical vessel weld.
- Item 4 represents "generic" estimated data as defined in 10CFR50.6l(b)(2)(iv) per footnote 1.
Based on this comparision CPCo concludes that this logic is in full agreement with the requirements of 10CFR50.61.
Based on application of this logic to previously reported vessel chemistry it is CPCo's position that the ht!1lt specific data can be used to meet the "best-estimate" criteria as defined in paragraph (b)(2)(iv). The significant difference between Palisades' previously reported chemistry and the 10CFR50.61 defined best-estimate chemistry is that the Cu values previously reported for the axial welds contained measurements from unknown or different heat numbers of RACO 3 welds; 4
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This previous weld data, with different heat numbers is not considered to best represent the welds used in the fabrication of the Palisades' vessel, and results in a more limiting Cu mean value. All measurements from different heat numbers or unspecified heats are deleted from the calculation of the Palisades reactor vessel axial welds best estimate Cu content.
The previously submitted chemistry for the vessel circumferential weld (Table 2) is presently b_ased on only two heat specific weld samples. The data search mentioned above determined that additional information was available for the Palisades specific weld heat 27204 for the circumferential weld. This additional data is included in the best estimate circumferential weld data. The previously submitted plate chemistry data represents Palisades' reactor vessel and remains unchanged.
DETERMINATION OF REACTOR VESSEL BELTLINE MATERIAL CHEMISTRY General Paragraph (b)(2)(iv) of 10CFR50.61 identifies the best-estimate chemistry as the mean of measured values for weld samples made with the weld wire heat number that matches the critical vessel weld. Palisades beltline welds were made using the weld wire heat numbers as specified in Table 5.
The information presented in Table 5 for beltline weld consumables reiterates the information provided in previous reactor vessel submittals. Figure 1 shows the location of each of these welds. The Palisades axial welds were made with a Raco 3 weld wire plus nickel addition wire heat number N7753A. The axial welds above the beltline circumferential weld were fabricated with the W5214 Raco 3 weld wire, while the axial welds below the beltline circumferential weld were fabricated using both weld wire heat numbers W5214 and 34B009. The beltline circumferential weld was fabricated with a Mil B-4 modified weld wire heat number 27204 with nickel included.
- * *ideally; the* reactor vessel beltline material chemistry would be determined from chemistry samples taken from the Palisades reactor vessel weld and plate material, or from surveillance specimens made from the same material and heats of wire as that used to fabricate the vessel.
As is the case with many vessels built during the same time frame, Consumers Power Company does not have chemistry measurements for the Palisades v.essel specific beltline welds, nor does it have a surveillance specimen made with the same material and heat of wire. Consumers Power Company does have copper and nickel measurements for the actual vessel beltline plate material.
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Axial welds In determining the best-estimate chemistry for the Palisades reactor vessel axial welds, CPCo has derived the copper and nickel content by using chemistry samples from weldments and surveillance specimens fabricated with weld wire heat numbers W5214 and 34B009 with nickel addition wire heat number N7753A. As previously explained, the additional chemistry data from Raco 3 weldments made with wire heats different than these or of unknown origin are no longer considered as applicable to Palisades' best-estimate chemistry. A total of seven copper measurements and one nickel measurement previously reported are no longer considered applicable to Palisades. One additional copper measurement from the Indian Point 2 surveillance program is included in the new axial weld chemistry values presented in Table 6.
The copper content of the upper axial welds is determined by calculating the mean of the copper measurements taken from weld deposits fabricated using weld wire heat number W5214. In determining a copper concentration for the lower axial welds, it was concluded that the mean copper concentrations for weld wire heat numbers W5214 and 34B009 should both be calculated, with the highest value assumed to be applicable. A mean copper concentration of 0.18 weight percent was assumed to represent all six axial welds. This represents a .01 Wt% decrease from previous submittals.
The nickel content of the axial welds is determined by calculating the mean of the nickel measurements taken from weld samples fabricated using nickel addition weld wire heat number N7753A with Raco 3 weld wire. One data point has been deleted from the previously submitted nickel data because the heat number was unknown. However, this is a conservative approach since the data point had the second lowest nickel content. This data point was included in previous submittals .however a review of mean value calculations indicate it was not included in previous calculations, therefore, the weight percent of nickel remains the same. A summary of the axial weld data and chemistry is presented in Table 6.
Circumferential Weld The best-estimate copper and nickel content of the circumferential weld was determined by calculating the mean of copper and nickel measurements taken from weld deposits fabricated with Mil B-4 modified weld wire heat number 27204. As mentioned earlier, a search of available data has determined that additional heat number 27204 data is available. As a result, the previous data (Table 3) has been revised. The source of this data is CE metallurgical reports.
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Also, the chemistry of the Diablo Canyon surveillance specimen has been adjusted by three additional chemistry measurements performed on irradiated samples from the same weld. 2 A summary of the circumferential weld data and chemistry is presented in Table 7. This data represents an increase of 0.01 wt% copper and 0.03 wt% nickel.
Baseplate material Table 8 identifies the chemical composition of the Palisades plate materials and is identical with information provided in the past. While Table 8 shows the calculated mean values for Cu and Ni, *cPCo will continue to conservatively use 0.25 w/o Cu and 0.54 w/o Ni of the limiting plate material.
Conclusion The revised Palisades reactor vessel beltline material chemistry is summarized in Table 9.
Sufficient data exists to accurately estimate the nickel and copper content of the Palisades reactor
- -***vessel axial and circumferential welds and the base plate material. Weld chemistry can be based on data which represents the same weld wire heat used in the Palisades reactor vessel welds.
