ML16342A657

To Calculation N-NCM-97009, DCPP 1 & 2 Reactor Vessel Adjusted Reference Temperature at 16 EFPY for RCS Heating/Cooldown Curves

ML16342A657
Person / Time
Site:	Diablo Canyon
Issue date:	01/14/1999
From:	Brashear P PACIFIC GAS & ELECTRIC CO.
To:
Shared Package
ML16342A656	List: DCL-99-005, Forwards Responses to 981105 & 981118 RAIs Re LAR 98-06,to Revise RCS Pressure & Temperature Limit Curves to Be Applicable Through 16 Efpy.Rev 0 to Calculation N-NCM-97009, Included ML16342A657
References
N-NCM-97009, N-NCM-97009-R, N-NCM-97009-R00, NUDOCS 9901270074
Download: ML16342A657 (40)

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Nuclear Cor Management Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0 Pacific Gas 8 Electric Company Nuclear Core Management Calculation Note N-NCM-97009 Rev. 0 DCPP 1

8c 2 Reactor Vessel Adjusted Reference Temperature at 16 EFPY for RCS Heatup/Cooldown Curves Structure S stem or Com onent:

Reactor Coolant System (System 07)

~Pur ose: This calculation determines the adjusted nil-ductilitY-transition Reference Temperatures (ART) for DCPP Units 1 and 2 at 16 EFPY, for use in updating the RCS Heatup and Cooldown Curves.

Number of Sheets:

19 pages

~Si nature Disci line/De t Date Preparer:

Reviewer:

Approval:

/ /X~

Supervisor bCs-e ei<<

PFS-TA Per IDAP CF3.1D4, enter registered engineer's full name and registration number, or stamp, or seal in this space.

Related Revisions/Documents/Memoranda:

990i270074

'P9'0i2261 PDR; ADQCK 05000275

. PD~

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0 TABLE OF CONTENTS Page 1.

PURPOSE 2.

BACKGROUND 3.

METHODOLOGY 4.

RESULTS

SUMMARY

5.

NOMENCLATURE 6.

DESIGN INPUT 7.

CALCULATIONS 4

7.1 Beltline Materials 7.2 Fluence 7.3 Fluence Factors 7.4 Chemistry Factors 7.5 Nil-Ductility-Transition-Temperature Shift 7.6 Margin 7.7 Adjusted Reference Temperatures 8.

RESULTS 9.

CONCLUSIONS 10.

REFERENCES 13 TABLE 1.

TABLE 2.

TABLE 3.

TABLE 4.

TABLE 5.

TABLE 6.

Unit 1 Reactor Vessel Fast Fluence at 16 EFPY Unit 2 Reactor Vessel Fast Fluence at 16 EFPY DCPP-1 Reactor Vessel Adjusted RTNpT (1/4t) at 16 EFPY DCPP-2 Reactor Vessel Adjusted RTN>> (1/4t) at 16 EFPY DCPP-1 Reactor Vessel Adjusted RTN>> (3/4t) at 16 EFPY DCPP-2 Reactor Vessel Adjusted RTN>> (3/4t) at 16 EFPY 15 15 16 17 18 19 Page 1 of 19

Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0

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1.

PURPOSE This calculation determines the adjusted nil-ductility-transition Reference Temperatures (ART) for DCPP Units 1 and 2 at 16 EFPY for use in updating the RCS Heatup and Cooldown Curves (Technical Specification 3/4.4.9 Pressure/Temperature Limits) from 12 EFPY (present interval) to 16 EFPY.

ART values determined herein are applicable to the vessel wall 1/4 and 3/4 thickness locations, and are appropriate for use as input to ASME BRPV Section III Appendix G calculations.

2.

BACKGROUND The current RCS Heatup and Cooldown Curves (Technical Specification 3/4.4.9 Pressure/Temperature Limits) were submitted to NRC in Reference 1, based on the fluence calculations of References 2 and 3, which were based on surveillance capsule data from both units.

The technical specifications indicated the Pressure-Temperature (P/T) Limits are valid for 12 EFPY.

Following approval of the P-T Limit LAR, Reference 4 was received informing PG5E that NRC does not consider the DCPP 1 Surveillance Capsule S results to be credible.

Per Reference 5, plant specific surveillance results cannot be used in determining the ART unless there are data from at least two credible surveillance capsules.

Therefore, with only one credible surveillance data point (Capsule Y), revisions to Calculation Files 930715-0, and 930818-0 were performed (Reference 6 and 7) to exclude use of the Unit 1 surveillance data, which were used in deriving a plant-specific Chemistry Factor and relaxing the Margin term used in determining the Adjusted Reference Temperature (ART). The results of Calculation File 930715-1 showed the ART values for the Unit 1 materials calculated for 12 EFPY did not bound those ART (1/4T and 3/4T) values which were input to the Appendix G Heatup/Cooldown Curves analysis of record (Reference 8), if the data from Surveillance Capsule S

could not be credited. Reference 6 showed that the existing technical specifications were actually only valid for 11

~ 5 EFPY.

This calculation is being performed to support updating the P-T Limits to 16 EFPY, before DCPP reaches 11.5 EFPY of service.

The limiting ART at 16 EFPY is also an input parameter in calculating the enable temperature for LTOP, referenced in Technical Specifications 3/4.1.2.2, 3/4.1.2.4, 3/4.4.1.3, 3/4.4.1.4.1, 3/4.4.9.3, and 3/4.5.3, and will be used in extending applicability of these Tech Specs.

