ML17228B499

Nonproprietary Version of Application of Reactor Vessel Surveillance Data for Embrittlement Mgt.

ML17228B499
Person / Time
Site:	Saint Lucie
Issue date:	07/31/1993
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
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ML17228B498	List: ML17228B497 ML17228B499
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CEN-405-NP, NUDOCS 9605210540
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L-96-112 Enclosure 2 COMBUSTlON ENGINEERlNG OWNERS GROUP CEN-405-NP APPLICATION OF REACTOR VESSEL SURVEILLANCE DATA FOR EMBRITTLEMENTlVRMAGEMENT Prepared for the C-E OWNERS GROUP July 1993 da I555 ABB Combustion Engineering Nuclear Power P'L115N ASEA BROWN BOVERI 9605210540 9605i4 PDR ADQCK 05000335 P PDR IQ

WINDSOR, CONNECTICUT TABLE F E T

~ec i n No. ~Pe

SUMMARY

INTRODUCTION BACKGROUND IV. RATIONALE 12 INTEGRATED SURVEILLANCE APPROACH 20

'I. MARGIN REDUCTION APPROACH 29 VII. CONCLUSIONS 37 VIII. REFERENCES 39 APPENDIX A STATISTICAL ANALYSIS, PLATE DATA BASE AND RESULTS A-1 APPENDIX B STATISTICAL ANALYSIS, WELD DATA BASE AND RESULTS B-1 APPENDIX C EFFECT OF WELD FLUX LOT ON WELD CHEMICAL CONTENT C-1

LI T F FI

+i~mr T~il Decision Tree for Selection of Integrated Surveillance Approach (ISA) or Margin Reduction Approach (MRA)

Predicted versus Measured Weld RTNO~ Shift Results 16 Trend Curve for A533B Reference Material 24 Margin Reduction Approach 36 I T FTABL Summary of Basic Statistics for Charpy Shifts Difference Sample Sets 19

i. MMARY The surveillance program for Combustion Engineering (C-E) designed reactor vessels provides for the monitoring of irradiation damage in accordance with 10CFR50, Appendix H."'any of those programs, however, do not include the controlling vessel material as presently defined using Regulatory Guide 1.99, Revision 02.+ One consequence is that data from an approved Appendix H program can not be used for evaluation of vessel integrity issues. More specifically, direct application of Regulatory Guide 1.99, Position 2.1 to refine embrittlement predictions is not available for many vessels.

This report presents two approaches for C-E owners to apply Regulatory Position 2.1 in the specific case where a subject plant's surveillance data are credible in all respects except that the controlling material in the vessel gs not one of the surveillance program materials. If the controlling material of one reactor vessel is located in the surveillance program of another vessel, the Integrated Surveillance Approach (ISA) may be applied. When the controlling material of a vessel can not be traced to any other vessel's surveillance program, the Integrated Surveillance Approach can not be used and the Margin Reduction Approach (MRA) is applicable. Figure 1 gives a decision tree defining under what circumstances the ISA or MRA is to be used.

In the Integrated Surveillance Approach, controlling material data for the subject vessel is available from another C-E fabricated vessel (the "host" vessel) surveillance program (e.g.,

from a Westinghouse vessel).['

Once the preceding have been established, the chemistry factor and margin are determined for the subject vessel following Regulatory Position 2.1.

In the Margin Reduction Approach, plant-specific surveillance data are used to reduce the margin to he added to the predicted shift. P Rationale is provided to support use of the two approaches. [

Figure 'I Decision Tree for Select.ion of I nt egr at ed Sur ve t ance Approach (

I I I SA) or Marg i n Reduct. i on Approach (MBA)

L imi t i ng Mater i a I i n Surve i I ance Capsu e I I YES Follow Regulatory Pos l t. l on 2. 1 Approach Limit.ing Mat.erial in Sist.er Vessel Surveillance Program Fo I l ow I SA per YES CEN-405P, Section V NO Follow MRA per CEN-405P,Section VI The reactor pressure vessel is designed such that its fracture toughness will be sufficient to provide adequate margins of safety against brittle fracture during its service life. Thus, the original construction employed thick section low alloy steel base and weld materials which were inherently tough, as characterized by the initial reference temperature, RT>>Y.

Particular attention was given to the reactor vessel beltline, the region of the reactor vessel that surrounds the effective height of the active core. This region is exposed to a relatively high level of neutron irradiation which, over time, will reduce the toughness of (i.e.,

embrittle) the base and weld materials. Each operating plant is required to have a reactor vessel surveillance program which monitors those irradiation induced changes in the toughness properties of the beltline materials.

