ML20140B475

Responds to 10CFR50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events. Results of Encl Calculations Justify Operation of Facility Through 41 Operating Yrs

ML20140B475
Person / Time
Site:	Point Beach
Issue date:	01/20/1986
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Harold Denton, Lear G Office of Nuclear Reactor Regulation
References
CON-NRC-86-008, CON-NRC-86-8, REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR TAC-59972, TAC-599728, TAC-59973, VPNPD-86-031, VPNPD-86-31, VPNPD-860-31, NUDOCS 8601270023
Download: ML20140B475 (25)

Text

.

8 l%sconsin Electnc roara coupar 231 W. MICHIGAN, P.O. BOX 2046 MILWAUKEE, WI 53201 VPNPD-86-031 NRC-86-008 January 20, 1986 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 -

Attention: Mr. G. Lear, Project Director PWR Project Directorate No. 1 Gentlemen:

DOCKET NOS. 50-266 AND 50-301 RESPONSE TO 10 CFR 50.61 PROTECTION AGAINST PRESSURIZED THERMAL SHOCK (PTS)

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 This letter responds to the Commission's regulations at 10 CFR 50.61,

" Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events". Plant-specific RT calculations have been performed for all materials throughout thgTgore height (beltline region) of the Point Beach Units 1 and 2 reactor vessels. The conclu-sion is that the RT values for all beltline materials in these reactor vessels wil{Tgot exceed the screening criteria defined in Part 50.61(b) (2) (270 F for plates, forgings, and axial weld materials and 300 F for circumferential weld materials) through the expiration date of the current operating licenses.

Furthermore, RT calculations were performed and are - nmitted here that demon![ hate that Point Beach reactor vessel RT values will not exceed the PTS screening criteria through 40 ye$I5 of operation from the date of issuance of the full-power license. These calculations support our " License Amendment Application No. 107 Extension of License Duration Point Beach Nuclear Plant, Units 1 and 2", submitted by Mr. C. W. Fay's letter to Mr. H. R. Denton dated June 28, 1985 (NRC-85-6).

, AD - J. Knight (1tr only)

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- Mr. H. R. Denton January 20, 1986

] Page 2 1

The attachment to this letter provides the results of the RT calculations for Point Beach, as well as the bases for the f$d$nce and material properties used in the RT calculations. The attachmentalsoprovidesbackgrounddobbkentationregardingcore i loading patterns, operator training and awareness of pressurized thermal shock, emergency operating procedures used to mitigate vessel integrity challenges, and Westinghouse Owners Group (WOG) programs that demonstrate the applicability of the generic PTS screening values to Westinghouse-designed plants.

We believe, based on comparison of our calculated RT values in Table 4 of the attachment to the PTS screening cr$Ibria, that this submittal justifies operation of the Point Beach reactor 4

vessels through 40 operating years. Please contact us if additional irformation is required.

Very truly yours,

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, C. W. Fay Vice President Nuclear Power Attachment Copy to NRC Resident Inspector 4

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a N. o ATTACHMENT RT CALCULATIONS AND BASES FOR PTS POINT BEACH NUCLEAR PLANT LICENSING BASIS FOR PRESSURIZED THERMAL SH0CK (PTS)

The NRC has established that RT values of 270'F for plates, forgings, and axial weld materials and 300 F b circumferential weld materials be used as screening criteria to determine the timing of plant specific evaluations of vessel integrity and of possible needed modifications to provide protection against PTS events.[1] For the purpose of comparison with this criterion, the value of RT for the ~ reactor vessel must be calculated for each weld and plate, or forgi k in the reactor vessel beltline. For each material, RT PTS is the lower of the results given by Equations 1 and 2.

Equation 1:

0 RT PTS = I + M + [-10 + 470 Cu + 350 Cu Ni] f .270 Equation 2:

0 RT PTS = I + M + 283 f .194 "I" is the initial reference temperature of the unirradiated material measured as defined in the ASME Code, NB-2331. If a measured value is not available, then a generic mean value is to be used: 0 F for welds made with Linde 80 flux, which includes all Point Beach reactor vessel beltli.le welds.

"M" is the margin to be added. In Equation 1. M=48 F if a measured value of I was used, and P=59 F if the generic mean value of I was used. In Equation 2, M=0 r if a measured value of I was used, and M=34 F if the generic mean value of I was used.

"Cu" and "Ni" are the best estimate respective weight percents of copper and nickel in the material.

"f" is the maximum neutron fluence in units of 1019n/cm2 (E greater than or equal to 1MeV) at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence for the period of service in question.

IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS Figures 1 and 2 identify and indicate the location of all beltline region materials for the Point Beach Unit I and 2 reactor vessels, respectively. The beltline region is defined to be "the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage".

Page 1 of 23

O DEFINITION OF PLANT-SPECIFIC MATERIAL PROPERTIES The pertinent chemical and mechanical properties of the beltline region plate, forging, and weld materials of the Point Beach Unit 1 and 2 reactor vessels are given in Table l'.