Base plate chemistry is *based on actual Palisades reactor vessel material. This weld chemistry data represents the highest tier confidence level, was treated in a conservative manner, and meets the "best estimate" criteria of 10CFR50.61 (b)(2)(iv) .
PALISADES PLANT REACTOR VESSEL FLUENCE Westinghouse performed an assessment of Palisades' reactor vessel fluence from beginning of life through Cycle 9. The methodology employed the DOT III W two dimensional discrete ordinate transport code, the SAILOR cross section library and Palisades plant specific 1/8th core symmetric geometry configuration.
The complete results of this assessment are provided in Enclosure 2 and a summary of the flux and :fluence data used to perform RTYrs calculations is provided in Tables 11 and 12. These tables include previous calculations performed by Westinghouse for Cycles 1 through 8 anc;l utilized in our submittal dated April 3, 1989 which is our present analysis of record. In general it can be seen that the present calculations are more conservative. This difference is attributed to refinements of the core geometry model.
2 WCAP-11567, "Analysis of Capsule S from the Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program", Yanichko, Anderson, Schmertz, and Albertin, Westinghouse Electric Corporation, December 1987.
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DETERMINATION OF RTPTs FOR THE VESSEL BELTLINE MATERIALS General Table 10 lists the limiting fluence and the RTPTs for each of the Palisades reactor vessel-beltline materials as required by 10CFR50.61. The time intervals addressed are the date of this submittal (December 16, 1991), the date each of the welds reaches the applicable screening criteria, and the date the operating license expires (March 14, 2007). Fluence is assumed to accumulate at the rate presently occurring in Cycle 9 at a continuous operating capacity of 75 %.
Although different values for capacity could be substantiated, the resulting variance in the adjusted reference temperature is minimal at a the accumulated levels of fluence applicable to Palisades.
Conclusions
- As shown in Table 10, it is projected under the current opentting assumptions that the Palisades reactor vessel circumferential weld will reach its screening criteria in March 2007 and the axial welds will reach their screening criteria in November 2005. Properly classifying the chemistry data as best-estimate has extended the date at which the axial weld will exceed the screening criteria by 4 years. On the other hand, the new additional chemistry data has moved up by 10 years the date at which the circumferential weld will exceed the screening criteria. The status of the base plate remains unchanged from previous submittals and does not exceed the screening criteria.
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. TABLE 1 AXIAL WELDS PREVIOUSLY REPORTED COPPER MEASUREMENTS ON RACO 3 WELDMENTS MADE BY COMBUSTION ENGINEERING Plant Weld Heat No. Copper Content (w/o)*
Humb~ltBay Surveillance NA+ .22 (6)
Zorita Surveillance 1248 .22 (1)
Big Rock Point Surveillance NA+ .26 (3)
Tarapur Surveillance NA+ .16 (5)
Conn. Yankee Surveillance 9565/86054B .22 (1)
San Onofre Surveillance NA+ .19 (1)
- Millstone Surveillance W5214 .19 (13)
- Indian Point 2 Surveillance W5214 .20 (4)
- Indian Point 3 Surveillance W5214 .15 (1)
Salem 1 Surveillance 39B196 .16 (1)
- Indian Point 3 Longitudinal Seam W5214 .15 (3)
- Robinson Head Weld 1 34B009 .19 (4)
- Robinson Head Weld 2 W5214 .16 (4)
- w/o Average of the number of measurements shown in parentheses
+ Data not available
- Data used to determine heat specific chemistry see Table 6
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- TABLE 2 AXIAL WELDS PREVIOUSLY REPORTED NICKEL CONTENT OF RACO 3 + Ni 200 I LINDE 1092 WELD:MENTS Plant Weld Ni 200 Heat Number Ni Content (w/o)*
Millstone Surveillance N7753A .98 (13)
Salem 1 Surveillance N7753A 1.26 (1)
Indian Point 3 Surveillance N7753A 1.02 (1)
Indian Point 3 Longitudinal Seam N7753A 1.09 (3)
Indian Point 2 Surveillance N7753A 1.15 (4)
MML Record (CE) Weld Deposit N7753A 1.09 (5)
Robinson Head 2 NA+ .99 (4)
- w/o average of the number of measurements shown in parenthesis.
