ML20024B736

Affidavit of Bj Elliot Supporting NRC Motion for Summary Disposition of Carolina Environ Study Group Contention 18, Palmetto 44.Amount of Reactor Matl Degradation for Reactor Vessels Can Be Accurately Measured

ML20024B736
Person / Time
Site:	Catawba
Issue date:	07/01/1983
From:	Elliot B Office of Nuclear Reactor Regulation
To:
Shared Package
ML20024B704	List: ML20024B703 ML20024B708 ML20024B711 ML20024B717 ML20024B719 ML20024B723 ML20024B733 ML20024B736 ML20024B739 ML20024B743
References
NUDOCS 8307110293
Download: ML20024B736 (39)

Text

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UNITED STATES OF AMERICA NUCLEAR REGULATORY CONilSSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of DUKE POWER COMPANY, ET AL.

Docket Nos. 50-413 1

50-414 (Catawba Nuclear Station, h

Units 1and2)

)

AFFIDAVIT OF BARRY J. ELLIOT IN SUPPORT OF

SUMMARY

DISPOSITION OF CESG CONTENTION 18 (PALMETTO 44)

I, Barry J. Elliot, state under oath that:

1.

I am a Materials Engineer in the Materials Engineering Branch, t

Division' of Engineering, Office of Nuclear Reactor Regulation of the Nuclear Regalatory Commission.

I am responsible for reviewing the pressure-temperature operating limit curves and beltline material d

surveillance programs for all reactor vessels in nuclear power plants.

I havd personal knowledge of the matters set forth herein and believe

~them to be true and correct. A statement of my professional qualifica-tions is attached.

2.

As originally proffered, CESG Contention 18 (Palmetto 44) stated:

The license should not issue because reactor degradation in the form of a much more rapid increase in reference temperature than had been anticipated has occurred at a number of PWR's including the Applicant's Oconee Unit 1.

Until and unless the NRC and the industry can avoid reactor embrittlement, Catawba should not be pennitted to operate.

This contention, though initially rejected, was " admitted as clarified"

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in an Order of the Licensing Board, dated July 8,1982. The " clarified 8307110293 830708 PDR ADOCK 05000413 0

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contention" states:

The. Board appears to have misread this contention. Reactor materials did indeed comply with 10 CFR Part 50; App. G requirements when tested.

Intervenors' concern is with the unanticipated " rapid increase in reference temperature" which has been found in essentially all reactor vessels examined of which Oconee and Robinson are merely two nearby' examples. All of these reactor vessels were required to conform to essentially the same ASME codes. Perhaps it should be stated that the " reference temperature" is the nil ductility reference teiaperature below which the application of siiTficient stress produces a glass-like brittle fracture rather than a chewing-gum-like ductile stretching. Taffy provides a comon example of ductile stretching. Above the nil ductility temperature it " pulls" to an extraordinary degree itTfhout breaking. Below the nil ductility temperature it breaks like peanut brittle in a fracture mode.

Premature reactor embrittlement increases hazard because ECCS water during a LOCA can chill a reactor vessel below an elevated nil ductility temperature under conditions of appreciabTii stress.

It is only prudent in light of what is clearly a comon problem to anticipate and avoid these consequences at Catawba where no evidence exists that these reactors will behave any differently than other ASME Section III, Subsection NA Components. The Board should note the languageinASMECode,SubsectionNA1130p.3,(1971ed.):

The Rules.....do not cover deterioration which may occur in service as a result of radiation effects, corrosion, erosion or instability of the materials.

It is this now somewhat illuminated blindspot which Intervenors i

, seek to address.

The Staff has interpreted this contention as a claim, principally, that the NRC's projection of the amount of increase in reference temperature RTNDT, which results from neutron irradiation damage, is nonconservative, that the amount of reactor material degradation for the Catawba reactor l

l; vessels cannot be accurately measured, and, as a result, that a safety t

hazard therefore exists.

3.