Chemistry values for the shell plates and forgings were derived from vessel fabrication test certificates and results from chemical analyses of surveillance capsule material performed by Westinghouse. The brackets [ ] following the entries in Table 1 indicate the references from which the data were taken.

The chemistry data for welds could not be derived in the same direct manner.

Fast neutron irradiation-induced changes in the tension, fracture, and impact properties of reactor vessel materials are largely dependent on chemical composition, particularly the copper concentration. The general variability in irradiation-induced property changes is further compounded by the variability of copper concentration within the weldments.

To address the variation in chemistry, West.inghouse, the Electric Power Research Institute (EPRI), Babcock & Wilcox (B&W), and others have performed reactor vessel beltline weld chemistry studies of B&W vessels. The Westinghouse Owners Group (WOG) has reviewed and evaluated the above sources of data for Westing-house reactor vessels manufactured by B&W and has compiled the "WOG Reactor Vessel Materials Data Base".[2] The WOG materials data base consists of the following primary files " WELD CHEMISTRY", " PLANT", and " FLUENCE". In the " WELD CHEMISTRY" and " PLANT" files, records may be retrieved by specifying data fields such as Wire Heat No., Wire Type, Flux Type, and Flux Lot No.

For the Point Beach reactcr vessel welds listed in Table 1, the WOG data base was searched for the exact weld wire heat, wire type, flux type, and flux lot.

These records were then averaged to obtain the best estimate Cu and Ni values to be used in the RT calcu!ations. In the case where there was minimal data

(<10 records)forthgThxactPointBeachweld,thesearchwasbroadenedby allowing all Linde 80 flux lots to be included in the search that were utilized with that particular weld wire heat. Again, these records were averaged to obtain the Cu and Ni values used in the RTPTS calculations for the particular weld.

The printouts from the various searches in the WOG data base, with all weld records utilized to obtain the numeric averages of Cu and Ni, are given in Appendix A to this attachment. The above procedure was utilized to obtain the best estimate of Cu and Ni values for all Point Beach reactor vessel beltline welds except SA-775/812, the axial weld in the intermediate shell of the Unit I vessel. For SA-775/812 a combination of weld wires and flux lots was used in manufacture, but in total, only four (4) records exist to document the chemistry of this weld. Therefore, the chemistry of SA-775 (Cu = .19) was used since it is more conservative than the average Cu value (Cu = .167) of SA-812.

REACTOR VESSEL NEUTRON FLUENCE As of December 1,1985 the Point Beach Units 1 and 2 reactors had been operated for a total of 10.68 and 10.66 Effective Full Power Years (EFPY), respectively.

Page 2 of 23 l

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Assuming a cumulative capacity factor of 80%, the periods of reactor operation until the expiration of the current licenses and until the expiration of the j revised licenses, if the license extensions requested in License Amendment No.  ;

107 are granted, can be determined. These periods and associated dates are l shown in Table 2. Also shown is the date corresponding to 32 EFPY, which corresponds to the originally presumed period of operation of 40 years at a cumulative lifetime capacity factor of 80%.

Figures 3 and 4 are excerpted from WCAP-10638, " Adjoint Flux Program for Point Beach Units 1 and 2,"[3] and show the maximum fast neutron (E>l MeV) fluence at the beltline weld locations as a function of full power operating time for Units I and 2, respectively. Table 3 results from converting the reactor operating periods in Table 2 to fluence utilizing Figures 3 and 4, as appropriate.

The following section, " Core loading Patterns", discusses the makeup and bases of the Point Beach fluence prediction curves, Figures 3 and 4.

CORE LOADING PATTERNS For Point Beach Unit 1 beginning with Cycle 8 in 1980, core loading patterns employed a Low Leakage Loading Pattern (LLLP) design, and assemblies with several previous cycles of burnup were positioned at certain locations on the core periphery. The LLLP design was fully implemented for PBNP Unit 2 in Cycle 7, also in 1980. Prior to this time, new fuel had been placed on the core periphery.

In 1984 the Westinghouse Adjoint Flux Program was utilized to assess the effects that past and present core management strategies have had on neutron fluence levels in the reactor pressure vessel. Figures 3 and 4 have been excerpted fromWCAP-10638[3] and represent the application of the " Adjoint Flux Program For Point Beach Units 1 and 2". In regard to Figures 3 and 4, the solid portions of the curves are based directly on calculations that utilized the PBNP plant specific core power distributions through fuel cycle 11 for Unit 1 and fuel cycle 10 for Unit 2. The dashed portions of these curves involve a projection into the future, based on the average neutron flux at the key loca-tions over the low leakage fuel cycles. For Unit I the neutron flux average over cycles 8 through 11 was used to project future fluence levels, while the neutron flux average over cycles 6 through 10 was employed for Unit 2.