+ Data not available TABLE 3 CIRCUMFERENTIAL WELDS PREVIOUSLY REPORTED COPPER AND NICKEL CONTENT OF l\1IL B-4 MODIFIED WIRE- HEAT NO. 27204 Chemistry (w/o)
Weld Designation Flux Type Lot No. Deposit Form Cu NI MML (CE) LINDE 1092/3774 Vessel Weld .18 .96 Contract 14166 LINDE 1092/3714 Surveillance .21 .98 Program Diablo Canyon 1
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- TABLE 4 PALISADES REACTOR VESSEL AND SURVEILLANCE PROGRAM MATERIAL - PLATE Material Vessel Drop Weight Chemical ComDosition (w/o)
Identification Location NOTT (°F) Cu Ni D-3803-1 Intermediate -30 .25 .48 Shell D-3803-2 Intermediate -30 .25 .50 Shell D-3803-3 Intermediate -30 .25 .48 Shell D-3804-1 Lower -30 NA+ .45 Shell D-3804-2 Lower -40 NA+ .50 Shell D-3804-3 Lower -30 NA+ .54 Shell D-3803-1 Surveillance -10 .25 .53 Material
+ Data not available TABLES PALISADES REACTOR VESSEL BELTLINE WELD CONSUMABLES Location Filler Heat Flux Batch Nickel Addition 2-112A/C Intermediate RAC03 W5214 LINDE 1092 3617 Ni-200, #N-7753A Shell Axial 3-112A/C Lower Shell RAC03 W5214 LINDE 1092 3692 Ni-200, #N-7753A Axial RAC03 34B009 LINDE 1092 3692 Ni-200, #N-7753A 9-112 Intermediate MIL-B4 27204 LINDE 1092 3714 None to Lower Modified LINDE 124 3687 None Girth
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TABLE 6 PALISADES REACTOR VESSEL AXIAL WELD CHEMISTRY RACO 3 WELD WIRE HEAT NUMBER W5214 Plant Weld Copper Content (w/o)*
Millstone Surveillance 0.19 (13)
Indian Point 2 Surveillance 0.20 (4)
Indian Point 3 Surveillance 0.15 Indian Point 3 Axial Seam 0.15 (3)
MML Record (CE) Weld Deposit 0.20 Robinson Head Weld 2 0.16 (4)
Mean= 0.18 RACO 3 WELD WIRE HEAT NUMBER 34B009 Plant Weld Copper Content (w/o)*
MML Record (CE) Weld Deposit 0.15 Robinson Head Weld 1 0.19 (4)
Indian Point 2 Surveillance 0.20 (6)
Mean= 0.18 NICKEL CONTENT OF RACO 3 + Ni200LINDE1092 WELDMENTS Plant Weld Ni Content (w/o)*
Millstone Surveillance 0.98 (13)
Salem 1 Surveillance 1.26 Indian Point 3 Surveillance 1.02 Indian Point 3 Axial Seam 1.09 (3)
Indian Point 2 Surveillance 1.15 (4)
MML Record (CE) Weld Deposit 1.09 (5)
Mean= 1.10
- Weight percent average of the number of measurements when shown in parenthesis.
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- TABLE 7 PALISADES REACTOR VESSEL CIRCUMFERENTIAL WELD CHEMISTRY COPPER AND NICKEL CONTENT OF MIL B-4 MODIFIED WIRE HEAT NUMBER 27204 Chemistry (w/o)
Plant Weld Cu Ni MML Record (CE) D5235, Production 0.22 (3)
D5236, D5237 MML Record (CE) D5395, Production Weld - Circ Seam 0.23 (2) 1.05 (2)
D5401 MML Record (CE) D5414 Circumferential Seam 0.18 0.96
D8658 Production Weld 0.22 1.00 Diablo Canyon 1 Surveillance 0.20 (4) 1.00 (4)*
Mean= 0.21 1.00
- Additional data points of .196, .192, .203 wt% copper and .99, .99, 1.03 wt% nickel were added and averaged with the previously reported data of .21 wt% copper and .98 wt% nickel (See Table 3).
NOTE: There is one additional production weld, MML Record D5207 with 0.32 wt% Cu and 0.82 wt%
Ni that does not appear to be an accurate chemistry for heat 27204. If included in the above chemistry, Cu would equal 0.22 and Ni would equal 0.97. This results in a chemistry factor of 227°F, 2° less conservative than the above.
- TABLE 8 PALISADES REACTOR VESSEL AND SURVEILLANCE PROGRAM BASEPLATE MATERIAL Material Chemistry (w/o)
Identification Vessel Location Cu Ni D-3803-1 Intermediate shell 0.25 0.48 D-3803-2 Intermediate shell 0.25 0.50 D-3803-3 Intermediate shell 0.25 0.48 D-3804-1 Lower shell 0.45 D-3804-2 Lower shell 0.50 D-3804-3 Lower shell 0.54 D-3803-1 Surveillance material 0.25 0.53 Mean= 0.25 0.50 TABLE 9 PALISADES REACTOR VESSEL LIMITING CHEMISTRIES Chemistry (w/o)
Best Estimate Previously Submitted Vessel Location Cu Ni Cu Ni Axial weld 0.18 1.10 .19 1.10 Circumferential weld 0.21 1.00 .20 .97 Baseplate 0.25 0.54 .25 .54 14
- TABLElO PALISADES REACTOR VESSEL MATERIALS WITH RESPECT TO THERMAL SHOCK CRITERION Material RTSPTs Chemistry Constants *RTPTs Wt% op Screening RTPTs - 0 P/Pluence (E19)
Cu Ni I M Criteria - 0 P 12/91 11/05 EOL Long. .18 1.H> -56 66 270 242/1.18 270/1.87 272/1.94 Girth .21 1.00 -56 66 300 269/1.61 298/2.61 300/2.71 Plate .25 .54 -5 34 270 219/1.61 240/2.61 242/2.71
- TABLE 11
- FLUX (E. > 1.0 MeV) AT THE CLAD-BASE METAL INTERFACE Azimuthal Flux (E + 10 ri./cm2 - sec)
Cycle Location (1) (2) 1-5 0°, Axial Weld 4.59 Combined 30°, Axial Weld 4.70 (3) 16 °, Peale (Base Metal, 6.03 Circumferential Weld) 6-7 0°, Axial Weld 4.87 4.23 Combined 30°, Axial Weld 4.79 4.17 16°, Peale (Base Metal, 6.25 5.62
. Circumferential Weld) 8 0°, Axial Weld 2.10 2.07 30°, Axial Weld 2.28 2.21 16°, Peale (Base Metal, 4.76 4.81 Circumferential Weld) 9 0°, Axial Weld 2.09 30~, Axial Weld 1.99 (4) 16°, Peale (Base Metal, 3.05 Circumferential Weld)