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0 3.

METHODOLOGY RT<pT for the reactor vessel increases with fast neutron fluence exposure.

Updating, or adjusting, RTN>> (ART) is performed by adding the initial unirradiated reference temperature (initial RTNpT), the expected shift in reference temperature (BRTNpT) due to neutron irradiation, and a margin term (M) i.a.w. the methodology provided in Reference 5.

Properties for the most limiting vessel beltline material will be used in determining the vessels'RT values.

4. RESULTS

SUMMARY

The adjusted reference temperatures (ART) for DCPP Unit 1 at 16 EFPY at 1/4t and 3/4t, are 180.2'F (183.7'F, if the chemical composition from Reference 14 is used) and 122,6'F, respectively, and were determined for the limiting beltline material, Lower Shell Longitudinal Weld 3-442 C.

The adjusted reference temperatures (ART) for DCPP Unit 2 at 16 EFPY at 1/4t and 3/4t, are 177.2'F and 151.4'F, respectively, as determined for the limiting beltline material, Intermediate Shell Plate B5454-2.

Therefore, for the 16 EFPY composite Pressure-Temperature Limits applicable to both units, the Unit 1 ART limits the Heatup/Cooldown Appendix G Curves at 1/4T, and the Unit 2 ART limits the Curves at 3/4T. The limiting ART values to be input to the 16 EFPY Appendix G analyses are 180.2'F (183.7'F, if the chemical composition from Reference 14 is used) and 151.4'F at 1/4T and 3/4T, respectively.

These ARTs exceed those input to the "12 EFPY" (11.5 EFPY in actuality, as discussed previously) Pressure-Temperature Limits (1/4t = 164'F, 3/4t = 141 'F); therefore, Technical Specifications 3/4.4.9 and the supporting analyses will require revision to support operation beyond 11.5 EFPY.

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Nuclear Core anagement Calculation Note NTS/Design Engineering Services/Technical Support Services N-NCM-97009 Revision 0

5. 'MOMCMCLATURE ART CF f

f.nm fif4~

f3I4t f,

ff X

RTNp~j b RTNOT RTNoT ai

- Adjusted Reference Temperature (also refered to as RT>>T) evaluated at vessel wall 1/4 and 3/4 thickness locations, at 16 EFPY, ('F).

- Chemistry Factor per Reference 5.

- Reactor vessel wall fluence at a particular location and time (n/cm ).

- Fluence at reactor vessel clad-base metal interface at 16 EFPY (n/cm ).

- Fluence at vessel wall 1/4 thickness location at 16 EFPY (n/cm ).

- Fluence at vessel wall 3/4 thickness location at 16 EFPY (n/cm ).

- Fluence at reactor vessel inner surface at 16 EFPY (n/cm

)

~

- Fluence Factor per Reference 5.

- Distance into the vessel wall from the inside diameter wetted surface, per Reference 5, (inches)

~

- Margin to be used in determining the Adjusted Reference Temperature to cover uncertainties per Reference 5 ('F).

- (or Initial RTNpT) Reactor vessel initial nil-ductility-transition reference temperature,

('F).

- Shift in nil-ductility-transition reference temperature due to vessel irradiation ('F).

- Adjusted Reference Temperature (ART) evaluated at vessel wall 1/4 and 3/4 thickness locations, at 16 EFPY, ('F).

- Standard deviation for the initial RTN>>, based on measured or generic data, per Reference 5 ('F).

- Standard deviation for 8 RT>>T, per Reference 5 ('F).

6.

DESIGM IMPUT CF:

See Tables 3-6 for value and Reference.

Vessel cladding thickness:

0.22 inch (Reference 8).

Cu & Ni: See Tables 3-6 for value and Reference.

Fluence (f): See Tables 1,2 for value and Reference.

Initial RTNQY. See Tables 3-6 for value and Reference.

Vessel shell basemetal thickness (t): 8.625 inches (Reference 8).

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0

7. 'CALCULATIONS 7.1 Belt!inc Materials The reactor vessel beltline materials are the plates and welds in the vicinity of the core which are exposed to a fast (E>1 MeV) neutron fluence which results in a significant increase in RT<<. "Significant" can be quantified with respect to 10CFR50 Appendix H, which requires materials monitoring (via a surveillance program) for exposures greater than 10 n/cm at EOL. Historically, the beltline materials have been considered to include only the shell courses opposite the active fuel region: the intermediate and lower shell courses.

However, since the upper shell course exposure at EOL is projected (Reference

9) to slightly exeed 10 n/cm, its'aterials will be included in this ART calculation.

The vessel materials with fluence greater than 10 n/cm at EOL for DCPP 1 & 2 are confined to the upper, intermediate and lower shell courses, and are listed in Tables 3-6, along with their material property parameters required for this calculation.

7.2 Fluence The fast fluences appropriate for use in heatup and cooldown curve calculations are the fluence values that occur at the 1/4 and 3/4 wall thickness postulated cracktip locations, at the peak axial neutron flux location in the core beltline, and at the appropriate azimuthal angle for the vessel plate or weld being evaluated.

These fluence values wilt be derived from the clad/base-metal best estimate fluences obtained from the DCPP surveillance capsule and reactor cavity dosimetry results.