Optimized fuel assemblies (0FA) were utilized at PBNP in core reloads beginning in the Fall of 1984 for Unit 2 and the Spring of 1985 for Unit 1. Complete transition to 0FA will occur at the fourth core reload of 0FA for each unit, specifically the Fall of 1987 for Unit 2 and the Spring of 1988 for Unit 1.

The fluence projections in Figures 3 and 4 are applicable to both 0FA and normal fuel, because the core power distributions of 0FA and normal fuel are basically the same. Hence, the plant specific fluence distributions, which are calculated by adjoint importance functions that directly relate the spatial distribution of fission density within the reactor core to the response of interest on the vessel, are unchanged.

The low leakage loading pattern described earlier and the transition to 0FA fuel have the benefit of both fuel cycle economy and lower neutron flux at the vessel wall. It is anticipated that PBNP will continue to operate at or below this level of leakage in the future.

Page 3 of 23

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RT VALUES FOR PBNP PTS RT@

50 calculations have been and are presented performed in Table according

4. Tables 1 andto 3requirements list the inputs of that 10 CFR were directly utilized in the RT equations. It is concluded from Table 4 that PBNPreactorvesselbeltlinhTbaterialsdonotexceedtheapplicablescreening PTS criteria through the expiration of the current operating licenses.

It is also concluded that the PBNP reactor beltline materials will not exceed the applicable PTS screening criteria during the operating license extension requested in Mr. C. W. Fay's letter to Mr. H. R. Denton dated June 28, 1985,

" Docket Nos. 50-266 and 50-301, License Amendments Application No. 107, Exten- ,

sion of License Duration, Point Beach Nuclear Plant, Units 1 and 2".

OPERATOR TRAINING AND AWARENESS OF PRESSURIZED THERMAL SHOCK The operators at Point Beach are trained to be aware of conditions which challenge vessel integrity. Since 1982, pressurized thermal shock has been directly and indirectly addressed in training conducted at Point Beach. The following chronology illustrates the PBNP continuing training effort:

June 1982 Supervisors and instructors were given overall scoping training on reactor vessel PTS by an engineer responsible for reactor vessel issues.

September 1982 Emergency operating procedures that deal with loss of coolant, steamline break, and steam generator tube rupture were reviewed with the operators in light of PTS considerations.

November 1982 All plant operators and some engineers attended background training on PTS, which included topics on reactor vessel construction; fundamentals of thermal-hydraulics, stress, and fracture mechanics; neutron embrittlement; and plant transient response.

1982-1984 Cycle update training for the operating crews considered PTS and vessel integrity concerns in review of Technical Specifications, industry events, and general plant limitations with respect to heatup, cooldown, and low temperature overpressure protection.

Cctober 1984 Classroom training on the new draft Westinghouse Owner's Group (WOG) Emergency Operating Procedures (EOPs) was held for all operating crews. As part on the review of the new E0P setup and network, the critical safety procedures (CSPs) were reviewed. The CSPs include procedures that directly focus on maintaining vessel integrity and prevention and mitigation of pressurized thermal shock.

Page 4 of 23 l

1

• The critical safety function " status tree" pertaining-to 4 vessel integrity, which directs the operator to the appro-priate CSPs is included in Appendix B. Also included in Appendix B is the current Point Beach PT (pressure-tempera-ture) limit curve for vessel integrity. These PTS operating limits are in addition to the Technical Specification (Appendix G) operational curves. The PT-limit curves define the symptom sets - cold leg temperature and RCS pressure - where  :

special attention to vessel integrity is required of the operator to prevent flaw initiation. The severity of the cooldown transient and resulting operator actions are priori-tized by color, with red being the most urgent.

Although there are numerous procedures that control reactor-plant transient (pressure and temperature) response, the 3

following E0P and CSPs directly respond to PTS conditions:

Procedure No. Title E0P-0 Reacter Trip or Safety Injection

, CSP-P.1 Response to Iminent Pressurized Thermal Shock Condition CSP-P.2 Response to Anticipated Pressurized Thermal Shock Condition Both Point Beach reactor vessel PT limit curves presently fall J into the WOG generic Category II Grouping, which consists of plants with an intermediate RT These curves were developed fromworkdevelopedfortheWeNIn.ghouse Emergency Response Guideline set, Revision 1.

- August 1985 Training cycle 85-6 provided classroom training to the operators on the background documentation that supports the new E0Ps.

- September / The new E0Ps, ECAs, and CSPs were utilized for the first October 1985 time in the simulator. All operating crews at PBNP spent 5 days in the simulator handling accident / transient scenarios k with thc CSPs.

+

November 1985 PBNP operators again received training on the reactor pressure e vessel and ir.ternals. The PTS enclosure for this training discussed vessel construction, materials, stress and thermal-

, hydraulic fundamentals, and PTS type events.

The initial operator training program provides a broad overview of materials and 1 theory in the reactor coolant portion of the systems description to the operator.  !