(1) Westinghouse Report, December 1991 By E.P. Lippincott, "Palisades Reactor Vessel Fluence Analysis" (2) CPCo Letter to NRC, Dated April 3, 1989 "Docket 50-255 - License DPR Palisades Plant - Compliance with Pressurized Thermal Shock Rule 10CFRS0.61 and Regulatory Guide 1.99 Revision 2 - Fluence Reduction Status (Tac No. 59970)"
(3) Combined Cycles 1-7 Flux Calculations were Performed (4) Cycle 9 (Present Cycle), Flux Calculations were not Performed
TABLE 12 FLUENCE (E > 1.0 MeV) AT TIIE CLAD-BASE METAL INTERFACE Azimuthal Accumulated Fluence (E + 19 n/cm 2)
Cycle Location (1) (2) 1-5 0°, Axial Weld 0.753 Combined 30°, Axial Weld 0.771 (3) 16°, Peale (Base Metal, 0.989 Circumferential Weld) 6-7 0°, Axial Weld 1.05 0.951 Combined 30°, Axial Weld 1.06 0.938 16°, Peale (Base Metal, 1.37 1.264 Circumferential Weld) 8 0°, Axial Weld 1.12 1.014 30°, Axial Weld 1.14 1.006 16 °, Peale (Base Metal, 1.52 1.411 Circumferential Weld) 9 0°, Axial Weld 1.17 30°, Axial Weld 1.19 (4) 16°, Peale (Base Metal, 1.60 Circumferential Weld)
(1) Westinghouse Report, December 1991 By E.P. Lippincott, "Palisades Reactor Vessel Fluence Analysis" (2) CPCo Letter to NRC, Dated April 3, 1989 "Docket 50-255 - License DPR Palisades Plant - Compliance with Pressurized Thermal Shock Rule 10CFR50.61 and Regulatory Guide 1.99 Revision 2 - Fluence Reduction Status (Tac No. 59970)"
(3) Combined Cycles 1.:.7 Flux Calculations were Performed, Fluence Values are at the End-of-Cycle 7 (4) Cycle 9 (Present Cycle), Flux Calculations were not Performed
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- at:c.lilU FIGURE 1 Palisades Reactor Pressure Vessel. Weld Map
ENCLOSURE 2 Consumers Power Company Palisades Plant Docket 50-255 WESTINGHOUSE REPORT PALISADES REACTOR VESSEL FLUENCE ANALYSIS December 1991 19 Pages
- ... t Palisades Reactor Vessel Fluence Analysis E. P. Lippincott December 1991 Introduction A new analysis has been performed to evaluate the Palisades reactor vessel fast neutron exposure through cycle 9 (the current cycle). The analysis was carried out to determine average_ flux values for cycles 1 to 5 and for cycles 6 to 7. Previous calculations reported for cycle 8 [l] and. for cycle. 9 [2] were utilized to integrate the flux through the end of cycle
- 9. All calculations were performed~using the modified fuel source input that takes into account the assembly burnup and the varying fuel assembly .
spacing unique to the Palisades reactor.
Neutron Transport Analysis Method Fast neutron exposure calculations for the reactor and cavity geometry were carried out for Cycles 1-5 and for Cycles 6-7 using di scre.te ordinates transport techniques. Similar calculations already reported for Cycle 8 [l] and for Cycle 9 [2] were used also. These calculations, provided the energy distribution of neutron flux for use in determining the actual exposure at key locations in the reactor vessel wall. The calculations all use an improved methodology that takes into acco~nt the irregular spacing of the Palisades fuel assemblies and the effect of the fuel burnup for each assembly on the neutron source from that assembly.
Thus these calculations represent an update of calculations previously
- reported for Cycles 1-5 average [3] and for Cycles 1-7 average [4]. The calculation reported in reference 1 is also an update of the Cycle 8 calculation reported in reference 4.
1
'" I A plan view of the calculational model of the reactor geometry at the core midplane elevation is shown in Figures 1 and 2. Due to the radial extent of the geometry modelled and in order to keep a fine spatial mesh definition, the calculation was carried out in two parts as shown in these figures. The inner part of the model encompassed the fuel region and the ex-core region through the water inside the reactor vessel. This part of the model contained 61 azimuthal mesh and 85 radial mesh. The outer part of the model utilized the boundary source calculated in the inner part of the model at the inside of the core barrel to calculate the neutron transport from the barrel through the reactor vessel and cavity and into the biological shield. The outer part of the model contained 83 radial mesh.
Since the reactor exhibits 1/8 core symmetry only a 0 to 45 degree sector is depicted in the figures. In addition to the core, reactor internals, reactor vessel, and the primary biological shield, the model also included explicit representations of a surveillance capsule attached to the vessel wall at 20 degrees, the vessel cladding, and the reflective insulation located external to the vessel. From a neutron transport standpoint, the inclusion of the surveillance capsules and associated support structures in the analytical model is significant for analysis of capsule dosimetry results and evaluation of capsule exposures. To a lesser extent, the capsules impact the reactor vessel exposure at locations close to the capsule. Inclusion of the surveillance capsule also allowed a comparison of the new calculated flux value in the capsule for Cycles 1-5 with the W-290 capsule exposure result.