The Staff's initial projection (as stated in its Safety Evaluation for the Catawba OL) of the amount of increase in RTNDT resulting from neutron irradiation damage for the Catawba reactor vessel i

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..x is based on the method presented in Regulatory Guide 1.99, Rev. 1

" Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."- This projection method was considered conservative because the trend curves in the Regulatory Guide fonn an upper bound for the data which was available at the time of its issuance.

4.

The conservative nature of the Regulatory Guide is demonstrated by comparing its projection method with the test results from the Oconee

?.eactor Vessel Surveillance Program and the projection method in Connission Report SECY-82-465, " Pressurized Thermal Shock." The Staff in its safety evaluation in a letter from E. L. Conner to H. B. Tucker dated March 11,1983(attached)comparedthechangeinRT for the NDT materials in the Oconee Reactor Vessel Surveillance Program to the projection method of Regulatory Guide 1.99. This report indicates that

- the actual increase in RT for the Oconee Capsule OC III-B and OC II-A NDT weld metals is 89'F and 104*F, respectively (Table 1 of Staff's Safety Evaluation in E. L. Conner letter).

For these capsule materials the Regulatory Guide 1.99 method would project an increase in RT of 170*F NDT

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and 226'F, respectively. Contrary to Intervenors' concern that there has been " unanticipated ' rapid increase in reference temperature'... "

at Oconee, the increase has been well below that which was predicted by the methods used for this purpose. With respect to Oconee, therefore, the increase in RTNDT projected by the Regulatory Guide is conservative, inasmuch as it is approximately twice the increase in RT of the t

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irradiated Oconee Reactor Vessel Surveillance material.

5.

In Commission Report Secy-82-465, the Staff statistically 4

evaluated the increase in RT resulting from irradiation damage from NDT 9-c--,-,,..,.n

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all.PWR reactor vessel surveillance materials. The source of this data is 25 operating PWR plant surveillance programs. The range of neutron 17 2

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fluence represented by the data is 8 x 10 n/cm to 8 x 10 n/cm (E1MeV). Since the end-of-life fluence for the Catawba reactor vessels will be within this range, the conclusions reached in the study are valid for Catawba for the life of the plant.

6.

The Comission evaluation resulted in the "Guthrie Formula" (Page E-6, Appendix E, SECY-82-465) which had a standard deviation of 24*F.

For the limiting Catawba Units 1 and 2 reactor vessel beltline materials, the "Guthrie Fomula" predicts at the end of life of the plant (40 years of operation) that the inside surface mean RT would increase NDT by 62*F and 61.5*F respectively. The Regulatory Guide 1.99 method predicts i

for the Catawba Units 1 and 2 materials that the increase in RT would NDT be 58'F and 94*F respectively. For Catawba Unit 1, the RT by the HDT Regulatory Guide prediction method is within 4*F of the mean RT of the HDT "Guthrie Formula" and for Catawba Unit 2, the RT by the Regulatory NDT Guide prediction method is within the 95% confidence interval (two standard deviations) of the "Guthrie Fomula." This shows that the Regulatory Guide and "Guthrie Fomula" prediction methods are consistent and are conservative with respect to the Oconee surveillance test data. Thus the staff considers these projection methods conservative for predicting the increase in RTNDT, which results from neutron irradiation damage.

7.

In addition to the prediction methods previously discussed, the Comission requires that all comercially operated reactor vessels comply with the requirements of Appendix H,10 CFR Part 50, " Reactor Vessel Material Surveillance Program Requirements." This program y

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requires that sanples from the limiting reactor vessel materials be placed inside reactor vessel surveillance capsules which are irradiated within the reactor vessel. According to the withdrawal schedule referenced in this Appendix, the capsules must be withdrawn and the materials must be tested to determine the amount of reactor vessel material embrittlement resulting from neutron irradiation damage. For the Catawba reactor vessels, for which it has been conservatively predicted that the end of service adjusted reference temperature will not exceed 200"F, the capsule withdrawal schedule is set forth in ASTM E 185, Section 7.6.2.

The results of these tests will be used to determine the actual increase in RT for NDT the Catawba reactor vessels. 'Thus the Staff considers that the combination of reference temperature increase prediction methods previously discussed and the Applicants' reactor vessel surveillance program will accurately detennine the amount of reactor material degradation for the Catawba reactor vessel materials.