7.2.1 Clad/Basemetal Fluence, f~

Fluences at 16 EFPY at the reactor vessel wall cladding/base-metal interface (f,+

)

for vessel azimuthal locations of 0', 15', 30', and 45're summarized in Tables 1 and 2, for Units 1 and 2, respectively.

These values were obtained from References 10 (Table 8.2-1) and 11 (Table 8.2-1) and are based on measured surveillance capsule and reactor cavity dosimetry data.

Fast fluxes used to determine these fluences were evaluated at the position of highest axial flux peaking for the beltline welds and plates, Page 5 of 19

Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0 7.2:3 Su'rface Fluence, f, The fast fluences at 16 EFPY at the vessel inner surface (wetted perimeter) are obtained from the base-clad interface fluences using the correction factor from Reference 12: 1.029, which accounts for neutron attenuation through the vessel stainless steel cladding.

fg 1 I029 fg)b~

For example, the inner surface fluence for Unit 1 at 45's:

f, = 1.029 (7,46 x 10

) n/cm f, = 7.68x10 n/cm The 16 EFPY inner surface fluence for Units 1

and 2 at reactor vessel azimuthal locations of 0', 15', 30', and 45', are determined similarly, and are summarized in Tables 1 and 2.

7.2.4 1/4t Fluence, f>14, The fluence values to be used for determining RT~, for use in Appendix G analyses are the fluences at the 1/4 and 3/4 thickness flaw locations, per 10CFR50, Appendix G.

The fluence at 1/4t can be obtained by attenuating the surface fluence according to the following equation from Reference 5:

(-0.24x) f,/4, f,xe

'or example, the Unit 1 fluence at a vessel azimuthal angle of 45', one fourth the way into the vessel wall basemetal (from the inside; i.e., 1/4t) at 16 EFPY is:

f f

(-0.24(0.2S(8.625) + 0.22))

inc =

s x e

fq14, 7.68 x 1 0' (0.5654) f>>4, 4.34x10'/cm Using this method, fast fluences at 1/4t, at 16 EFPY, and for reactor vessel azimuthal locations of 0', 15', 30', and 45', have been determined.

Results are summarized in Tables 1 and 2, and are also included in Tables 3 and 4.

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0 7.2:5 3/4t Fluence The fluence at 3l4t can be obtained by attenuating the surface fluence according to the following equation from Reference 5:

(-0.24x) f3/4t fs x e For example, the 3/4t fluence for Unit 1 at 45't 16 EFPY is:

(-0.24(0.75(8.625)

+ 0.22))

3/4t = sXe f3/4, = 7.68 x10'(0.2008) f3/4t 1.54 x 1 0 n/cm'8 Using this method, fast fluences at 3/4t, at 16 EFPY, and for reactor vessel azimuthal locations of 0', 15', 30', and 45', have been determined.

Results are summarized in Tables 1 and 2, and are also included in Tables 5 and 6.

7.3 Fluence Factors Fluence Factors (ff) are defined by Reference 5:

ff f( 28- 0.10)og f) where f is the appropriate fluence (e.g., f1/4,) in units of 10'/cm (E>1 MeV).

For example, for Unit 1 at an azimuthal angle of 45'e.g., weld 3-442-C), at 1/4t, and at 16 EFPY, ff is:

0 434(.28 - 0.10 Iog 0.434) ff = 0.768 Fluence Factors for the remaining beltline materials for Units 1 5 2 were determined similarly and are provided in Tables 3-6.

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0 7.4'hemistr Factors Chemistry Factors (CF) were initially determined in Reference 3 for the Units 1 & 2 intermediate and lower shell course materials, in accordance with Reference 5;

i.e., when two credible surveillance capsule data points are available, plant-specific Chemistry Factors are permitted in determining the Adjusted Reference Temperature.

For Unit 1, two surveillance capsules (Capsule S & Y) have now been evaluated; however, the NRC recently (Reference

4) classified the Capsule S results as noncredible.

Therefore, the chemistry factors from Reference 13, excluding allowance for surveillance capsule results, will be used in this analysis.

Chemistry Factors for the upper shell course materials were determined in Reference 6.

The Chemistry Factors for the Unit 1 reactor vessel are listed in Tables 3 and 5.

Also shown in Tables 3 and 5 are the results of a sensitivity analysis on the Unit 1 lower shell longitudinal weld 3-442 C (which historically has been the limiting materialfor RT<< in both the Unit 1& 2 vessels) weld properties, to assess the impact of slightly different copper and nickel chemistry reported in Reference 14 for weld wire heat 27204 (0,203 Cu, 1.018 Ni), compared to the currently assumed composition (0.198 Cu, 0.999 Ni, Reference 13).

The weld compositions reported in Reference 14 are the result of a Combustion Engineering Owners Group (CEOG) task to determine the "Best Estimate" chemistry for the welds used in C-E fabricated reactor vessels.

The C-E evaluation accessed all credible 27204 weld composition data points in the industry known to date, and determined what may potentially be perceived as "the" generic Best Estimate chemistry for this weld wire heat.

This task and its product were a commitment by GEOG to NRC, and represent an exhaustive search by C-E of the industry RPV weld fabrication and surveillance program test records.

While PG&E has not made any commitments to use the CEOG Report results, an evaluation of the reported chemistry results and impact on our limiting weld is prudent and will be performed in this calculation.

The Chemistry Factor for Weld 3-442 C based on the results from Reference 14, is determined from Table 1 of Reference 5 using linear interpolation for 0.203 Cu and 1.018 Ni, and is 226.8.