The transport calculation for each of the two parts of the reactor model depicted in Figures 1 and 2 was carried out in R,0 geometry using the DOT-IIIW two-dimensional discrete ordinates transport theory code [5] and the SAILOR cross-section library [6]. The SAILOR library is a 47 neutron energy group ENDF-8/IV based data set produced specifically for light water reactor applications~ In these analyses, anisotropic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an s8 order of angular quadrature. The calculation was normalized to the axial peak power location and for operation at a thermal power level of 2530 MW.
2
The SAILOR library was chosen because of the extensive testing of this set for application in LWR environments. It is recognized that improved cross section sets have been developed and are currently being tested [7]. In particular, the ENDF/B-VI cross section for iron inelastic scattering has been modified to better fit data that indicates increased transmission of high ener.gy neutrons. This would increase the calculated response of dosimetry located in the cavity, but when cavity dosimetry results are used to extrapolate back to the vessel inner radius, lower projected fluences would be obtained. Therefore, the SAILOR cross section set will give conservative results for the vessel IR when cavity dosimetry results are applied. In addition, calculations of neutron flux (E>l MeV) transmission through the vessel indicates the cross section change to have a relatively small effect (about 2% through a 9 inch thick vessel) on the flux (E>l MeV).
The spatial core power distributions utilized in the calculations were supplied by Consumers Power [8], They were used to determine the neutron source time-averaged over the actual operation period for each calculation. The power distribution was supplied as a pin-by-pin power distribution for each of the outer assemblies and as cycle burnups for each assembly. The neutron source was derived for each fuel pin and for each fuel assembly using burnup dependent values of the fission neutron energy spectrum, neutrons per fission, and energy per fission. The source spectrum was calculated by determining the fraction of fissions occurring in each of the important uranium and plutonium isotopes for the mid-cycle burnup and calculating a resultant average fission spectrum using the ENDF-B/V fission spectrum for each isotope. Axial peaking factors were also entered for each assembly. Since almost all the assemblies peaked near the same axial height, the peak power for all assemblies was used, and the calculation therefore represents the fluence at the maximum axial point as if the entire core was at that peak power. The source from each assembly was spatially located to take into account the varying gaps between fuel assemblies and thus represents the location of source from each pin as accurately as possible.
3
The source was converted from the X-Y pin geometry to the R-9 DOT geometry by distributing the source over a square area equal to the pitch for each pin. This area was then divided into a 10 x 10 array, each source element of which was placed into the DOT mesh containing the center point of the source element. The error in source positioning, both radial and azimuthal, was thus kept to less than +/-0.07 cm for each of these source elements. Averaging the error over a very large number of source elements produces an extremely small resultant bias in source positioning. Thus the error due to the geometry conversion can be assumed to be negligible. The neutron source was input to the DOT calculation for the inner part of the model as a 47 group fixed source for each DOT spatial mesh point in the fuel region.
Details of fuel assembly locations, core geometry, and other reactor parameters were taken from the information supplied by Consumers Power for the Cycle 8 calculation [9]. Values of water temperature were slightly different in Cycle 8 and the normal values were used for the other cycles.
The fine mesh spacing used in the calculations ensured that the neutron flux changes between mesh were sufficiently small that the calculation converged to an accurate solution. In particular, changes in the neutron flux above 1 MeV between mesh in the surveillance capsule and surroundings were held to about 10% and in the reactor vessel to less than about 20%.
Experience with Sn calculations indicates that these mesh-to-mesh variations are well below allowable levels for attainment of reliable results.
Uncertainties in reactor vessel fluence evaluati~n arise from the approximations and data errors that are an integral part of any transport calculation. The reactor model used has a finite number of points that have to represent geometry in a symmetry not always appropriate. In the case of typical two dimensional calculations of LWR geometries, the fuel and shroud region is in an x-y configuration while the other reactor components in the calculation (barrel, water gap, reactor vessel, etc.)
are cylindrically symmetric. Errors due to such modeling difficulties were minimized by use of a sufficient number of points to define the 4
geometry and by paying close attention to preserving the volume of each component in the model. Other uncertainties may be introduced by tne inclusion or exclusion of other structures such as former plates and surveillance capsules that are not present symmetrically. Of importance also are cross section uncertainties that can have large impacts on transport of neutrons through the region from the fuel to the reactor cavity. The existence of these uncertainties, together with the reactor geometry uncertainties, require that the calculations be benchmarked to measured results, both related controlled benchmarks (such as the those measured under the NRC-sponsored LWR Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) [10]), and plant specific measurements at Palisades (discussed below). The Westinghouse methodology has been benchmarked against a number of benchmarks including the PCA at ORNL [11], the VENUS mockup at Mol, Belgium [12], and the H. B. Robinson PWR [13], as well as against a large database of reactor surveillance capsule measurements and reactor cavity dosimetry measurements [14,15].
These benchmark calculations support the accuracy of the methods and cross sections, especially in the interpolation of the results between the capsules and the cavity to get the fluence throughout the vessel.
However, the benchmark results of most significance for establishing the accuracy of the Palisades calculated fluence results are measurements made in the Palisades plant itself. Such measurements serve not only to check the calculational accuracy, but also provide a check on plant specific parameters such as actual geometry, material densities, and neutron source distributions.
Results The calculated flux and fluence (E>l MeV) results are tabulated in Table 1 for four azimuthal angles. These include the two longitudinal weld angles of 0 and 30 degrees, the fluence maximum at about 16 degrees which is applicable for the maximum fluence to the vessel plate and circumferential weld, and a point at 45 degrees. Results are tabulated for each cycle, but average flux values are input for Cycles 1-5 and for Cycles 6-7 since these cycles were not calculated individually.