8.

The Staff ensures safe operation of the reactor vessel during normal, anticipated upset and test conditions by requiring the vessel to be operated within the limits of Appendix G,10 CFR Part 50, " Fracture Toughness Requirements." According to this Appendix, the RT for the NDT limiting reactor vessel material is the basis for the reactor vessel l

operating limits. The Staff will compare the results of the l

surveillance program with the Staff projaction methods (i.e., "Guthrie,"

orReg. Guide 1.99)andwillusethehigherRT for :alculating ET l

operating limit curves. Since the Catawba reactor vessel materials will have their RT accurately detennined throughout the life of the plant, NOT the Staff considers that the reactor vessels can be safely operated

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4 during nonna1, anticipated upset and test conditions.

In addition, the operating curves for normal and anticipated upset conditions are calculated using a safety factor of two on the pressure which will account for variance in physical parameters, such as weld chemistry.

9.

The Staff ensures safe operation of the reactor vessel during faulted and emergency conditions by requiring the vessel RT to comply NDT with the screening criteria of Commission Report SECY-82-465, I

" Pressurized Thennal Shock." This report on page 6 states that "the risk from PTS events for reactor vessels with RT values less than the NDT proposed screening criterion (270*F for axial welds and 300'F for circumferential welds) is acceptable."

10. The projected end of life RT for the Catawba Units 1 and 2 NDT reactor vessels are identified in Table I.

The Staff considers that the 95% confidence RT represents the upper bound RT for the Catawba NDT NDT Units 1 and 2 reactor vessels, but that the RTET. projected by the

- Regulatory Guide 1.99 method and the mean RT Projected by the NDT "Guthrie Formula" represent the amount of embrittlement at end-of-life expected for the Catawba reactor vessels. The upper bound 95%

confidence RT for the Catawba Units 1 and 2 reactor vessels are 162'F NDT and 124.5*F, respectively. These values are well below the PTS screening criteria and indicate that the risk to the vessel during faulted and emergency conditions is acceptable.

11. The Staff therefore believes that the amount of reactor material degradation for the Catawba reactor vessels can be accurately measured and that the methods used to predict such degradation are conservative. Since Appendix G requires the vessel operating limits to

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7-be based upon the values so detemined for the limiting vessel material, there is reasonable assurance that the Catawba reactor vessels can and will be operated well within acceptable safety margins for material degradation.

Barry J.' Elliot Subsc i ed a worn to before me this day o 46 1983 neadu; Notary Public My comission expires: 7///((

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Table I Projected End-of-Life Reference Temperature, RT NDT RT by Reg. Guide Mean RT by "Guthrie Upper Bound 95%

NDT NDT 1.99 Method Formula" Confidence RTNOT b; "Guthrie Formula" Catawba Unit 1 110*F 114

'F 162

'F Catawba Unit 2 109'F 76.5'F 124.5'F i

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• s.% ~ o 3 s......r Cockets los. 50-259, 50-270 an:: 50-237 Mr. H. B. Tucker, Vice President Nuclear Production Cep. - e-t Duke Power Company P. O. Box 33189 422 South Church Stmet Charlotte, florth Carolins 152 4 Cear lir. Tucker:

The Comission has issued the enclosed Amendments ::cs.119,

119, and 116 to Licenses i;os. CPR-33, CPR-47 sr.d DPR-55 for the Oconee Nuclear Station, Units tics.1, 2 and 3.

"?:ese =end ene ::nris:

of changes to the Station's comon Technical Specifications (TSs) in nsponse to your request dated November 12, 1982, as sucolemented on February 24, 1983.

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These amendnents revise the TSs concerning the heatup, cooldown and inservice test limitations for the reactor coolant systems of each Oconee unit.

Copies of the Safety Evaluation and the Notice of Issuance are also enclosed.

Sincerely, ft h:- - e Eben L. Conner Project Manager Operating Reactors Branch #4 Division of Licensing

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Enclosures:

1.. Amendment No.119 to DPR-38 2.

Amendment No.119 to DPR-47 3.

Aaendnent No.116 to DPR-55 4.