For Unit 2, Chemistry Factors were determined in Reference 15 for the intermediate and lower shell course materials, and include use of plant-specific surveillance results for Intermediate Shell Course Plate B5454-1, and Longitudinal Welds 2-201 A,B,C, based on the results of surveillance capsules U, X, and Y.

Chemistry Factors for the upper shell course materials were determined in Reference 6.

The Chemistry Factors for Unit 2 materials are listed in Tables 4 and 6.

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0

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Y 7.5'il-D'uctilit -Transition Tem erature Shifts The shift in RT>> is determined from the correlation in Reference 5:

ERTNpT = CF ff For example, for Unit 1 (weld 3-442 C), at the vessel wall 1/4 thickness location at EOL, hRTNpT is:

ARTNpT = 223.3(0.7680) dRTNpT = 170.7 'F d,RTN>>s for the remaining beltline materials for Units 1 & 2 at 16 EFPY were determined similarly and are provided in Tables 3-6.

7.6

~INar in "Margin", M as defined in Reference 5, is the quantity ('F) that must be added to obtain conservative, upper-bound values of Adjusted Reference Temperature for use in calculations necessary to meet the requirements of 10CFR50 Appendix G.

M = 24(cri

+ 0~ )

Gi is the standard deviation for the initial RTNpT and cr~ is the standard deviation for the shift in RTN>>. ai was determined in References 15 & 16 to be O'F when the initial RT<<was determined from measurement, and 17'F when the initial RT<<

was determined from generic data.

From Reference 5, a< is 28'F for welds and 17'F for plates, except that cr> need not exceed half the shift in RTNpz. Also, per Reference 5, a~ may be cut in half when results from two surveillance capsules are available and "credible" (as defined in Regulatory Guide 1.99, Revision 2) plant-specific Chemistry Factors are being used.

DCPP 1 has had 2 surveillance capsules removed and evaluated.

The surveillance data can be used to determine plant-specific chemistry factors and margins provided they meet the credibility criteria outlined in Regulatory Guide 1.99, Revision 2. These criteria are evaluated for DCPP 1, specifically:

1)

The surveillance capsule materials should be the reactor's limiting materials with respect to radiation embrittlement prediction using R.G. 1.99, Rev. 2.

For Unit 1 the surveillance weld metal represents Lower Shell Longitudinal Weld 3-442 C, which is the limiting material for RTndt and USE at end of life (EOL). The Page 9 of 19

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Nuclear Core i anagement Calculation Note N-NCM-97009 NTSIDesign Engineering Services/Technical Support Services Revision 0

. surveillance plate material is Intermediate Shell Plate B4106-3, which is equally lirriitingfor USE at EOL.

2) The Charpy energy versus temperature curves (irradiated and unirradiated) for the surveillance materials should show minimal scatter.

The Unit 1 Charpy specimens for the surveillance materials provide data at 8 test temperatures, and are sufficient for determination of RT<<at 30 ft-Ibs and the upper shelf energy.

3)

The scatter of RT<<shifts about a best fit line, as discribed in R.G. 1,99, Reference 2, Regulatory Position 2.1, for Intermediate Shell Plate B4106-3,is 23.7'F (Capsule S) and 14.5'F (Capsule Y). The ERT<<scatter for the surveillance weld is 32.6'F (Capsule S) and 21.8'F (Capsule Y). The bRT<<scatter for all Unit 1 surveillance materials is well under the prescribed 2a~ allowance of 34'F for the surveillance base metal and 56'F for the surveillance weld metal.

However, for Capsule S, the hRT<<scatter exceeds the surveillance database 1a~ standard deviation of 17'F for base metal and 28'F for welds and has been deemed not credible by NRC (Reference 4):

4)

The surveillance capsule irradiation temperature should match the vessel wall IO temperature within 25'F.

The surveillance capsules at Unit 1 are mounted on the exterior of the reactor thermal shield, opposite the reactor vessel inner wall.

Both the surveillance capsules and the vessel wall inner surface are in the vessel downcomer region, and are maintained very close to the nominal cold leg temperature of 547'F during operation.

The vessel wall cladding/base metal interface is well within 25'F of the capsule components.

5)

The surveillance data for the correlation monitor material in the capsules should fall within the scatterband for this material.

The surveillance Correlation Monitor IVlaterial (HSST Plate 02) material test results are very close to prediction, and therefore meet this criteria.

The surveillance data from DCPP Unit 1 meet Criteria 1-5 above for Capsule Y, but only Criteria 1, 2, 4, 5 5 for Capsule S.

Therefore, without at least 2 credible surveillance datapoints available, Section 1.1 of R,G. 1.99, Revision 2 (and not plant-specific data) must be used in determining the ART for Unit 1.

DCPP 2 has also had 2 surveillance capsules removed and evaluated.

The surveillance data meet the credibility criteria outlined in Regulatory Guide 1.99, Revision 2, as discussed in References 17 and 18, and can be used to determine plant-specific chemistry factors and margins (v~ can be cut in half when determining the appropriate margin term, M).

Based on these rules, a, and a~ values for the Units 1 5. 2 beltline materials are listed in Tables 3-6.

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0 For 'example, the Margin, M, for Unit 1 longitudinal weld 3-442 C is:

M = 24[(17)

+ (28) ]

M = 65.5'F The Margins for the remaining Units 1 5 2 beltline materials are determined similarly and are listed in Tables 3-6.