Integrated operations, transient and accident analysis, and E0P use are also l d

covered in the initial operator training program, which addresses PTS. l I

l 4

l Page 5 of 23 1

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Operations requalification training covers PTS training whenever the training needs analysis performed by PBNP Training Group indicates that retraining is appropriate.

APPLICABILITY OF PTS SCREENING VALUES TO WESTINGHOUSE DESIGNED PLANTS Westinghouse through the WOG has provided a number of submittals that demon-strate that the relative Trequency of PTS events in Westinghouse designed plants is no more likely than that considered by NRC for the generic plant. In addition, probabilistic fracture mechanics calculations by Westinghouse have demonstrated that severe cooldown transients will not result in producing significant flaw extension in Westinghouse designed plants when the vessel RT is < 310 F for longitudinal flaw orientation and <335*F for circumferen-ti$kTflaw orientation.

WCAP-10319, "A Generic Assessment of Significant Flaw Extension. Including Stagnant Loop Conditions, from Pressurized Thermal Shock of Reactor Vessels on W Nuclear Power Plants", December 1985, refines all previous WOG submittals on FTS. It also provides a rigorous treatment of transients that lead to loop stagnation. It describes the probabilistic methodology developed by Westing-house for treating the PTS issue and applying this methodology to W - designed PWRs. In short, it demonstrates for Westinghouse plants that transients in general, including those which lead to stagnant loops, can be treated in a prob-abilistic sense and do not represent " outliers" which would undermine the risk studies used to develop the PTS screening criteria.

CONCLUSION The RT values for Point Beach reactor vessel beltline materials do not exceed 9escreeningcriteriathroughforty(40)yearsofplantoperation.

Point Beach has implemented measures that prevent and/or mitigate the severity 1

of PTS transients. The PTS screening criteria, though generically developed, are applicable to Point Beach. Therefore, we conclude that the Point Beach reactor vessels with current fuel management programs are safe for continued operation.

Page 6 of 23

FIGURE 1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE POINT BEACH UNIT N0. 1 REACTOR VESSEL CIRCUMFERENTIAL SEAMS VERTICAL SEAMS Z.

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FIGURE 2 l IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE POINT BEACH UNIT NO. 2 REACTOR VESSEL CIRCUMFERENTIAL SEAMS TOP VIEW Z

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TABLE 1 POINT BEACH UNITS 1 AND 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni I UNIT 1 (Wt.%) (Wt.%) ('F)

Intermediate Shell Plate A-9811:(c)[4,5) 0.20 0.056 -2(a)

Lower Shell Plate C-1423:(c)[5,6] 0.12 0.065 -20(a)

Axial Weld - Intermediate Shell SA 775/812:(d) 0.19 0.63 0(b)

Weld Wire Heat Nos. IP0815/IP0661 Linde 80 Flux Lots 8350/8304 Axial Weld - Lower Shell SA-847: I') 0.25 0.55 0(b)

Weld Wire Heat No. 61782 Linde 80 Flux Lot 8350 Circumferential Weld - Intermediate:I') 0.20 0.55 0(b) to Lower Shell SA-1101 Weld Wire Heat No. 61782 Linde 80 Flux Lot 8350 UNIT 2 Intermediate Shell Forging 123V500:(c)[8,9] 0.09 0.70 40(a)

Lower Shell forging 122W195:IC)[9,10] 0.05 0.72 40(a)

Circumferential Weld - Intermediate:(*) 0.26 0.60 C(b) to Lower Shell SA-1484 Weld Wire Heat No. 72442 Linde 80 Flux Lot 8579 NOTES:

(a) The initial RT values for plates and forgings are estimated according to BranchTechnicggTPositionMTEB5-2.[7]

(b) The initial RTN values for welds are generic mean values defined by the PTSruleat10OfR50.61(b)(2)(ii).

(c) The chemistry values for the shell plates and forgings were derived from vessel fabrication test certificates and surveillance capsule chemistry measurements.

(d) The chemistry data for SA-775 was utilized since this will result in a conservative PTS calculation for this weld. See Appendix A.

(e) The chemistry values for these welds were derived from searches in the WOG data base [2] and represent the rounded, average values.

Page 11 of 23

TABLE 2 POINT BEACH UNITS 1 AND 2 ACCUMULATED YEARS (EFPY) 0F REACTOR OPERATION

• PREDICTED UNIT I PERIOD DATE EFPY To present 12/1/1985 10.68**

Until current license expiration (midnight) 7/19/2007 28.0 Until license expiration, if License Amend- 10/5/2010 30.6 ment No. 107 approved Original design EFPY value for 40 years of 7/27/2012 32 operation UNIT 2 To present 12/1/1985 10.66** -

Until current license expiration (midnight) 7/25/2008 28.8 Original design EFPY value for 40 years of 8/3/2012 32 operation

Until license expiration, if License Amend- 3/8/2013 32.5 ment No. 107 approved

• Assumes a cumulative capacity factor of 80% for predicted EFPY values.