Safety Evaluation 5.

Notice e'

cc w/enclosums:

See next page

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,;sa 7coer C :::a.,y cc s/anciosurels):

• Mr. William L. Porter Duke Power Comoany P. O. Box 33189 422 South Church Street Office of Intergovernmental Relations Charlotte. North Carolina 28242 116 West Jones Street Raleign, North Carolina 27603 Honorable James M. Phinney l

County Supervisor of Oconee County Walhalla, South Carolina 29621 Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission, Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 i

Regional Radiation Aepresentative EPA Region IV 345 Courtland Street, M.E.

Atlanta, Georgia 30308, William T. Orders Senior Resident Inspector U.S. Nuclear Megulatory Connission Route 2, Box 610 Seneca, South Carolina 29678 Mr. Robert B. Borsum Babcock & Wilcox l

Nuclear Power Generatton Division Suite 220, 7910 Woodmont Avenue

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l Bethesda, Maryland 20814 i

Manager, LIS NUS Corporation

.I 2536 Countryside Boulevard Clearwater. Florida 33515 J. Michael McGarry, III. Esq.

DeBevoise & Liberinan 1200 17th Street, N.W.

Washington, D. C.

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O'.'XE PC'JE?. CCM?lY CCCXE NO.

EC-250.

OC07;EE NUCLEAR STATIC'l, UNIT'NO.1 N'.ENDMENT TO FACILITY OPERATING LICENSE Amendment No.119 License No. DPR-38 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The apoliestien for amendment by Duke Power Cogany (the licensee) dated Nove.mer 12, 1982, as. supplemented February 24, 1983, coglies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), angl the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in confomity with the application, the pro-visions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of with the Comission)s regulations; thepublic,and(ii that such activities will be conducted in complidnce

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D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR part 51 of the Comissien's regulations and all applicable requirements have been satis-ffed.

i 2.

Accordingly, the license is amended by changes to the Technical Specifications as ' ndicated in the attachment to this license amendment and. paragraph 3.8 of F4c lity Operating License No. DPR-38 is hereby amended to re4d as follows:

3.BITechnicalSpecifications-

,e The Technical Specifications contained in Appendices A and B, as.,

revised through Amendment No.119., are, hemby incorcorated in tht.,

license. The licensee shall operate the facilits.in accordance !

y with the Technical Specifications.

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3.

This license amendment becomes effective on itarch 14, 1983.

l FOR THE ;;UCLEAR REGULATOR'.' 00;at::::0:1

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b6efabngkeact Brand #4 Di ision of Licensing

Attachment:

Changes tp the Technical Specipestions Date of Issuance: MAR 1 1 1983 em O

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DUKE POWER COMPANY COCKET NO.

50- 270 0007:22 ::L'CLIAR STATIO!!, L"!!T N0.2 A".E:: !'E::T TO FACILITY CPE?ATIMG LICE'!SE

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Amendment No.119 License No. DPR-47 1.

The Nuclear Regulatory Comission (the Comission) has fcund that:

A.

The application for emnd-ent by Duke Power Ccapany (the licensee}

dated November 12, 1982, as sucplemented February 24, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), angl the Comission's rules. and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the application, the pro-visions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that tha activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requ'irements have been satis-fled.

2.

Accordingly, the license is amended by changes to the Technical Specifications as "ndicated in the attachment to this license amendment and paragraph 3.8 of Fac'lity Operating License No. DPR-47 is hereby amended to read as follows:

3.8 Technical Specifications-

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The Technical Specifications contained in. Appendices A and 8, as revised through Amendment No.119. aret hereby incorporated 'in the'

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license. The licensee shall opera.te the facility in accordance '

with the Technical Specifications.

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3.

This license amendment becores effective on March 14, 1983.

FOR THE NUCLE.aD. REGULA. TORY COM11SSION

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Joh 'F. Stolz Chief 00 rating Reac, tors Branch #4 vision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: MAR 11 $83 O

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CUXE PCWER CCMP%iY CCCXI"' ?!O.