7.7 Ad'usted Reference Tem eratures The Adjusted Reference Temperature is the initial NDT Reference Temperature adjusted for material embrittlement for the exposure period being considered, and includes a margin term:

RTNDT RTNDTi + ~RTNDT +

For example, for the Unit 1 Longitudinal Weld 3-442 C at 1/4t and at 16 EFPY:

ART = -56 + 170.7 + 65.5 'F ART = 180.2'F ARTs for the remaining Units 1 5 2 beltline materials for the 1/4t and 3/4t locations, at 16 EFPY, were determined similarly and are listed in Tables 3-6.

8.

RESU LTS The adjusted reference temperatures (ART) for DCPP Unit 1 at 16 EFPY at 1/4t and 3/4t, are 180.2'F and 122.6'F, respectively, and were determined for the limiting beltline material, Lower Shell Longitudinal Weld 3-442 C.

The chemical composition sensitivity evaluation of this weld, using the generic chemistry data from Reference 14, resulted in an increase in ART of 3.5'F at 1/4t and 3.2'F at 3/4t (predicted ARTs of 183.7'F and 125.8'F at 1/4t and 3/4t, respectively), over those ART values calculated using only the measured composition data from Diablo Canyon reactor vessel materials.

These differences and their associated uncertainties are small compared to the large margin (65.6'F) term added to the calculated Reference Temperatures.

Since the purpose of the margin term is to account for uncertainties in measurement, composition, calculated fluence, etc., it Page 11 of 19

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0

.is concluded that either the plant-specific ARTs, or the generic ARTs, can be technicall'y justified for use in the DCPP P/T Limit Technical Specifications.

The adjusted reference temperatures (ART) for DCPP Unit 2 at 16 EFPY at 1/4t and 3/4t, are 177.2'F and 151.4'F, respectively, as determined for the limiting beltline material, Intermediate Shell Plate B5454-2.

Therefore, for the 16 EFPY composite Pressure-Temperature Limits applicable to both units, the Unit 1 ART limits the Heatup/Cooldown Appendix G Curves at 1/4T, and the Unit 2 ART limits the Curves at 3/4T, The limiting ART values to be input to the 16 EFPY Appendix G analyses are 180.2'F (183.7'F, if the chemical composition from Reference 14 is used) and 151.4'F at 1/4T and 3/4T, respectively.

9.

CONCLUSIONS The ARTs calculated for 16 EFPY exceed those input to the "12 EFPY" (11.5 EFPY in actuality, as previously discussed)

Pressure-Temperature Limits (1/4t = 164'F, 3/4t = 141 'F). Therefore, Technical Specifications 3/4.4.9 and the supporting analyses will require revision to support operation beyond 11.5 EFPY.

An LAR will need to be made to revise the Pressure-Temperature Limittechnical specifications.

In addition, the plant P-T Limits provide input assumptions to LTOP setpoint analyses.

Updating the P-T Limits will require the LTOP setpoints and analyses to be evaluated for continued applicability or the need for revision.

These actions are being tracked in A0408079.

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Nuclear Core anagement Calculation Note N-NCM-97'009 NTS/Design Engineering Services/Technical Support Services Revision 0

'tO 'EFERENCES 1.

2.

7.

8.

9.

PG&E Letter No. DCL-94-158, Chron ¹222814, Diablo Canyon Units 1

and 2 License Amendment Request 94-09 Revision of Technical Specifications 3/4.4.9.1 Figures 3.4-2 and 3.4-3, 3/4.1.2.2, 3/4.1.2.4, 3/4.4.1.3, 3/4.4.1.4.1, 3/4.4.9.3, and 3/4,5.3 RCS Pressure/Temperature Limits and LTOP Actuation Pressure and Enable Temperature",

W. H. Fujimoto to USNRC, August 17, 1994.

PG&E Calculation File No. 930818-0, "DCPP Reactor Vessel Fluence Projections for Input to RCS Heatup and Cooldown Curves at 12 EFPY",

P. F. Brashear, August 19, 1993.

PG&E Calculation File No. 930715-0, "DCPP 1 & 2 Reactor Vessel Adjusted Reference Temperature at 12 EFPY for RCS Heatup/Cooldown Curves and at End-Of-License for FSAR Update", P, F. Brashear, September 8, 1993.

USNRC to G. M. Rueger, "Diablo Canyon 1: Assessment of Diablo Canyon Surveillance Material For Issuance of Revision 1 of the Reactor Vessel Integrity Database",

PG&E Chron. 230563, June 28, 1996.

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

PG&E Calculation File No. 930715-1, "DCPP 1 & 2 Reactor Vessel Adjusted Reference Temperature at 12 EFPY for RCS Heatup/Cooldown Curves and at End-Of-License for FSAR Update", P. F. Brashear, July 23, 1 996.

PG&E Calculation File No. 930818-1, "DCPP Reactor Vessel Fluence Projections for Input to RCS Heatup and Cooldown Curves at 12 EFPY",

P. F. Brashear, July 18, 1996.

PG&E Calculation File No. 890206-0, "Appendix G Heatup/Cooldown Curves", M. D. Sullivan, Febuary 3, 1989.

PG&E Calculation File No. 921130-0, "DCPP 1 & 2 Reactor Vessel Fluence Projections for input to Charpy Upper Shelf Energy Analysis", P.