Neutron irradiation embrittlement of the reactor vessel beltline is addressed for both normal operation and for design base accidents. Heat-up and cool-down limits on pressure and temperature are adjusted to account for the predicted irradiation induced elevation of RT~Y.

Accident analyses ascertain that vessel integrity will be maintained in the event of a postulated transient, such as pressurized thermal shock <'", despite predicted embrittlement of the vessel. In either case, predictions of irradiation embrittlement are based on Regulatory Guide 1.99<". In situations where "credible" surveillance program results are available, those predictions for establishing operating limits can be adjusted based on the surveillance data.

This report addresses the specific case where a plant's surveillance data are credible in all respects except that the controlling material in the vessel is not one of the surveillance program materials. Two prescriptive approaches are developed in order to maximize the value from plant specific surveillance measurements. In the Integrated Surveillance Approach, the chemistry factor and margin is adjusted using surveillance data from another Combustion Engineering (CE) fabricated reactor vessel of CE or Westinghouse design. In the Margin Reduction Approach, plant-specific surveillance data are used to reduce the margin to be added to the predicted shift, Rationale is provided to support use of Regulatory Position 2.1 of Regulatory Guide 1.99"'n these specific cases. Supplemental criteria are provided for demonstrating the viability of each approach.

Surveillance programs were designed to provide a means of ~itin irradiation behavior of reactor vessel beltline materials. The concept in the 1960's was to measure the extent of embrittlement to verify the original design estimates. Surveillance capsules contained monitors to measure peak temperature and neutron flux, and test specimens to measure changes in strength and toughness. The test specimens were from materials selected to

~re re en the beltline materials, where selection criteria reflected the then current understanding of radiation embrittlement trends.

Presently, surveillance program requirements are given by 10CFR50, Appendix H+ and, by reference, ASTM E185-82~'. The stated purpose is still to monitor property changes with the addition that the resultant data are to be used in support of 10CFR50, Appendix G analysis. Regulatory Guide 1.99<" provides a means for predicting RT>>~ shift based on the of the

~

chemical content vessel material and the neutron fluence. ~ The Guide also presents a

~

method, Regulatory Position 2.1, by which credible surveillance data can be used to refine

~

~

the shift prediction and to reduce the uncertainty factor (margin) which must be added to the mean predicted shift.

A dichotomy exists given surveillance programs designed in the 1960's and early 1970's and the current Regulatory Guide 1.99"'. The means by which surveillance materials were selected for encapsulation differs substantially from the method currently prescribed in the Guide. For example, for the Palisades reactor vessel, the beltline plates were compared on the basis of drop weight NDTT and the Charpy impact test results;+ the plate with the highest NDLT and the highest temperature at the 30 ft-lb Charpy impact value was selected for inclusion in the surveillance program. The assumption was that differences in the initial toughness would be retained after irradiation for plates purchased to the same specification.

It was not recognized at the time that small differences in residual chemistry content could result in a large difference in irradiation sensitivity (NDTT shift). If the same plates were evaluated using the current Regulatory Guide 1.99"', a more rigorous analysis would be performed on the basis of RTNpy and irradiation induced changes in the transition temperature (shift) and upper shelf energy. The plate selected on this basis for the surveillance program would very likely be different from the one originally selected.

Therefore, the post-irradiation surveillance plate data would not be credible because it would not meet the first Regulatory Guide 1.99 credibility criterion:

"Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement according to the recommendations of this guide.""'ence, the dichotomy exists; the non-credible reactor vessel surveillance data cannot be used to compute the best-fit chemistry factor, and any further use of the surveillance data must be justified to the NRC even though the surveillance program complies with the version of ASTM E185 in effect at the time the program was designed.

This situation exists for many reactor vessels because of differences in methods to define

~ ~ ~

initial toughness and to predict shift. RT~~ was used in the surveillance material selection

~

~

process for those vessels built to the Summer 1972 Addenda to the ASME Boiler and Pressure Vessel Code."@ For earlier vessels, NDTl'r a Charpy index temperature was used to differentiate initial toughness properties of the candidate materials. Shift predictions for material selection were based on ranges of copper content; separate trend curves were used for 0 to 0.10% Cu, 0.11 to 0.15% Cu, and greater than 0.15% Cu. Typically, the beltline plates from one vessel would all fall within one range of copper content and, therefore, be predicted to exhibit the same shift. The beltline welds would have a similar situation. In contrast, Regulatory Guide 1.99"'redictions are based on explicit values of copper and nickel content such that each beltline plate and weld would have a unique predicted shift. Given the major differences in methods used to select the surveillance material and present day criteria"'or identifying the controlling material, the probability of having the precise controlling material in the surveillance capsule is low. As a consequence, present rules will severely limit the number of reactor vessels to which Regulatory Position 2.1tu can be applied.