• • Denotes actual years of reactor operation until December 1, 1985.

Page 12 of 23

J TABLE 3 POINT BEACH UNITS 1 AND 2 ACCUMULATED NF'lTRON FLUENCE (E>l MeV)

• PREDICTED CUMULATIVE FLUENCE 19 2 (10 n/cm )

VESSEL AZIMUTHAL LOCATIONS UNIT 1 DATE 0* 15*

Present 12/1/1985 1.32 0.81 Current License Expiration 7/19/2007 3.05 1.90 License Expiration, if License Amend- 10/5/2010 3.30 2.10 ment No. 107 approved Original design value 7/27/2012 (32 EFPY) 3.45 2.20 VESSEL AZIMUTHAL UNIT 2 DATE LOCATION-0 Present 12/1/1985 1.30 Current License Expiration 7/25/2008 3.12 Original design value 8/3/2012(32EFPY) 3.45 License Expiration, if License Amend- 3/8/2013 3.50 ment No. 107 approved

• Figures 3 and 4 were utilized to convert reactor operating period (EFPY) to the maximum, inside surface fluence values for the reactor vessels at the azimuth angles shown.

Page 13 of 23

TABLE 4 RT VALUES FOR REACTOR VESSEL BELTLINE MATERIALS

• PTS POINT BEACH NUCLEAR PLANT UNIT 1 A-9811** C-1423** SA-775/812** SA-847** SA-1101***

Intermediate Lower Axial Weld Axial Weld Circumferential Weld RTPTS( F) Values At Shell Plate Shell Plate Inter. Shell Lower Shell Inter. to Lower Shell Prcsent (12/1/85) 140.8 81.0 200.2 173.5 191.0 Current License Expiration 164.8 94.4 236.8 203.1 224.5 License Expiration, if License 167.4 95.8 241.7 207.1 228.1 Amendment NO. 107 Approved Original design EFPY value for 40 168.8 96.6 244.0 208.9 230.1 years of operation (32 EFPY)

UNIT 2 123V500** 122W195** SA-1484***

Intermediate Lower Circumferential Weld RTPTS(.F) Values At Shell Forging Shell Forging Inter. to Lower Shell Present(12/1/85) 146.3 116.0 238.0 Current License Expiration ,' 161.9 123.5 285.8 Original design EFPY value for 40 163.9 124.5 292.0 years of operation (32 EFPY)

License Expiration, if License 164.2 124.6 292.9 Amendment No. 107 approved (Predicted RT PTS values assume a cumulative (lifetime) capacity factor of 80%.

• • Applicable PTS screening' criterion - 270 F.

oc* Applicable PTS screening criterion - 300 F.

Page 14 of 23

REFERENCES

1. NUCLEAR REGULATORY COMMISSION, 10 CFR PART 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events",

July ?3, 1985.

2. Westinghouse Owner's Group (WOG), " Reactor Vessel Materials Data Base",

Revision 0, March 1985.

3. WCAP-10638, " Adjoint Flux Program for Point Beach Units 1 and 2",

December 1984.

4. Lukens Steel Company Test Certificate No. RM129655-NS, January 3, 1966 for Babcock and Wilcox Company.

5. WCAP-10736, " Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Radiation Surveillance Program", December 1984.

6. Lukens Steel Company Test Certificate No. RM61766-BB, January 20, 1966 for Babcock & Wilcox Company.

7. NUREG-0800, U.S. NRC Standard Review Plan, Branch Technical Position 5-2.

Revision 1 July 1981.

8. Bethlehem Steel Corporation Test Report No. 911, July 15, 1968 for Babcock

& Wilcox Company.

9. WCAP-7712 " Wisconsin Michigan Power Co. and the Wisconsin Electric Power Co. Point Beach Unit No. 2 Reactor Vessel Radiation Surveillance Program",

June 1971.

10. Bethlehem Steel Corporation Test Report No. 917, July 18, 1968 for Babcock

& Wilcox Company.

Page 15 of 23

O 1

9 Appendix A PBNP Reactor Vessel Weld Chemistry Data from WOG Material Data-Base T

Page 16 of 23

UNIT 1 SA-775/812: Axial Weld - Intrnnediate Shell SA-812: Weld Wire Type: Mn-Mo-Ni Weld Wire Heat No.: IP0815 Flux Type: Linde 80 Flux Lot Nos.. 8304, 8350*, 8544

• Denotes actual PBNP weld flux lot.