F. Brashear, November 30, 1992.

10. Westinghouse Electric Corporation, WCAP-14284, "Pacific Gas and Electric Company Reactor Cavity Neutron Measurement Program for Diablo Canyon Unit 1 - Cycles 1 through 6, January, 1995.

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0

11. Westinghouse Electric Corporation, WCAP-14350, "Pacific Gas and Electric Company Reactor Cavity Neutron Measurement Program for Diablo Canyon Unit 2-Cycles 1 through 6, November, 1995.

12. Westinghouse Letter PGE-88765 (Chron'23479),

Fluence at Vessel Surface", J.C. Hoebel to J.E.Thomkins, December 14, 1988.

13. Westinghouse Electric Corporation WCAP-13771, J.M. Chicots, "Evaluation of Pressurized Thermal Shock for Diablo Canyon Unit I",

July, 1993.

14. Combustion Engineering Report CE NPSD-1039, Rev. 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds",

Appendix A, "CE Reactor Vessel Weld Properties Database",

Vol. 1, June, 1997.

15. Westinghouse Electric Corporation WCAP-14364, P.A. Peter, "Evaluation of Pressurized Thermal Shock for the Diablo Canyon Unit 2 Reactor Vessel", August, 1995.

16. PGBcE Report 420-85.687, "Evaluation of Diablo Canyon Power Plant Reactor Vessel Materials by the NRC Pressurized Thermal Shock Screening Criteria", M.D. Sullivan, January 13, 1986.

17. M. D. Sullivan to J. Gisclon, "Surveillance Capsule X Results", January 29, 1991.

18. Westinghouse Electric Corporation, WCAP-14363, "Analysis of Capsule Y From the Pacific Gas and Electric Company Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program", August 1995

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Nuclear Core anagement Calculation Note N-NCM-97009 NTS/Design Engineering Services/Technical Support Services Revision 0 Table 1

Unit 1 Reactor Vessel Fast Fluence (E) 1MeV) at 16 EFPY (n/cm )

0'5'0'5's f3/4t 2.74 E+18 2.66 E+18 1.55 E+18 5.50 E+17 4.24 E+18 5.31 E+18 4.12 E+18 5.16 E+18 2.40 E+18 3.00 E+18 8.51 E+ 17 1.07 E+ 18 7.68 E+18 7.46 E+18 4.34 E+18 1.54 E+18 Table 2 Unit 2 Reactor Vessel Fast Fluence (E)1lVleV) at 16 EFPY (n/cm

)

00 15'0'50 fs 3 81 E+ 18 6.1 5 E+ 18 f,

3,70 E+18 5.98 E+18 fi 4~

2 15 E+18 3 48 E+18 fa 4, 7.65 E+17 1.24 E+18 7.79 E+18 7.58 E+18 7.57 E+18 7.37 E+18 4.40 E+18 4.29 E+18 1.56 E+18 1.52 E+18 Page 15 of 19

I

Nuclear Core Management Calcula ion c'l eI NTS/Design Engineering Services/Technical Su).art Services N-NCM-97009 Revision 0 Material (A)

(%)

TABLE 3 DCPP-1 REACTOR VESSEL ADJUSTED REFERENCE TEMPERATURE (1/4t) AT 16 EFPY Initial 16 EFPY 1 4 Ni

~(AI RTNpT (r

crc(

Margin Fluence (CI (Al BRTI(pT

(%)

CF( I

('F)

('F)

('F)

('F)

(n/cm )

(fI) l'F)

RTI(pr

('F)

Upper Shell Plate 84105-1 B4105-2 B4105-3 0.12 0.56 82.2' 28' 17 4.2 35.0 9.00E+ 16 0.1019 8.4 0.12 0.57 82.4' 9Q'7 4.2 35.0 9.00E+ 16 0.1019 8.4 0.14 0.56 98.2' 14+'7 5

35.4 9.00E+ 16 0.1019 10.0 71.4 52.4 59.4 Inter Shell Plate B4106-1 B4106-2 B4106-3 0.125' 0.53 85.3 0.12' 0.50 81.0 0.086' 0.476' 55.2

-10 0

17 34.0 4.34E+ 18 0.7680 65.5

-3 0

17 34.0 4.34E+ 18 0.7680 62.2 30' 17 17 48.1 4.34E+ 18 0.7680 42.4 89.5 93.2 120.5 Lower Shell Plate B4107-1 84107-2 B4 107-3 013 056 898 0.12 0.56 82.2 0.12 0.52'8 81.4 15 20

-22 0

17 0

17 0

17 34.0 4.34E+ 18 0.7680 69.0 34.0 4.34E+ 18 0.7680 63.1 34.0 4.34E+ 18 0.7680 62.5 118.0 117.1 74.5 Upper Shell Long.

Welds 1-442 A,B,C 0.19 0.97 215.7'A'20 Upper Shell to Inter Shell Weld 8-442 0

11 22.0 9.00E+ 16 0.1019 22.0 0.25 0.73 197.5'A'56' 17 10.1 39.5 9.00E+16 0.1019 20.1 24.0 3.6 Inter. Shell Long.

Welds 2-442 A,B 2-442 C 0.198' 0999' 222.3 0.198' 0999' 222.3

-56' 17 28 65.5 3.00E+ 1 8 0.6703 149.0 1 58.5

-56' 17 28 65.5 1.55E+18 0.5102 113.4 122.9 Inter. Shell to Lower Shell Weld 9-442 0.18 0.68 167.2

-56K'7 28 65.5 4.34E+ 18 0.7680 128.4 137.9 Lower Shell Long.