50 237 CCC?;EE ?!UC:. EAR STAT:Cf!, "?l!T ?:0.3

%1E?iCME'iT TO FACILITY OPERATI?iG LICE?lSE Amendment No.115 License No. DPR.

1.

The Nuclear Regulatory Ccmission (the Comission) has found that:

A.

Tne application for amendment by Duke Pcwer Company (tt.e licensee) dated Novenber 12, 1982, as supplemented February 24, 1983, conplies with the standards and req)uirements of the Atomic Energy Act o

-set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the application, the pro-visions of the Act, and the rules and regulations of the Cossr.ission; l

C.

Thereisreasonableassurance(1)thattheactivitiesauthorizedbythis amendnent can be conducted without endangering the health and safety of the public, and (ii).that such activities will be conducted in comp 1tince with the Comission's, regulations; l

D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and i

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the I

Comission's regulations and all applicable requirements have been satis-fled.

2.

Accordingly, the license is amended by chanDes to the Technical S;mifications as <ndicated in the attachment to.this license amendment and paragraph 3.8 of Fac lity Operating License No. DPR-55 is hertby amended to read as follows:

3.8 Technical Specifications-The Technical Specifications contained in Appendices A and B. as revised through Amendment No..116., are, hereby ingprcorated in the e'

license. The licensee shall operate the facility in.accordance '

with the Technical Specifications.

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~'"I 'i:anse amend:nent becomes effective on March 14, 1983.

FOR THE NUCLEAR REGULATORY COMMISSION n

a J n Stolz Chief Ipe iting Reac, tors Branch 4 i sion of Licensing 1

2-::ac., ment:

Oh:350: t the Technical E:a "ications Cate of Issuance: MAR 111983 e

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3.1.2

.trs22urica: ion. Restuo. and Cooldown Limitation 5: :ifi: ;:-

2.1.2.1 Tra r2::t:r :::1 ct pressure and the system heatup and cocidcun rates O.

n tce exception of the pressurizer) shall be limited

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Heatup:

• • aa: p rs:es and allowable combinations of pressure and tempera-ture shall be limited in accordance with Table 3.1-1 and Figure 3.1.2-1A *.*ni.

1 3.1.2-1B Unit 2 2.1.0- C Unit 3

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Cooldown rates and allowable combinations of pressure and tempera-ture sh.11 be limited in accordance with Table 3.1-2 and Figure 3.1.2-2A Unit 1 3.1.2-2B Unit 2 3.1.2-2C Unie. 3 3.1.2.2 Leak test,s required by Specification 4.3 and ASE Section XI shall be limited to the heatup and cooldown rates and allowable combina-tions of pressure and temperature provided in Tables 3.1-1, 3.1-2 and Figure 3.1.2-3A Unit 1 3.l.2-38 Unit 2 3.1,2-3C Unit 3 3.1.2.3 For thermal steady state system hydro tests required by ASE Section II the system may be pressurized to the limits set forth in Specification 2.2 and 3.1.2.2.

I 3.1.2.4 The secondary side of the steam generator shall not be pressurized i

above 237 psig if the temperature of the vessel shell.is below 110'F.

3.1.2.5 The pressurizer heatup and cooldown rates shall not exceed 100'F/hr.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 410*F.

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l 3 1-3 Amendnents Nos.119, 119, & 116 i

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1.1.2.e Prior to exceeding fifteen ' Unit IJ fifteen (Unit 2) fifteen (Unit JJ I

effective full power years of operation.

Figures J.i 2-1A (Unit 1), J.l.2-2A LUnit IJ 3.1.2-la (Unit 2), 3.1.2-2B (Unit 2) 3.1.2-IC (Unit 3), 3.1.2-2C (Unit 3) and 3.1.2-3A (Unit 1) 3.1.2-3B (Unit 2) l 3.1.2-3C (Unit 3) and Technical Specification 3.1.2.;, 3.*.2.2 and 3.1.2.3 shall be updsted for the next service period in acrordance with l0 07R 30, Appenaix G, seet;:n *. 3 nr.d y,g, 3.1.2.7 The updated proposed technical specification referred to in 3.1.2.6 shall be submitted for NRC review at least 00 days prior to tne end of the service period for Units 1, 2 and 3.