SELECT REPORT

=========================================================================================

ID WIRE WIRE FLUX FLUX WELDCHEN Cu Mi P Si HEAT TYPE TYPE LOT DATA SOURCE tel-MO-NI LINDE 80 8304 BW,We 0.250 0.480 0.019 0.440 0254 1P0815 0.130 0.570 0.022 0.420 0255 1P0815 MN-NO-NI LINDE 80 8544 BW WQ tel-MO-MI LINDE 80 8350 BW.WQ 0.120 0.520 0.017 0.400 0271 1P0815 meen 0.166667 0.523333 0.019333 0.420000 cts.dev. 0.072342 0.045092 0.002517 0.020000 as====================================================================================================

4 SA-775: Weld Wire Type: Mn-Mo-Ni Weld Wire Heat No.: IP0661 Flux Type: Linde 80 Flux Lot No.: 8304 SELECT REPORT

==.................................................... .............. ...............................

ID WIRE WIRE FLUX FLUX WELDCHEN Cu Ni P St HEAT TYPE TYPE LOT DATA SOURCE HN-NO-NI LINDE 80 8304 BW,WQ 0.190 0.630 0.024 0.460 0253 1PO661 mean 0.190000 0.630000 0.024000 0.460000 0.000000 0.000000 0.000000 0.000000 ctd.dev.

s===================================================================================================

Page 17 of 23

UNIT 1 SA-847: Axial Weld - Lower Shell Weld Wire Type: Mn-Mo-Ni L' eld Wire Heat No.: 61782 i Flux Typr: Linde 80 Flux Lot Nos.. 8350*. 8373. 8436. 8457. 8754

• Denotas actual PBNP weld flux lot.

SELECT REPORT .

====================================== -=================================================

ID WIRE WIRE FLUX FLUX WELDCHEN Cu N1 P 81 HEAT TYPE TYPE LOT DATA BOURCE 0306 61782 MN-HO-NI LINDE 80 6436 RGE.SC 0.220 0.500 0.006 0.410 0240 61782 MN-HO-NI LINDE 80 8350 BW.WQ 0.200 0.390 0.012 0.450 0268 61782 MN-HO-MI LINDE 80 8373 BW.WQ 0.220 0.490 0.016 0.460 0272 61782 MM-HO-NI LINDE 80 8457 BW.WQ 0.170 0.500 0.015 0.470 4

0282 61782 MN-HO-MI LINDE 80 8436 RGE.SC 0.230 0.580 0.012 0.590 0334 61782 MN-HO-NI LINDE 80 8436 BW.WQ 0.310 0.640 0.017 0.500 0343 61782 MN-HO-NI LINDE 80 8754 BAW-1799.WQ 0.290 0.470 0.017 0.420 0489 61782 MN-HO-MI LINDE 80 8436 BAW-1799.ESA O.270 0.490 0.014 0.580 0490 61782 MN-HO-NI LINDE 80 8436 BAW-1799.ESA 0.270 0.490 0.014 0.600 0491 61782 MM-HO-NI LINDE 80 8436 BAW-1799.ESA 0.240 0.490 0.014 0.570 0492 61782 MN-HO-HI LINDE 80 8436 BAW-1799.ESA 0.220 0.490 0.013 0.550 3

0493 61782 MN-HO-h! LINDE 80 8436 BAW-1799.ESA 0.230 0.4 90 0.013 0.540 0494 61782 MN-HO-MI LINDE 80 8436 BAW-1799.ESA 0.220 0.480 0.013 0.560

0495 61782 MM-HO-NI LINDE 80 8436 BAW-1799 ESA 0.210 0.480 0.013 0.550

! 0496 61782 MM-HO-NI LINDE 80 8436 BAW-1799 ESA 0.160 0.490 0.013 0.530 0497 61782 MM-HO-MI LINDE 80 8436 BAW-1799.ESA 0.180 0.4 90 0.013 0.540 0498 61782 MM-HO-NI LINDE 80 8436 BAW-1799 ESA 0.170 0.500 0.013 0.540 0499 61782 MM-HO-WI LINDE 80 8436 BAW-1799.ESA 0.150 0.490 0.013 0.540 0500 61782 MN-HO-NI LINDE 80 8436 BAW-1799.ESA 0.150 0.490 0.013 0.570 j 0501 61782 MM-HO-MI LINDE 80 8457 BAW-1799.ESA 0.270 0.580 0.013 0.530 i 0502 61782 HN-HO-NI LINDE 80 8457 BAW-1799.ESA 0.310 0.580 0.013 0.540 0503 61782 MN-HO-NI LINDE 80 8457 BAW-1799 ESA 0.310 0.580 0.013 0.5 90

} 0.510 0504 61782 MM-HO-MI LINDE 80 8457 BAW-1799 ESA 0.270 0.580 0.012 0505 61782 MM-HO-NI LINDE 80 8457 BAW-1799 ESA 0.270 0.580 0.013 0.580 0506 61782 iM-MO-NI LINDE 80 8457 BAW-1799.ESA 0.280 0.580 0.014 0.600 0507 61782 MM-HO-NI LINDE 60 8457 BAW-1799 ESA 0.290 0.580 0.012 0.520