Reactor vessels designed and built by Babcock & Wilcox encountered a problem with the surveillance program which prompted the establishment of an integrated surveillance program (see for example Reference 6). The problem involved both the forced removal of surveillance capsules from several reactor vessels and the need for continued surveillance of the beltline materials from those vessels. The B&W situation was the impetus for Section II.C of 10CFR50, Appendix H@ which detailed requirements and acceptance criteria for establishing an integrated surveillance program. The basic approach entails irradiation of representative material in a host reactor for use by other reactors having similar design and operating features. Presumably, that data could also be used in accordance with Regulatory Position 2.1<u for Appendix G"'nalyses.

In the B&W situation, numerous vessels were fabricated using similar materials and processes, including those vessels made for use in Westinghouse designed PWRs. Similar to B&W, Combustion Engineering also fabricated vessels for Westinghouse. Therefore, surveillance materials from Westinghouse designed vessels represent a potential source of data on specific heats and types of vessel beltline materials. Establishment of an integrated surveillance program between two vessels supplied by the same fabricator could, in numerous cases, provide surveillance data on the controlling vessel material for one or both of those vessels.

In contrast to the B&W situation, where some vessels had to have the surveillance capsules removed, all of the C-E designed PWRs still have surveillance capsules. Therefore, measurements of neutron flux, irradiation temperature, and surveillance material irradiation sensitivity can be obtained for each C-E vessel. This provides for the monitoring required by Appendix H,"'nd implementation of an integrated surveillance program could provide the data on the controlling material as input for analyses required by Appendix G"'nd 10CFR50.61.""

The objective of this report is to establish two approaches for implementing Regulatory Position 2.1"'or C-E designed PWRs for which the surveillance program is credible in all respects except that the controlling material in the vessel is not in the surveillance program.

In the Integrated Surveillance Approach, data for the controlling material is obtained from another reactor vessel surveillance program. In the Margin Reduction Approach, the plant-specific data are used when data on the controlling material are unavailable. Both approaches are intended to address the Regulatory Position 2.1"'ase in which "...

surveillance data are credible in all respects except that the (surveillance) material does not represent the critical material in the vessel..." The primary purpose is to realize as fully as reasonable the benefits from the surveillance program once certain criteria have been satisfied. Those criteria include the surveillance data credibility factors given in Regulatory Guide 1.99<" and, where applicable, the criteria given in 10CFR50, Appendix H+ for an integrated surveillance program. The added certainty obtained through surveillance capsule measurements justifies realization of the benefits from application of Regulatory Position 1

0)

The purpose of this section is to describe the rationale used in establishing the Integrated Surveillance and Margin Reduction Approaches.

(Note that the second and third rationale elements are directed primarily at the Integrated Surveillance Approach.) The three elements are detailed below.

PREDICTED VERSUS MEASURED WELD RTNDT SHIFT RESULTS 400 0 SURVEILLANCE DATA 350 EX P ERIMENTAL DATA (WELD-1, LCP) 300 250

~r o

rr rQ 200 0

z 150 100 POSITIOH 2.1. RG 'I.99 REY. 2 CURVE FIT TO SURVEILLANCE WELD DATA WITH lrr = 28F BOUNDS 50 0.1 0.2 0.4 0.6 0.8 1.0 2.0 4.0 6.0 8.0 10.0 ~

FLUENCE (E > 1 MeV). 10 n/cm Table 1 V. I E RATED R EILLA E APPR A H The purpose of this section is to establish the criteria which need to be addressed and the procedure to be followed in order to utilize surveillance data from another reactor vessel in support of a Regulatory Position 2.1<" analysis. The approach combines the credibility criteria of Regulatory Guide 1.99tu with the concept of integrated surveillance programs defined in 10CFR50, Appendix H.o'rior to applying Regulatory Position 2.1 per this approach as shown in Section V.E., the following must be established: traceability of the controlling material (Section V.A.), the credibility of the subject vessel (Section V.B.) and host vessel surveillance data (Section V.C.), and the similarity of the irradiation environment of both vessels (Section V.D.).

-A '<<h 'di h h program is equivalent to the controlling material in the subject vessel.