5
- .... t Data available for validation of the calculated results includes the dosimetry results from the W-290 surveillance capsule [3] and the Cycle 8 reactor cavity measurements [l]. An updated evaluation of the capsule fluence was performed and is described below. Following this description, comparisons of the calculated fluence with measurement is made.
Evaluation of Surveillance Capsule Fluence Measured reaction rates for the Palisades W-290 surveillance capsule are given in the capsule report [3]. These values were assigned uncertainties in accordance with the data given, and experience with similar capsule measurements. In particular, average values were used for the capsule to minimize stochastic variation in the data. This averaging ignores possible axial flux gradients in the capsule, which the data indicates could be as large as 10%. Thus the average value could be biased below the maximum reaction rate value by about 4%. This possible bias is less than the assigned reaction rate-uncertainties but should be considered independently in the uncertainty analysis.
Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code
[16]. The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the sensor set location (surveillance capsule center point) as input and proceeded to adjust the a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The exposure parameters along with their associated uncertainties were then obtained from the adjusted spectrum.
In the FERRET evaluation, a log-normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values fare linearly related to the flux~ by some response matrix A:
(s,a) (s) (a) fi .= l Aig ~g g
6
where i indexes the measured values belonging to a single data set s, g designates the energy group, and a delineates spectra that may be simultaneously adjusted. For example, relates a set of measured reaction rates Ri to a single spectrum
~g by the multigroup cross-section uig* {In this case, FERRET also adjusts the cross-sections.) The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
In the FERRET analysis of the dosimetry data, the continuous quantities
{i.e., fluxes and cross-sections) were approximated in 53 energy groups.
The calculated fluxes from the reference forward calculation were expanded into the FERRET group structure.using the SAND-II code [17]. This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 energy group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620 group spectrum was then easily collapsed to the group structure used in FERRET.
The cros~-sections wer~ also collapsed into the 53 energy group structure using SAND-II with calculated spectra {as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF-B/V dosimetry file. Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross-section. Correlations between cross-sections were neglected due to data and code limitations, but this omission does not significantly impact the results of the adjustment.
For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight. In some cases, as for the cross-sections, a multigroup covariance matrix is used.
More often, a simple parameterized form is employed:
7
I ,
where Rn specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The fractional uncertainties Rg specify additional random uncertainties for g~oup g that are correlated with a correlation matrix:
The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 1 (8 specifies the strength of the latter term).
For the a priori calculated fluxes, a short-range correlation of 1 = 6 groups was used. This choice implies that neighboring groups are strongly correlated when 8 is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R.E.
Maerker [18]. Maerker's results are closely duplicated when 1 = 6.
For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.
In performing the least squares adjustment with the FERRET code, the neutron flux spectrum at the center of the 20° surveillance capsule from the Cycle 1-5 transport calculation (Table 1) was input to the analysis.
The specific assignment of uncertainties in the measured reaction rates and the input (a priori) spectra used in the FERRET evaluations was as follows:
REACTION RATE UNCERTAINTY 5-10 percent FLUX NORMALIZATION UNCERTAINTY 30 percent 8
FLUX GROUP UNCERTAINTIES (E > 0.0055 MeV) 30 percent (0.68 eV < E <'0.0055 MeV) 58 percent (E < 0.68 eV) 104 percent SHORT-RANGE CORRELATION (E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 FLUX GROUP CORRELATION RANGE (E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 The reaction rate uncertainty e$timates include both a statistical (counting) uncertainty and systematic uncertainty. The latter is based on known errors such as uncertainty in counter efficiency, unknown errors that are derived from consistency of counting results, and estimates of error arising from the power history and corrections for competing reactions.
It should be noted that the uncertainties listed for the upper energy ranges extend down to the lower range. Thus, the 58 percent group uncertainty in the second range is made up of a 30 percent uncertainty with a 0.9 short-range correlation and a range of 6, and a second part of magnitude 50 percent with a 0.5 correlation and a range of 3. These input uncertainty assignments were based on prior experience in using the FERRET least squares adjustment approach in the analysis of neutron dosimetry.
from surveillance capsule, reactor cavity, and benchmark irradiations.
The values are liberal enough to permit adjustment of the input spectrum to fit the measured data for all practical applications.*
9
Results of the FERRET analysis are summarized in Table 2, and a comparison of the FERRET adjusted values with the measured reaction rate values is given in Table 3. The FERRET result for the flux (E > 1 MeV) at the surveillance capsule center is 6.88El0 n/cm 2-sec which is quite close to the evaluation in the original capsule report (6.95El0). The uncertainty in FERRET least squares fit value is 9%. The quality of the derived value is supported by the excellent agreement of the reaction rates calculated with the adjusted flux-spectrum with the measurements.
Comparison of Calculations with Measurements The value for flux (E>l MeV) at the surveillance capsule location for Cycles 1-5 was calculated to be 7.36 ElO. This value may be compared to the reported measured value of 6.95 EIO which indicates that the calculated value is conservative by 6%. *The reported values of cavity dosimetry measurements for Cycle 8 also support a conclusion that the calculated flux (E>l MeV) is copservative, in this case by an average of 4%. Uncertainties in the cavity dosimetry measurements were calculated to be about 6%.