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3*I'3' Amendments tios. 119, 119. & 116

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tim:tstians of Il0*F and 227 psig are based on the highest estimated RT.D 4

v *Ja' and *.he preepersti:::1 :y: tem hyd::stati-test pressure of 101; psig.

The average metal temperature is assumed to be equal to or greater than the coolant temperaturee The limitatiens include sargias of 25 psi and 10*J for pcss:ble instrument error.

The spray temperature difference is imposed to maintain the' thermal stresses the pressurized spary line nozzle below the design limit.

at REFERENCES (1) Analysis of Capsule OCII-A from Duke Power Company Oconee Unit 2 Reactor Vessel Materials Surveillance Program, BAW-1699, December 1981.

(2) Analysis of Capsule OCIII-B from Duke Power Company Oconee Unit 3 Resetor Vessel Materisis Surveillance Program, SAW-1697, Detober 1981.

(3) Analysis of Capsule OCI-E from Duke Power Company Oconee Unit 1 Reactor Vessel Materials Surveillance Program, BAW-11.36, September,1977.

de' Amendnents Nos.119 119, & 116 3.1-5 i

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e TA2LE 3.1-2 OPERATIONAL GUIDANCE FOR PLANT CCOLD0Qi I.

RC Temperature Constraints RC Temperature (2)

!!azimum Cooldown Rate (2)

T > 280*F

$, 50'T in any \\ hour period 150*F < T < 280*F

< 25'T in any h hour period T < 150*F

< 10*F in any I hour period f3)

RCS depressurized

< 50'T in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period (1) RC temperature is cold leg temperature if one or more RC pumps are in operation or if on natural circulation cooldown; otherwise it is the LPI cooler outlet temperature.

(2).These rate limits must be applied to the change in temperature indication from cold les temperature to LPI cooler outlet tempera-ture per Note (1).

(3) When the RCS js depressurized such that all three of the following conditions exi,st:

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'RCS temperature < 200'F, l

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RCS pressure < 50 psig, c)

All RC Pumps off, the maximum cooldown rate shall be relaxed to f,50'T in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

II. RC Pump Constraints For Validity of Guidance RC Temperature Allowed Pump Combinations

> 270*F

.Any 270-200'T No more'than 1 pump per loop

< 200*F No more p an 1 pump Amendments.Nos. 119. 119. & 116 3,2 5b i

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.MUCLEAR REGULATCRv COMMIS3iCN a 9. j wasumcTON, 3. C. 20555 y..v,/

SAFETY EVALUATION SY THE OFFICE OF NbCLEAR REACTOR REGULATION SUPDORTING #iEND"ENT NO.119 TO RCILIT/ OPERATING LICENSE NO. DPR-33 AMEND!1ENT NO.119 TO FACILITY OPERATING LICENSE NO. OPR-47 A? E!!DMENT tio.116 TO FACILITI OPERATING ' LICENSE NO. DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, L::TS N05.1, 2 AND 3 OCCY,ETS N05, 50-2E3, 50-270 ANO 50-237 Introduction By letter dated November 12,1_:52,)*es revised on February 24, 1983, Duke Power Company (DPC or the licensee pmposed a chance to the Otonee Nuclear Station, Units 1, 2 and 3 Technical Specifications (TSs).

This 4

change is a revision to the reactor vejsel pressure-temperature limits.

Batkaround The licensee indicated that the bases for the proposed pressum-tegerature limits were the material properties data in Babcock & Wilcox (B8W) Reports BAW-1697 and BAW-1699. The curves for each Oconee reactor vessel are to be valid for 15 effective full power yean (EFPY).

The B8W Reports BAW-1697 and BAW-1699 contain the B&W analysis of reactor vessel material surveillance capsules DC III-B and OC II-A, nspectively.

l These capsules are part of the 88W Owners Group Integrated Surveillance Program. As a result, the capsules were irradiated in both the Oconee and Crystal River 3 reactor vessels.