'ec V el rv ill n D redi ili - The surveillance data from the subject reactor vessel must be credible in all respects except that the surveillance material does not represent the critical material in the vessel. This is established by satisfying the following five criteria which are taken from Reference 1 with modifications to items 1 and 5:

1. Materials in the capsule shall be representative of the,,reactor vessel beltline materials, including both base metal and weld metal.[

2. Determination of the 30 ft-lb index temperature and the upper-shelf energy shall be done unambiguously for both the irradiated and unirradiated Charpy data Two or more sets of post-irradiation surveillance data for both base and weld metal shall be available from the subject reactor vessel, and a Regulatory Position 2.1 analysis shall be performed. The measurements shall be within

+lo~ of the mean curve of the actual surveillance data, where a~ is 17'F for base metal and 28'F for weld metal."'f the fluence range is two orders of magnitude or greater, the measurements must be within +2r~.<'>

4. The irradiation temperature of the Charpy specimens shall be within J25'F of the vessel wall temperature at the cladding/base metal surface. " The Charpy specimen temperature shall be estimated based on evaluation of the temperature monitors included in the surveillance capsules or from heat transfer calculations. The vessel wall temperature shall be based on cold leg or vessel wall measurements.

5. One of the surveillance capsules used in the evaluation should include Charpy specimens from a standard reference material. (CEOG vessel surveillance programs use HSST01 for reference material.) The measured shift for the standard reference material shall fall within the scatter band (J2og of the data base for that material as given in Figure 3.""

C. H st Reac r urveillance Data r ibili - The principal interest in the host reactor data is the one surveillance plate or weld representing the subject reactor's controlling material. The following criteria are based on the five criteria from Reference 1, with modifications to items 1, 4 and 5. Each of the following criteria are to be satisfied to establish the credibility of data for that one material:

1. The traceability between the controlling material from the subject reactor and surveillance material from the host reactor is to have been established as detailed in V.A.

2. Determination of the 30 ft-lb index temperature and the upper-shelf energy shall be done unambiguously for both the irradiated and unirradiated Charpy data.<u

3. Two or more sets of post-irradiation surveillance data for the controlling material shall be available from the host reactor vessel, and a Regulatory Position 2.1 analysis shall be performed. The measurements shall be within

+la~ of the mean curve of the actual surveillance data, where o~ is 17'F for base metal and 28'F for weld metal."'f the fluence range is two orders of magnitude or greater, the measurements must be within +2a~.<'>

4, The irradiation temperature of the Charpy specimens shall be within J25'F of the vessel wall temperature at the cladding/base metal surface."'he Charpy specimen temperature shall be estimated based on evaluation of the .

temperature monitors included in the surveillance capsules if available. The vessel wall temperature shall be based on cold leg or vessel wall measurements.

5. Standard reference material (SRM) is not available in some host reactor vessel surveillance programs.

r However, if SRM is available for one or more host reactor capsules, the measured shift for the SRM shall fall within the scatter band (J2rg of the database for that material as given in Figure 3.""

i25 200 v v I i00 v

O7 i50 v

m a ao ta.

/'e 75 o

uO ill 100 / 50 I-CI l

I cc 50  % Survoillanco Capouloo HSSTOZ 0 Survolllanco Capauloa HSST01 25 h 8SR Xrradiat iona (HSSTO?)

V ORR-PSF Xrradlat lone (HSST03) 0.00 0.50 i.00 i.50 Z.OO 2.50 "'3.00 3.50 4.00 4.50 5.00<Ei9 Fluanco (6 > i.0 Hov)

FIGURE 3 TREND CURVE FOR A533B REFERENCE MATERIAL Embrittlement of the A533B reference material relative to the draft Reg. Guide 1.99, Revision 2. The values for HSST01 and HSST03 plates-are adjusted relative to HSST02 plate to account for differences in chemistry. The upper and lower curves are the 34oF uncertainty bounds (20) specified by Reg. Guide 1.99.

(Source: Reference 15)

D. I di ti n Envir nmen m ri n - Ascertain that the irradiation environment for the host reactor surveillance capsule is comparable to that for the subject reactor surveillance capsule using the following factors and provide a qualitative ranking of the two capsules in terms of the significance of any differences on RTNpr shift.

1. Reactor Coolant Inlet Temperature

2. Neutron Flux (

E. li i n f Re ul P i i n 2 1 -Once the four preceding items have been satisfactorily addressed, the host reactor surveillance data for the controlling material may be used to determine the adjusted reference temperature for that material in accordance with Regulatory Position 2.1"'s follows:

- 25-

The overall equivalence of the subject reactor vessel's controlling material and the host reactor vessel's surveillance material was established in Section V.A.