Based on the benchmarking evaluations and the measurements, the uncertainty in the vessel fluence (E>l MeV) evaluation is about 10%
(la) when the measurement error and calculational extrapolation error is included. It is also concluded that the calculated fluence is conservative by about 5%. Thus, a high confidence can be attached to the use of the values in Table 1 to provide a conservative prediction of the current vessel fluence.
The agreement between the calculated and measured results for the Palisades in-vessel and ex-vessel cases is somewhat better than what has been achieved in similar plants [15]. In some plant designs, a consistent*
bias of about 13% has been observed with the calculations falling below the measurements. That this effect is not seen in these calculations could be due to differences in method of fuel power distribution calculation or plant geometry uncertainties. One such uncertainty is the vessel diameter for which the design value of 172 inches was used in these
. 10
- .,,,, f calculations. Although the cause of the biases in flux calculations are not presently understood, the use of the actual plant measurements to benchmark the calculations largely eliminates such biases as a source of uncertainty. The continuing program at Palisades for in-vessel and cavity measurements will support the continuing accuracy of the cycle-by-cycle vessel fluence determination.
11
~----
At
- th~
Table 1 Palisades Fluence Through Cycle 9 Reactor Vessel Clad-Base Metal Interface Cycle Cycle Cycle Cumulative Cycle Length Fl~x Fluen2e Flu2nce CEFPDl (n/cm /s) Cn/cm } (n/cm }
O Degrees 1 379.4 4.59E+10 1. 50E+18 1.50E+l8 2 449.1 4.59E+10 1.78E+l8 3.28E+18 3 349.5 4.59E+10 l.39E+18 4.67E+18 4 327.6 4.59E+l0 1.30E+18 5.97E+18 5 394.6 4.59E+10 1. 56E+18 7.53E+18 6 333.4 4.87E+10 1.40E+18 8.94E+18 7 369.9 4.87E+10 1. 56E+l8 1.05E+19 8 373.6 2. lOE+lO 6.78E+17 1.12E+l9 9 304.2 2.09E+10 5.49E+17 1.17E+19 16 Degrees 1 379.4 6.03E+10 1. 98E+l8 1.98E+18 2 449.1 6.03E+10 2.34E+18 4.31E+18 3 349.5 6.03E+10 1. 82E+l8 6.13E+l8 4 327.6 6.03E+10 1. 71E+18 7.84E+18 5 394.6 6.03E+10 2.05E+18 9.89E+18 6 333.4 6.25E+10 1. 80E+18 1.17E+19 7 369.9 6.25E+10 2.00E+l8 1.37E+19 8 373.6 4.76E+10 1.54E+l8 1.52E+19 9 304.2 3.05E+10 8.01E+17 1.60E+19 30 Degrees 1 379.4 4.70E+10 1. 54E+l8 1. 54E+18 2 449.l 4.70E+10 1. 82E+l8 3.36E+18 3 349.5 4.70E+10 1.42E+l8 4.78E+18 4 327.6 4.70E+10 1.33E+18 6.11E+18 5 394.6 4.70E+10 1.60E+18 7. 71E+18 6 333.4 4.79E+10 1.38E+l8 9.09E+l8 7 369.9 4.79E+10 1.53E+l8 1.06E+19 8 373.6 2.28E+10 7.36E+17 1.14E+l9 9 304.2 1. 99E+l0 5.23E+17 l.19E+l9 45 Degrees 1 379.4 2.98E+l0 9.78E+l7 9.78E+l7 2 449.1 2.98E+l0 l.16E+l8 2.13E+l8 3 349.5 2.98E+l0 9.00E+17 3.04E+l8 4 327.6 2.98E+l0 8.44E+17 3.88E+l8 5 394.6 2.98E+10 l.02E+18 4.90E+l8 6 333.4 3.03E+10 8.73E+17 5.77E+18 7 369.9 3.03E+10 9.68E+17 6.74E+18 8 373.6 1. 73E+10 5. 58E+17 7.30E+l8 9 304.2 l .14E+l0 . 3.00E+l7 7.60E+l8 12
~ ~
- Table 2 FERRET-SANDI! Results Palisades Capsule W-290 Cycle 1-5 Input and Adjusted Neutron Flux Energy Energy A Pr;or; Flux* Adjusted Flux % Uncerta~n Group {MeV) {n/cm**2/sec) {n/cm**2/sec) {l STD) 1 l.733E+Ol l.105E+07 9.487E+06 24.
2 l.492E+Ol 2.665E+07 2.268E+07 22.
3 l.350E+Ol l.186E+08 l.003E+08 19.
4 l.162E+Ol 2.969E+08 2.504E+08 16.
5 l.OOOE+Ol 7.098E+08 5.995E+08 14.
6 8.607E+OO l.274E+09 l.086E+09 12.
7 7.408E+OO 3.221E+09 2.777E+09 10.
8 6.065E+OO 4.612E+09 4.040E+09 10.
9 4.966E+OO 8*.119E+09 7.243E+09 9.
10 3.679E+OO 7.695E+09 7.020E+09 11.
11 2.865E+OO 1.234E+l0 l .155E+10 12.
12 2.231E+OO l.133E+10 1. 084E+10 14.
13 1.738E+OO l.126E+10 l.094E+10 17.
14 l.353E+OO 8.519E+09 8.343E+09 20.
15 L 108E+OO 1.166E+l0 l.152E+l0 23.
16 8.208E-01 1.054E+l0 1. 046E+l0 25.
17 . 6.393E-Ol 9.339E+09 9.280E+09 27.
18 4.979E-01 6. 716E+09 6.662E+09 29.
19 3.877E-01 7.644E+09 7.552E+09 31.