! 0508 61782 MM-HO-NI LINDE 80 8457 BAW-1799.ESA 0.270 0.580 0.012 0.510 0509 61782 MM-HO-NI LINDE 80 8457 BAW-1799.ESA 0.280 0.580 0.012 0.500 0510 61782 MN-HO-WI LINDE 80 8457 BAW-1799 ESA 0.290 0.580 0.012 0.520 0511 61782 MN-MO-NI LINDE 80 8457 BAW-1799.ESA 0.280 0.590 0.012 0.510 0512 61782 MN-MO-NI LINDE 80 8457 BAW-1799.ESA 0.290 0.580 0.012 0.520 0513 61782 MM-HO-NI LINDE 80 8457 BAW-1799.ESA 0.240 0.580 0.011 0.470 0514 61782 MN-HO-MI LINDE 80 8457 BAW-1799.ESA 0.250 0.580 0.011 0.480 0515 61782 MM-HO-NI LINDE 80 8457 BAW-1799 ESA 0.220 0.590 0.011 0.460 0516 61782 MN-MO-NI LINDE 80 8457 BAW-1799 ESA 0.220 0.580 0.012 0.480 0517 617P2 MM-HO-NI LINDE 80 8457 BAW-1799.ESA 0.310 0.590 0.012 0.510 0518 61782 MN-HO-NI LINDE 80 8457 BAW-1799 ESA 0.320 0.580 0.011 0.480 0519 61782 MN-HO-NI LINDE 80 8457 BAW-1799.ESA O.310 0.590 0.011 0.470 0520 61782 MM-HO-NI LINDE 80 8457 BAW-1799.ESA 0.340 0.590 0.011 0.480 C521 61782 MM-HO-MI LINDE 80 8457 BAW-1799.ESA 0.340 0.590 0.010 0.470 0522 61782 MM-HO-MI LINDE 80 8457 BAW-1799.ESA 0.330 0.590 0.010 0.450 0523 61782 MN-MO-NI LINDE 80 8457 BAW-1799.ESA O.260 0.580 0.010 0.430 0524 61782 MM-HO-MI LINDE 80 8457 BAW-1799.ESA 0.220 0.600 0.011 0.460 0525 61782 MN-HO-NI LINDE 80 8457 BAW-1799.ESA 0.220 0.600 0.010 0.420 0526 61782 MN-HO-NI LINDE 80 8457 BAW-1799 ESA 0.230 0.600 0.010 0.440 0527 61782 MN-HO-NI LINDE 80 8457 BAW-1799 ESA 0.210 0.590 0.010 0.440 0528 61782 MM-HO-HI LINDE 80 8457 BAW-1799 ESA 0.230 0.600 0.009 0.380 0529 61782 MN-HO-NI LINDE 80 8457 BAW-1799 ESA 0.170 0.590 0.009 0.390

mean 0.248333 0.550208 0.012146 0.502292 Page 18 Of 23

. ctd.dev. 0.051996 0.053055 0.002000 0.057064

e UNIT 1 SA-Il01: Circumferential Weld - Intennediate to lower Shell Weld Wire Type: Mn-Mo-Ni W2ld Wire Heat No. 71249 Flux Type: Linde 80 Flux Lot No.: 8445 ,

SELECT REPORT

=======================================================================-=================

ID WIRE WIRE FLUX FLUX WELDCHEM Cu Ni P Si HEAT TYPE TYPE LOT DATA SOURCE 0223 71249 MN-MO-NI LINDE 80 8445 BW.WQ 0.210 0.570 0.021 0.520 0296 71249 MN-MO-NI LINDE 80 8445 FPL SC 0.310 0.570 0.011 0.660 0454 71249 MN-MO-NI LINDE 80 8445 BAW-1799,ESA 0.180 0.550 0.019 0.540 0455 71249 MN-MO-N1 LINDE 80 8445 BAW-1799,ESA 0.150 0.540 0.018 0.550 0456 71249 MN-MO-NI LINDE 80 8445 BAW-1799,ESA 0.180 0.550 0.019 0.540 0457 71249 MN-MO-NI LINDE 80 8446 BAW-1799,ESA 0.190 0.540 0.019 0.810 0458 71249 194-MO-NI LINDE 80 8445 BAW-1799,ESA 0.150 0.550 0.020 0.600 0459 71249 MN-MO-NI LINDE 80 8445 BAW-1799,ESA 0.170 0.540 0.019 0.620 0460 71249 kN-MO-HI LINDE 80 8445 BAW-1799,ESA 0.200 0.540 0.020 0.630 0461 71249 MN-MO-NI LINDE 80 8445 BAW-1799.ESA 0.200 0.540 0.019 0.630 0462 71249 MN-MO-NI LINDE 80 8445 BAW-1799,ESA 0.230 0.520 0.017 0.620 mean 0.195455 0.546364 0.018364 0.592727 ctd.dev. 0.045687 0,.014334 0.002656 0.046710 a=====================================================================================================