If the reported copper and nickel content of both materials is identical, then proceed to step 2. If, however, they are not identical, then the measured values of shift, ART>>T, shall be adjusted by multiplying them by the ratio of the chemistry factor for the vessel material to that of the surveillance material in accordance with Regulatory Position 2.1.

2. Fit the surveillance data to obtain the relationship of ART>>T to fluence using the following equation:

AzrNDT = (t"z) s<'-'""g'~

where:

CF = chemistry factor f = neutron fluence (10" n/cm', E ) 1 Mev)

To do so, calculate the chemistry factor, CF, for the best flt by multiplying each ART>>T (or the adjusted values from step 1) by its corresponding fluence factor, summing the products, and dividing by the sum of the squares of the fluence factors. The resultant value of CF is then to be entered into equation 1 for calculating ART>>T.

Note: If the host reactor surveillance data are less than predicted, but the subject reactor surveillance data are greater than predicted, then application of the CF derived from the host reactor could be non-conservative for the subject reactor. In this situation, the host reactor calculated chemistry factor can be adjusted as follows; CF(H) x QFF'~ = Adjusted CF CF (S,P) where:

CF(H) is the calculated CF for the host reactor from step 2 CF(S,C) is the calculated CF for the corresponding plate or weld for the subject reactor following step 2 CF(S,P) is the predicted CF for the same subject reactor material based on Table 1 or 2 of Reference 1 The larger of the two values, Adjusted CF or CF(H) should then be entered into equation 1.

3. Calculate margin as follows:

2 2 Margin = 2 or+a~ (2) where:

0; = standard deviation for the initial RT>>T 0~ = standard deviation for d,RTNDT

) gn cases where a generic value is assumed for SA 533B Class 1 or SA 302B (Modified) plates, 0; is to be determined based on the variance of the data used to derive the generic value.)

The standard deviation for shift, 0~, can be reduced to 8.5'F for base metal and 14'F for welds if the surveillance data credibility has been established in accordance with sections V.B and V.C per Regulatory Position 2. 1"'nd the irradiation environments have been established to be comparable in accordance with Section V.D.

4. Calculate adjusted RT~(ART) as follows:

ART = RT~ + hRT~~ + Margin where:

RTNDv initial RTN ~, measured or generic value b,RTND~ = shift, using revised chemistry factor in equation 1 Margin = value using equation 2 including values of 0; and 0~ from step 3

5. Document results of the evaluation following Section V.A through V. E, including assumptions employed and conclusions reached.

VI. MAR I RED TI N APPR A H The previous section addressed the special case in which an integrated surveillance program approach could be used to augment plant-specific surveillance data. The purpose of this section is to address the case in which surveillance data on the controlling material is not available from either the plant-specific program or another reactor vessel, but the surveillance data are credible in all other respects. An approach is given for determining how much the standard deviation for dRTND~, 0~, can be reduced based on the degree of credibility of the surveillance capsule measurements or, as phrased in Regulatory Position 2.1<", "depending on where the measured values fall relative to the mean calculated for the surveillance materials".

The following procedure presents a recommended approach for satisfying Regulatory Position 2.1"'n order to reduce the value of 0~.

The procedure which follows addresses two issues, representative materials and predictability of the surveillance measurements, as the basis for reducing the value of 0~.

A, r n iv M Each surveillance program for a C-E designed reactor vessel includes encapsulated Charpy specimens from a plate, weld, heat-affected-zone, and standard reference material (SRM). In accordance with the edition of ASTM E185 in effect at the time, the surveillance materials (exclusive of SRM) were selected to represent the reactor vessel beltline. The surveillance plate was selected from one of the six (typical) beitiine plates. [

The surveillance weld was fabricated using portions of two beltline plates following the same procedure and weld consumables as one of the beltline welds In summary 911 -Rgl Gdjd 199"'9 11* 1 jdgjgg credibility of surveillance data. The four Regulatory Guide criteria dealing with I) scatter of the Charpy data, 2) scatter about the dRTN ~ versus neutron fluence best fit curve, 3) capsule irradiation temperature, and 4) SRM test results shall be evaluated. Once those criteria are satisfied data predictability needs to be evaluated to provide the basis for reducing the value of 0~.

C. A lica i n f Re l P ii n 1 1 with Mar in Reduction - Once the two preceding items have been satisfactorily addressed, the known information on the controlling material may be used to determine the adjusted reference temperature for that material in accordance with Regulatory Position 1.1"'s follows:

l. Obtain the value of d,RT>>~ based on the chemistry factor of the controlling material and the fluence using the following equation:

P0.28-0.10 logj)

(pe (1) where:

CF = chemistry factor f= neutron fluence (10" n/cm', E ) 1 Mev)

To do so, calculate the chemistry factor, CF, for the controlling material using the known chemical composition as prescribed in Regulatory Position 1.1 of Reference 1. If the exact chemical composition is not known, use the guidelines of Reference 1 to determine the appropriate chemistry factor.