20 3.020E-Ol l.106E+10 1. 087E+l0 32.
21 1.832E-Ol 9.389E+09 9.168E+09 33.
22 1.lllE-01
- 7. 598E+09 7.373E+09 34.
23 6.738E-02 6.305E+09 6.082E+09 35.
24 4.087E-02 4.667E+09 4.478E+09 35.
25 2.554E-02 3.246E+09
- 3.101E+09 36.
26 1. 989E-02 2.626E+09 2.500E+09 36.
27 1. 503E-02 4.702E+09 4.464E+09 36.
28 9.119E-03 4.899E+09 4.642E+09 62.
29 5.531E-03 5.363E+09 5.074E+09 62.
30 3.355E-03 1.736E+09 l.641E+09 62.
31 2.839E-03 1. 680E+09 1. 587E+09 62.
32 2.404E-03 l.643E+09 1. 552E+09 62.
33 2.035E-03 4.797E+09 4.529E+09 62.
34 l.234E-03 4.671E+09 4.409E+09 62.
35 7.485E-04 4.556E+09 4.301E+09 62.
36 4.540E-04 4.482E+09 4.230E+09 62.
37 2.754E-04 4.663E+09 4.401E+09 62.
38 l.670E-04 4.738E+09 4.472E+09 62.
39 l.013E-04 4.750E+09 4.482E+09 62.
40 6 .144E-05 4.747E+09 4.480E+09 62.
Note: Energy is upper energy bound of group.
13
- Table 2 (Continued)
FERRET-SANDI! Results Palisades Capsule W-290 Cycle 1-5 Input and Adjusted Neutron Flux Energy Energy A Priori Flux* Adjusted Flux % Uncertain Group (MeV) (n/cm**2/sec) (n/cm**2/sec) '(l STD) 41 3.727E-05 4.722E+09 4.456E+09 62.
42 2.260E-05 4.678E+09 4.414E+09 62.
43 l .371E-05 4.633E+09 4.372E+09 62.
44 8.315E-06 4.580E+09 4.321E+09 62.
45 5.043E-06 4.517E+09 4.262E+09 62.
46 3.059E-06 4.433E+09 4.183E+09 62.
47 l.855E-06 4.293E+09 4.0SOE+09 62.
48 l.125E-06 4.028E+09 3.800E+09 62.
49 6.826E-07 3.807E+09 3.592E+09 106.
50 4.140E-07 3.957E+09 3.734E+09 106.
51 2.511E-07 I. 071E+l0 l.OlOE+lO 106.
52 I. 523E-07 l .811E+l0 l.709E+l0 106.
53 9.237E-08 3.932E+l0 3. 710E+l0 106.
Integral Quantity A Priori Value Adjusted Value Uncertainty FLUX, E> 0.0 MEV 3.295E+ll 3 .136E+ll 27.%
FLUX, E< 0.414 EV 7.209E+l0 6.802E+l0 86.%
FLUX, E> 0.1 MEV l.375E+ll l.319E+ll 16.%
FLUX, E> I. 0 MEV 7.353E+l0 6.876E+l0 9.%
DPA/SECOND I. 058E-10 9.872E-ll 10. %
14
- Table 3 Comparison Of Measured And Calculated Reaction Rates PALISADES Cycle 1-5 Capsule W-290 Reaction Reaction Rate (dps/nucleus) Ratio Cale/Meas Meas A Priori Cale Adj Cale A Priori Adj Cale Cu63{n,ct)Co60 l.02E-16 l.24E-16 l.05E-16 1. 21 1.03 Fe54(n,p)Mn54 8.59E-15 9.78E-15 8.66E-15 1.14 1. 01 Ni58(n,p)Co58 1. OBE-14 l.26E-14 l. llE-14 1.16 1.03 U238(n,f)Csl37 3.15E-14 3.25E-14 3.00E-14 1.03 0.95 Ti46{n,p)Sc46 1.68E-15 l .85E-15 1. 63E-l 5 1.10 0.97 15
- FIGURE 1 REACTOR GEOMETRY SHOWING A 45 DEGREE R,0 SECTOR FOR THE INNER PART OF THE MODEL 140 120 100 FUEL EXCEPT SS PINS IN CYCLE 8 c
- 10 T
I FUEL
- I REGION T
I s
- 10 SHROUD----*
40 20 FUEL (SS PINS CYCLE 8 ONLY)
A&OIUS ** C!NTllETEIS oo 16
I FIGURE 2 REACTOR GEOMETRY SHOWING A 45 DEGREE R,9 SECTOR FOR THE OUTER PART OF THE MODEL 250 225
. 200 STEEL _ _--..all\. CONCRETE 115 LINER SHIELD 150 125 INSULATION 100 BARREL 15 REACTOR CAVITY 50 25 IAOIUS ** CENTl*ET!IS **
17
I References
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EA-PTS-87-004 (1987), transmitted by FAX November Z, 1990.
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, Reactor Dosimetry: Methods. Appljcations. arid Standardization. ASTM S.Ie lQQ.l, 1989, pp 308-313.
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NUREG/CR-1861, 1981.
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WCAP-11104; NUREG/CR-4576t February 1987.
18
- 14. Lippincott, E. P., Anderson, S. L., and Fero, A. H., "Application of Ex-Vessel Neutron Dosimetry for Determination of Vessel Fluence 1l ,
Reactor Dosimetry: Methods, Applications, and Standardization, ASTM STP 1001, 1989, pp 147-154.
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\
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