Evaluation A comparison of the materials in the Oconee 1, 2 and 3 reactormssels and the DC III-B and DC II-A capsules indicates that the limiting weld mateHal in the Oconee 1, 2 and 3 mactor vessels is n6t contained in the DC III-B and OC II-A capsules. The limiting material in the Oconee 1, 2 and 3 mactor vessels is weld material SA 1430 WF 24, and WF 67, rerpectively. The weld materials in DC III-B and OC II-A are WF 233-18 and WF 209-1 A respectively. 'Although tSe weld materials in the vessel and the capsules are not. identical, theyMre prepared'hy the same manufactumr, using the same type of.wtre and flux and heat treated to an equivalent metallurgical condition. As a result, the fracture toughness data from capsules OC III.-B and OC II-A~isy be utilized for evaluating the proposed pressure-temperature limits.

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=C 2-h The cM nge in uccee sheH energy (USE) and reference temoerature resulting from neutron irradiaticn damage of the limiting materials in the DC III-3 and OC. II-A caosules are comoared in Table 1 to the values predicted by Regulatory Guide 1.99, Rev.1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials", a.nd the values predicted by

. BW Report BAW-1511P dated October 1980. This comparison indicates that

- the Regulaton Guide 1.99 method for predicting change in RT res ulting from neutron irradiation damage is conservative.

In additiekTthe method in Figure 13 of BW Report BAW-1511P for predicting the chan5e in weld material USE pmperties resulting from neutron irradiation damage is more a: curate than the method in Regulatory Guide 1.99.

Hence, we utilized Regulatory Guice 1.99 methodology for estimating the change in vessel material RT and Figure 3 in SW Report BAW-lE11P for estimating the change in reach, vessel material USE. We believe that Figure 3 in BW Report BAW-1511P is more accurate than Regulatory Guide 1.99 for estimating the change in USE resulting from irradiation damage for Oconee vessel and surveillance weld materials because Figure 3 in BW Report BAW-1511P was generated fmm reactor vessel surveillance weld materials similar to the Oconee vessel and surveillance weld materials, and the Regulatory Guide 1.99 curve was generated from reactor vessel weld materials utilized throughout the nuclear industry.

The USE for the Oconee beltline materials must exceed.50 ft-lbs at the 1/4 thickness location in order to meet the safety nargins required by Paragraph IV.A.2 of Appendix G,10 CFR Part 50.

Using Figure 3 in BW Report BAW-1511P, we estimate that the limiting materials in Oconee 1, 2 l

and 3 reactor vessel beltlines will have USE less than 50 ft-lbs at the 1/4 thickness location when their neutron f' gence (E>1MeV) exceeds 2

5 x 10 8n/cm', 4.8 x 1018n/cm2 and 7.5 x 10, u /cm, mspectively. Baseo 1

n on the naution f1 bence estimated by the licensee for each beltline material

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and the uncertainty in vessel dosimetry identified bf BW*, we

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concluc4 that the USE energy. at the 1/4 thickness Iocation for th'e Oconee beltline reactor vessel materials will excded'1io ft-lbs for the

_ period of time that the proposed pressure-temperature curves are applicable.

Using the method for predicting shift in RT in Regulatory Guide 1.99, Rev.1, the neutron fluence estimates of th5Dficensee, the unirradiated material properties in BW Reports BAW-1511P, October 1983, and BAW-10046P, March 1976, and the method of calculating prassure-temperature limits identified in Standard Review Plan Section 5.3.2, the proposed pressure-tenparature limit curves for Oconee Units 1, 2 and 3 meet the safety margins of Appendix G 10 CFR Part 50, and are acceptable for 15 EFPY.

• C. Whitmarsh, Draft BW Report to be Published.

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e Environmental Consideration We have determined that the amendments do not authorize a change in effluent types or total amounts ncr an increase in power level and will not result in any significant e'nvironmental impact. Having mada this determination, we have f.urther concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, or negative declaration and environ-mental impact appraisal need not be psepared in connection with the issuance of these amendments.