2. Calculate margin as follows:

Margin = 2 ol + (RFa~) (2) where:

0; = standard deviation for the initial RTi T 0~ = standard deviation for hRTNDT RF = reduction factor from Section VI.B and as described below (In cases where a generic value is assumed for SA 533B Class 1 or SA 302B (Modified) plates, 0; is to be determined based on the variance of the data used to derive the generic value.)

The standard deviation in dRT>>Y, 0~, as defined in Regulatory Guide 1.99 is:

0~ = 28'F for welds and o~ = 17'F for base metal. The Margin Reduction Approach may then be applied to the controlling material based on the results of testing of the available credible surveillance data to reduce the required value of 0~ as follows:

where:

P = predicted shift based on Equation 2 of Regulatory Guide 1.99+

M = measured shift forbase metal (longitudinalor transverse orientation) and weld 0~ = standard deviation for ART>>Y which is 17'F for plates and 28'F for welds<n

3. Calculate adjusted RT>>Y(ART) as follows:

ART = RTNDz + B,RT>Dr + Margin where:

RTND initial RT>>Y, measured or generic value ART>>Y shift, using equation 1 Margin value using equation 2 including values of 0; and 0~ from step 2 Th* I fhp&igd i bd&.TM should include demonstration that the surveillance materials are representative of the reactor vessel beltline materials as described in Section VI.A. The surveillance data credibility shall be established using data sets from two or more surveillance capsules and the four Regulatory Guide 1.99 criteria identified in VI.B. Finally, the surveillance material predictability and the resultant 0~ reduction factor shall be established following the procedure given in Section VI.C.

rrcuRE 4 MARnx REnVCYrOx A.ppROA.eH N L I Advancements in technology for predicting the extent of neutron irradiation embrittlement in pressure vessel steels have caused significant changes in the definition of the "controlling" vessel beltline material. One consequence is the surveillance capsules do not contain that controlling material. Regulatory Position 2.1 of Regulatory Guide 1.99"'ermits the use of data from the reactor vessel surveillance program to refine shift predictions but only for those programs which include the controlling vessel material.

This report provides the justification for employing surveillance data in support of vessel embrittlement analyses when the surveillance data are credible in all respects except the controlling material is not one of the surveillance program materials. Approaches are provided for Integrated Surveillance and Margin Reduction to maximize the value gained from plant-specific sur veillance measurements.

~

In the Integrated Surveillance Approach, data for the controlling material from a host reactor

~ ~

vessel surveillance program is used in conjunction with plant-specific data to adjust the chemistry factor and reduce the required standard deviation for shift. This data sharing between Westinghouse and Combustion Engineering reactor vessel surveillance programs is justified based on the following rationale:

The Integrated Surveillance Approach identifies criteria to be addressed and a procedure to follow in order to utilize the host reactor surveillance data in support of a Regulatory

~

~

Position 2.1<'>

~ analysis for a C-E vessel.

In the Margin Reduction Approach, the plant-specific surveillance data are used to reduce the margin for predicted shift of the controlling material based on the predictability of the surveillance measurements.

The two approaches presented in this report are applicable to surveillance programs for C-E designed reactor vessels, Their use is not intended for other vessel designs and surveillance programs because the approaches were based in part on practices and design characteristics

~

unique to C-E. ~

V IL

l. USNRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

2. American Society for Testing and Materials, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels" ASTM E185-82, July 1982.

3. 10CFR Part 50, Appendix H, "Reactor Vessel Materials Surveillance Program Requirements," Federal Register, Vol. 48, No. 104, May 27, 1983.

4, 10CFR Part 50, Appendix G, "Fracture Toughness Requirements," Ibid.

5.~ "Summary Report on Manufacture of Test Specimens and Assembly of Capsules for

~

Irradiation Surveillance of Palisades Reactor Vessel Materials," Combustion

~ ~

Engineering Report P-NLM-019, April 1971.

6. "Babcock & Wilcox Owners Group Program for Evaluation of Reactor Vessel Properties," BAW-1474, Rev. 4, December 1986.

7. American Society for Testing and Materials, "Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (DPA),"

ASTM E693-79, August 1979.