Page 19 of 23

e 1

UNIT 2 SA-1484: Circumferential Weld - Intennedtate to Lower Shell Weld Wire Type: Mn-Mo-Ni Weld Wire Heat No.. 72442 Flux Type: Linde 80 Flux Lot No.: 8579 SELECT REPORT

===============================================================t=========================

FLUX WELDCHEM Cu Ni P Si ID WIRE WIRE FLUX TYPE TYPE LOT DATA HEAT SOURCE LINDE 80 8579 BW.WQ 0.250 0.640 0.018 0.420 0220 72442 991-MO-MI 0.460 MN-MO-NI LINDE 80 8579 BAW-1799,ESA 0.240 0.570 0.016 0602 72442 BAW-1799,ESA 0.230 0.600 0.016 0.430 0603 72442 191-MO-HI LINDE 80 8579 LINDE 80 8579 BAW-1799,ESA 0.240 0.590 0.015 0.420 C604 72442 MN-MO-MI 0.430 0605 72442 tel-MO-NI LINDE 80 8579 BAW-1799,ESA 0.240 0.600 0.015 2

LINDE 80 8579 BAW-1799,ESA 0.220 0.600 0.018 0.470 C606 72442 MM-MO-NI 0.450 0607 72442 tei-MO-NI LINDE 80 8579 BAW-1799,ESA 0.270 0.600 0.016 8579 BAW-1799,ESA 0.260 0.600 0.017 0.480 C608 72442 MM-MO-HI LINDE 80 0.450 tel-MO-NI LINDE 80 8579 BAW-1799,ESA 0.270 0.690 0.016 0609 72442 BAW-1799,ESA 0.230 0.600 0.016 0.480 C610 72442 MN-MO-NI LINDE 80 8579 LINDE 80 8579 BAW-1799,ESA 0.280 0.590 0.016 0.490 0611 72442 191-MO-MI 0.500 MN-MO-HI LINDE 80 8579 BAW-1799,ESA 0.290 0.590 0.016 C612 72442 B4W-1799,ESA 0.300 0.600 0.017 0.510 0613 72442 tei-MO-MI LINDE 80 8579 LINDE 80 8579 BAW-1799,ESA 0.310 0.600 0.017 0.500 C615 72442 MN-MO-NI 0.490 LINDE 80 8579 BAW-1799,ESA 0.290 0.590 0.016 0614 72442 194-MO-MI BAW-1799,ESA 0.300 0.590 0.016 0.490 0616 72442 MN-MO-NI LINDE 80 8579 0.500 tel-MO-MI LINDE 80 8579 BAW-1799,ESA 0.260 0.590 0.017 0617 72442 BAW-1799,ESA 0.240 0.600 0.018 0.490 0618 72442 798-MO-NI LINDE 80 8579 LINDE 80 8579 BAW-1799,ESA 0.240 0.800 0.018 0.500 0619 72442 191-MO-MI mean 0.261053 0.596842 0.016421 0.471579 ctd.dev.

0.027466 0.012933 0.000902 0.029863

=========================================================n================================

1 i

Pagle 20 Of 23

Appendix B PBNP Emergency Operating Procedure Integrity Status Tree l

l I

l Page 21 of 23 l

)

ST-4 Major INTEGRITY Revision 0 07-01-85 i i

P P.1 PRESSURE -

TEMPERATURE NO gagg GO TO POINTS IN CSP P.1 BOTH COLD LEGS TO E RIGHT OF YES LIMIT A* 'T , GO TO TEMPERATURES NO .

{ IN BOTH yv ' j CSP P.2 COLD LEGS - O ss GREATERTHAN O 283'F YES TEMPERATURES NO

• SEE FIGURE 1 *

{ IN BOTH COLD LEGS GREATERTHAN 315'F YES TEMPERATURE DECREASEIN NO BOTH COLD CSF

+ LEGS LESS SAT THAN 100*F IN THE LAST YES 60 MINUTES g TO p_p TEMPERATURES NO IN BOTH I-- COLD LEGS GREATERTHAN 283'F YES RCSPRESSURE . ,

LESS THAN S / I'. l

, ,,  ; GO TO

CSP P.2 YES' _.-

i l

l CSF l SAT NO TEMPERATURES IN BOTH COLD LEGSGREATER THAN 354*F YES i CSF

( SAT i Page 22 of 23 1

I

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tEEsEEsammessamassamameenamensasseusessassa '

mEWEs a -

sWEs Ege ns sa men marew m ean s sp ese = =is es w ese s m ese 1 0 sWEEa m , IEEnEEmmseammasemmameansmanneasemeamemamman 200 250 350 400 450 l l 300 l 500 )

231 283 315 1 1

I l

)

COLD LEG TEMPERATURE ( F)

UMIT A T = 231 + .0204 P P T P T 0 231 1400 259 200 235 1600 263 400 239 1800 267 l 600 243 2000 271 l 800 247 2200 275 1000 251 2400 280 1200 255 2550 283 Page 23 of 23