Welds 3-442 A,B 3-442 C 0.198' 0 999' 222.3

-56Ã'.198' 0 999' 222.3

-56' (0.203)'R (1.018)'R (226.8)'"

17 28 65.5 2.40E+18 0.6138 136.5 146.0 17 28 65.5 4.34E+ 18 0.7680 170.7 180.2 (1 74.2)'"

(1 83.7)'R A.

B.

C.

D.

Reference 6, unless otherwise noted.

Reference 13, unless otherwise noted.

Reference 16, unless otherwise noted.

Estimated per NUREG 0800.

Per 10 CFR 50.61.

Weld 3-442 C evaluated using Reference 14 chemistry.

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Nuclear Core Management Calcula ion NTS/Design Engineering Services/Technical Su).art Services r)

~

N-NCM-97009 Revision 0 Material TABLE 4 DCPP-2 REACTOR VESSEL ADJUSTED REFERENCE TEMPERATURE (1/4t) AT 16 EFPY Initial 16 EFPY 1 4t Cu Ni (A)

(A)

RT>>

(r(

rr~

Margin Fluence (C)

(Al BRTN>>

(%)

(%)

Cfl )

(of)

(of)

(of)

(of)

(n/cm )

(ff)

(of)

RT(r>>

('F)

Upper Shell Plate B5453-1 B5453-3 B5011-1R 0.11 0.60 74'.11 0.60 74' 0.11 0.65 74.8' 28 0

3.8 7.6 8.90E+ 16 0.1011 7.5 5Q'7 3.8 34.8 8.90E+16 0.1011 7.5 OQ'7 3.8 34.8 8.90E+ 16 0.1011 7.6 43.1 47.3 42.4 Inter Shell Plate B5454-1 B5454-2 85454-3 P.14'>>

P.65(>>

1P2.5 52 0.14 0 59 99 6 67 0.15 0.62 110.5 33 0

8.5 17 4.29E+ 18 0.7649 78.4 147.4 0

17 34 4.29E+ 18 0.7649 76.2 177.2 0

17 34 4.29E+18 0.7649 84.5 151.5 Lower Shell Plate B5455-1 85455-2 B5455-3 0.14 0.56 98.2 0.14 0.56 98.2 0.10 0.62 65.2

-15 0

15 0

17 0

17 0

17 34 4.29E+ 18 0.7649 75.1 94.1 34 4.29E+ 18 0;7649 75.1 109.1 34 4.29E+ 18 0.7649 49.9 98.9 Upper Shell Long.

Welds 1-201 A,B,C 0.22 0.87' 211.7'

-50 0

14 28 8.90E+ 16 0.1011 21.4

- 0.6 Upper Shell to Inter Shell Weld 8-201 0.18 0.68 167.2'A'56' 17 8.5 38 8.90E+ 16 0.1011 16.9 Inter. Shell Long.

Welds 2-201 A 2-201 B,C 0 22 0 87'>>

211.7

-50 0.22 0.87' 211.7

-50 0

14 28 0

14 28 2.15E+ 18 0.5868 124.2 102.2 4.40E+ 18 0.771 7 1 63.4 141.4 Inter. Shell to Lower Shell Weld 9-201 0.04 0.03 26.9

-56%'7 10.3 39.8 4.29E+ 18 0.7649 20.6 4.4 Lower Shell Long.

Welds 3-201 A,C 3-201 B 0.26 0.19 129.2

-56Ã'7 28 65.5 4.40E+ 18 0.7717 99.7 109.2 0.26 0.19 129.2

-56' 17 28 65.5 2.15E+18 0.5868 75.8 85.3 Reference 6, unless otherwise noted.

Reference 15, unless otherwise noted.

Reference 16, unless otherwise noted.

Estimated per NUREG 0800.

Per 10 CFR 50.61.

Welds 1-201 A,B,C are the same weld materials as Welds 2-201 A,B,C (Wire Heat 21935/12008 B4 Modified, Linde 1092 Flux, Lot 3869).

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Nuclear Core Management'Calculat'TS/Design Engineering Services/Technical Jrt Services N-NCM-97009 Revision 0 TABLE 5 DCPP-1 REACTOR VESSEL ADJUSTED REFERENCE TEMPERATURE (3/4t) AT 16 EFPY Material

)A)

N )A)

(%)

(%)

CF' Initial

)CI RT))or

('F)

)A) l'F) l'F)

Margin l'F)

Fluence (n/cm

)

16 EFPY 3 4 ART)ror l'F)

RTNor

('F).

Upper Shell Plate 84105-1 B4105-2 84105-3 0.12 0.56 82.2' 2&'7 2

34.2 3.20E+ 16 0.0476 3.9 0.12 0.57 82.4' 9'7 2

34.2 3.20E+16 0.0476 3.9 0.14 0.56 98.2'A'4' 17 2.4 34.3 3.20E+16 0.0476 4.7 66.1 47.1 53.0 Inter Shell Plate 84106-1 B4106-2 84106-3 0.125' 0.53 85.3 0.12N'.50 81.0 0.086' 0.476' 55.2

-10 0

17 34.0 1.54E+ 18 0.5087 43.4

-3 0

17 34.0 1.54E+ 18 0.5087 41.2 30+'7 17 48.1 1.54E+18 0.5087 28.1 67.4 72.2 106.2 Lower Shell Plate 84107-1 B4 107-2 84 I0/-3 0.13 0.56 89.8 0.12 0.56 82.2 0.12 0.52' 81.4 15 20 0 17 0

17 0

17 34.0 1.54E+ 18 0.5087 45.7 34.0 1.54E+ 18 0.5087 41.8 34.0 1.54E+ 18 0.5087 41.4 94.7 95.8 53.4 Upper Shell Long.