8. W. N. McElroy, "LWR Pressure Vessel Surveillance Dosimetry Improvement Program: LWR Power Reactor Surveillance Physic-Dosimetry Data Base Compendium," NUREG/CR-3319, prepared by Hanford Engineering Development Laboratory, HEDL-TME 85-3, August 1985.

9. R. K. Nanstad, et.al, "Effects of 50'C Surveillance and Test Reactor Irradiations on Ferritic Pressure Vessel Steel Embrittlement," presented at the 14th International ASTM Symposium on the Effects of Radiation on Materials, Andover, Massachusetts, June 1988.

10. "Evaluation of the First Maine Yankee Accelerated Surveillance Capsule," Effects Technology, Inc., Report CR75-317, August 15, 1975.

11. "Maine Yankee Nuclear Plant Reactor Pressure Vessel Surveillance Program-Capsule 263," Battelle Columbus Laboratories Report BCL-385-21, December 12, 1980.

12. "Analysis of the Maine Yankee Reactor Vessel Second Accelerated Surveillance Capsule," Westinghouse Report WCAP-9875, March 1981.

J. R. Hawthorne, J. J. Koziol, and S. T. Byrne, "Evaluation of Commercial Production A533-B Steel Plates and Weld Deposits with Extra-Low Copper Content for Radiation Resistance," NRL Report 8136, October 21, 1977.

14. J. R. Hawthorne, "Notch Ductility Degradation of Low Alloy Steels with Low-to-Intermediate Neutron Fluence Exposures," NRL Report 8357 (NUREG/CR-1053),

January 14, 1980.

15. F. W. Stallman, "Analysis of the SA302B and SA533B Standard Reference Materials in Surveillance Capsules of Commercial Power Reactors," NUREG/CR-4947, prepared by Oak Ridge National Laboratory, ORNL/TM-10459, January 1988.

16. American Society of Mechanical Engineering, Section III, "Nuclear Power Plant Components," of ME B iler nd Pre r V 1 e, New York.

"Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCA's with Loss of Feedwater for the Combustion Engineering NSSS," Combustion Engineering Report CEN-189, December 1981.

18. F. W. Stallman, et.al., "PR-EDB: Power Reactor Embrittlement Data Base, Version 1," NUREG/CR-4816, prepared by Oak Ridge National Laboratory, ORNL/TM-10328, June 1990.

19. 10CFR Part 50, Section 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Vol. 56, No. 94, May 15, 1991.

0 APPENDIX A STATISTICAL ANALYSIS PLATE DATA BASE AND RESULTS

APPENDIX A A.1 DATA BASE The reactor vessel surveillance plate data base used in the statistical analysis is listed in Tables A-1 and A-2 for Combustion Engineering and Westinghouse NSSSs, respectively.

- A The latter two tests (c and d) were performed as an overcheck on the more traditional F and t tests.

- A Tabl REACTOR VESSEL PLATE SURVEILLANCE DATA COMBUSTION ENGINEERING NSSS

Table A-1 (contin Table -2 REACTOR VESSEL PLATE SURVEILLANCE DATA WESTINGHOUSE NSSS

Table A- tinued) 0 Table A- tinued)

I 00 I

Table A- ntinued) tinued)

Table A-3 A302 B PLATE NICKEL CONTENT

~Ve el M ri 1 I ifi i n ikl n n o Yankee Rowe Upper Shell Plate 0.21 Test Material YA9 0.19 N/A HSST Plate 0.18 Big Rock Surveillance Plate 0.18 H. B. Robinson W-9807-3 0.10 W-9807-5 0.10 W-9807-9 0.15 W-10201-1 0.11 W-10201-2 0.2S W-10201-3 0.08 W-10201-4 0.09 W-10201-5 0.12 W-10201-6 0.09 Mean Value: 0. 14% Ni Range: 0.08 to 0.25% Ni Standard Deviation: 0.055% Ni

FIGURE A-I (PRED-ACT) CVT SHIFT:va CU CONTENT C-E 8 Meat,inghouae P lactea Cu Content. (%)

o C-E

-A + Meatinghnuse

0 HiGURK A-2 (PRED-ACT) CVT SHIFT us NX CONTENT C-E a Nest.3.nghouse elates Ni Cunt ent'. (% )

C-E

+ Mestinghnuse

'lGURE A-3 (PRED-ACT) CVT SHIFT va FAST FLUENCE C-E 8 Mant'.inghouaa Platen Faat Fluence (x18".18 n/cm" 2) o C F

+ Meetinghouae

-A FIGURE A-4 (PRED-ACT) CVT SHIFT vs NEUTRON FLUX C-E 8 Masts.nghuusa Plates Flux ( xi.8" 18/cm" 2 sec) o C-E