Welds 1-442 A,B,C 0.19 0.97 215.7'

-20 0

5 2 104-3 20E+16 0 0476 10 3 0.7 Upper Shell to Inter Shell Weld 8-442 0.25'.73 197.6 '56' 17 4.7 35.3 3.20E+ 16 0.0476 9.4

-11.3 Inter. Shell Long.

Welds 2-442 A,B 2-442 C 0.198' 0 999'0 222.3 0.198' 0 999' 222.3

-56Ã'7 28 65.6 1.07E+ 18 0.4305 95.7 105.2

-56' 17 28 65.5 6.50E+ 17 0.3081 68.5 78.0 Inter. Shell to Lower Shell Weld 9-442 0.18 0.68 167.2

-56Ã'7 28 65.5 1.54E+ 18 0.5087 85.1 94.6 Lower Shell Long.

Welds 3-442 A,B 3-442 C 0.198' 0.999N'22.3

-56Ã'7 28 65.5 8.50E+ 17 0.3854 85.7 95.2 0.198' 0.999' 222.3

-56' 17 28 65.5 1.54E+18 0.5087 113.1 122.6 (0.203)'R (1.018)'n (226.8)'0 (1 ${j.3)

(125 8)) )

A.

Reference 6, unless otherwise noted.

B.

Reference 13, unless otherwise noted.

C.

Reference 16, unless otherwise noted.

D.

Estimated per NUREG 0800.

E.

Per 10 CFR 50.61.

F,.

Weld 3-442 C evaluated using Reference 14 chemistry.

Pave IIIof'Ig

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r Nuclear Core Management Calculat

..e NTS/Design Engineering Services/Technical Sul.qort Services N-NCM-97009 Revision 0 Material TABLE 6 DCPP-2 REACTOR VESSEL ADJUSTED REFERENCE TEMPERATURE (3/4t) AT 16 EFPY Initial 16 EFPY 3 4 Cu Ni lAI

.IAI RTIlpr col aA Margin Fluence ICI lAI

~RTNor

(%)

(%)

CF

('F)

('F)

('F)

('F)

(n/cm )

(ff)

('F)

RTIioz

('F)'pper Shell Plate 85453-1 85453-3 85011-1R 0.11 0.60 74'*'.11 0.60 74'A'.11 0.65 74.8' 28 0

1.8 3.6 3.10E+ 16 0.0471 5'

17 1.8 34.2 3.10E+ 16 0.0471 OQ 17 1.8 34 2 3.10E+ 16 0 0471 3.5 3.5 3.5 35.1 42.7 37.7 Inter Shell Plate 85454-1 85454-2 85454-3 0.14' 0.65' 102.5 52 0.14 0 59 99.6 67 0.15 0.62 110.5 33 0

8.5 17 1.52E+ 18 0.5058 51.8 0

17 34 1.52E+ 18 0.5058 50.4 0

17 34 1.52E+ 18 0.5058 55.9 120.8 151.4 122.9 Lower Shell Plate 85455-1 85455-2 85456-3 0.14 0.56 98.2 0.14 0.56 98.2 0.10 0.62 65.2

-16 0

15 0

17 34 0

17 34 0

17 34 1.52E+ 18 0.5058 49.7 1.52E+ 18 0.5058 49.7 1.62E+ 18 0.5058 33.0 68.7 83.7 82.0 Upper Shell Long.

Welds 1-201 A,B,C 0.22 0.87'n 211.7R

-60 0

14 28 3.10E+ 16 0.0471 10.0 12.0 Upper Shell to Inter Shell Weld 8-201 0..1 8 0.68 167.2'

-66' 17 4

34.9 3.10E+ 16 0.0471 7.9

-13.2 Inter. Shell Long.

Welds 2-201 A 2-201 B,C Inter. Shell to Lower Shell Weld 9-201 O.22 O.87'"

0.22 0.87' 211.7

-60 211.7

-60 0

14 28 7.70E+ 17 0.3666 77.6 0

14 28 1.66E+ 18 0.5116 108.3 0.04 0.03 26.9

-66' 17 6.8 36.6 1.52E+ 18 0.5058 13.6 65.6 86.3

-5.8 Lower Shell Long.

Welds 3-201 A,C 3-201 8

~ 0.26 0.19 0.26

. 0.19 129.2

-56' 17 28 129.2

-56' 17 28 65:5 1.56E+ 18 0.5116 66.1 65.5 7.70E+ 17 0.3666 47.4 75.6 56.9 A.

Reference 6, unless otherwise noted.

B.

Reference 15, unless otherwise noted.

C.

Reference 16, unless otherwise noted.

D.

Estimated per NUREG 0800.

E.

Per 10 CFR 50.61.

F.

Welds 1-201 A,B,C are the same weld materials as Welds 2-201 A,B,C (Wire Heat 21935/12008 B4 Modified, Linde 1092 Flux, Lot 3869).

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