+ Mast inghouse

FIGURE A-5 (PRED-ACT) CVT SHIFT FOR C-E PLATES F v.aquanaut Hi at'.ng r am GVT Shif't D if'f'er enae (Deg. F)

-A FIGURE A-6 (PRED-AGT) GVT SHXFT FOR MEST. PLATES Fr equenc:g Hint'ogr am CVT Shit'0 Di,f'f'ar ant"a (Dag. F)

FIGURE A-7 (PRED-ACT) CVT SHIFT FOR ALL PLATES Frequenc:g Hi@tagl"am CVT Shif't Dif'f'er ent"e (Deg. F)

FIGURE A-8 (PRED-ACT) CVT SHIFT FOP PLATES C-E 8 Meat. Nor mal Prub. Deneitg Fane

<x e. eel>

CVT Shit't Dif'f'aranc:m C Dag. F )

C-E Meat inghouae

FIGURE A-9 (PRED-ACT) CVT SHXFT FOR PLATES Nar mal Pr obabi1ity Denaity Functinn (x e.eel)

CVT Shi f't Dif'f'er ence (Deg. F)

-A APPENDIX B STATISTICAL ANALYSIS WELD DATA BASE AND RESULTS

APPENDIX B B. 1 DATA BASE The reactor vessel surveillance weld data base used in the statistical analysis is listed in Tables B-1 and B-2 for Combustion Engineering and Westinghouse NSSSs, respectively.

- B The latter two tests (c and d) were performed as an overcheck on the more traditional F and t tests.

- B REACTOR VESSEL WELD SURVEILLANCE DATA COMBUSTION ENGINEERING NSSS

Table B- 1 (continu Table B-2 REACTOR VESSEL SURVEILLANCE DATA WESTINGHOUSE NSSS

Table B-2 (continu I

txf OO I

Table B-2 (continu FIGURE B-1 (PRED-ACT) CVT SHIFT vs CU CONTENT C-E 8 Westinghouse Welds Cu Content (%)

0 C-E

+ Westinghouse

-B FIGURE B-2 C

(PRED-ACT) CVT SHXFT vs NX CONTENT C- E 8 Mestinghouse 4le 1 ds Ni Cont ent. (%)

C-E

-B + Meat.inghuuae

FIGURE B-3 (PRED-ACT) CVT SHXFT vs FAST FLUENCE C- E 8 Mes t.inghuuse Me 1 ds Fast'. F luenc:e ( xi,8" 18 n/t"m 2) o C-E

+ Mestinghouse

-B FIGURE B-4 (PRED-ACT) CUT SHXFT va NEUTRON FLUX C-E 8 Meetinghouse Melds Flux (xi8"i,8 n/am"2 sec) 0 Q P

+ Meetinghouse

-B FIGURE B-5 (PRED-ACT) CUT SHIFT FOR C-E MELDS Frequency Histogram.

CUT Shit't'. Dif'f'erence (Deg. F3

-B FIGURE B-6 (PRED-ACT) CVT SHIFT FOR MEST. MELDS Fr equencLI H j.at'.ogr am CVT Shi f't. Dif'ter ence (Deg. F)

-B FIGURE B-7 (PRED-ACT) CUT SHIFT FOR ALL MELDS Fraquanc:LI Hiat ngram CUT Shit't'. Dif'f'er enae (Deg. F )

I C-E 8 FIGURE B-8 (PRED-ACT) SHXFT FOR WELDS Wast. Nurmal Prob. Dansi<g Fncs

~

Q 0

L 5

0 II~

lg 0

Q.

CVT Shi f't Dif'f'er ence (Deg. F)

C-E West inghouse

-B FIGURE B-9 (PRED-ACTi SHXFT FOR 4lELDS

<x e.eel>

Nor mal P I"ah@bi 1 itg Density Function CVT Shit't Dif'f'er ence (Deg. F )

-B APPENDIX C EFFECT OF WELD FLUX LOT ON WELD CHEMICAL CONTENT

APPENDIX C C. I INTRODUCTION

]

C.2 WELD FLUX TYPE

- C C.3 WELD FLUX LOT

-C TABLE C-1 Expected Effect of Flux Type on the As-Deposited Weld Properties Using the Same Weld Wire

TABLE C-2 As-Deposited Weld Chemistries as a Function of Flux Tyye

- C As-Deposited Weld Chemistries as a Function of Flux Lot (Weld Wire Heat ¹4P7869)

- C ll te 0

e-