Conclusion

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We have concluded, based on the considerations discussed above, that:

i (1) because the amendnents do not involve a significant increase in the probability or consequences of an accident prc'iously evaluated, do not create the possibility of an accident of a type different from any evaluated previously, and do not involve a significant I

reduction in a margin of safety, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of these amendnents will not be inimical to the common defense and security or to the health and safety of the public.,

Dated:

M II W3 The following NRC personnel have contributed to this 3afety Evaluation:

L. Lois, 8..Elliot E. Conner.

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UNITED STATES riUCLEAR REGULATORY C0:2*33:0 !

00CXETS tiOS. 50-259, 30-27: aid 30-237 DUKE POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPEMTIf4G LICENSE 5 The U. S. Nuclear Regulatory Commission (the Comission) has issued kendments Nos. 119,119 and 116to Facility Operating Licenses Nos. DPR-38, DPR-47 and DPR-55, respectively, issued to Duke Power Company, which revised the Tech-nical Specifications (TSs) for operation of the Oconee Nuclear Station, Units Nos.1, 2 and 3, located in Oconee County, South Carolina. The amend-ments become effective pn March 14, 1983.'

These amendments revise the TSs concerning the heatup, cooldown and inservice test limitations for the reactor coolant systems of each Oconee unit.

The application for the amendnents complies with the standards and require-monts of the Atomic Energy Act of 1954, as amended (the Act), and the Consiission's rules and regulations. The, Commission has made appropriate findings as requiref by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments. Prior public notice of these amendnents was not' required since the amendnents do not involve a significant hazards consi-

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deration.

The Commission has determined that the issuance'of these amendmen result in any significant environmental iapact and that pursuant to 10 CFR'Section*

51.5(d)(4) an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of these amendments. -

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For further details with y pect to this action, see (1) the applicatien fdr amenoments dated Novemcar 12, 1382, as suppla:::antad.9brus y 2",

1983, (2) Amendments Nos.119,

119, and llG to Licenses Mcs he-22, e

DPR-47 and DPR-55, respectively, and (3) the Comission's relatad I-faty Evaluation. All of these items are available for.public inspection at the e

Comission's Public Document Rcom,1717 H Street, N. W., Washington, D. C.

and at the Oconee County Library, 501 West Southbroad Street, Walhalia, South Carolina 29691. A copy of tems (2) and (3) may be obtained coon request addressed to the U. S,Muclear Reg,ulatorf Comission, Washing::n, D. C. 20555, Attention:

Director, Division of Licensing.

Dated at Bethesda, Maryland, this lith day of March 1903.

FOR THE NUCLEAR REGULATORY CDMMISSION Jo

. Stolz,. Chief Op rating Reactors Branch'#4 vision of Licensing O

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Ti:le No.

Pate 2.3-1A b :cter Protective System Trip Setting Limits - Unit 1 2.3-11 2.3-1B Reactor Protective System Trip Setting Limits - Unit 2 2.3-12 2.3-1C Reactor Protective System Trip Setting Limits - Unit 3 2.3-13 3.1-1 Operational Guidance for Plant Heatup 3.1-Sa 3.1-2 Operational Guidance for Plant Cooldown 3.1-5b

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3.5-1-1 Instruments Operating Conditions 3/5-4

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3.5-1 Quadrant Power Tilt Limits 3.5-14 3.7-1 Operability Requirements for the Emergency Power Switching

3. 7-13 Logic Circuits 3.17-1 Fire Protection & Detection Systems 3.17-5

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4.1-1 Instrumens Surveillance Requirements 4.1-3 4.1-2 Mini== Equipment. Test Frequency 4.1-9 4.1-3 Minf=um Sampling Frequency 4.1-10 4.2-1 Oconee Nuclear Station Capsule Assembly Withdrawal Schedule 4.2-3 at Crystal River Unit No. 3 4.4-1 List of Penetrations with 10CTR50 Appendix J Test 4.4-6 Requirements 4.11-1 Oconee Environmental Radioactivity Monitoring Program 4.11-3 4.11-2' Offsite Radiological Monitoring Program 4.11-4 l

'4.11-3 Analytical Sensitivities 4.11-5

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4.17-1 Steam Generator Tube Inspection 4.17-6 6.1-1 Minimum Operating Shift Requirements with Fdel in Three 6.1-6 Reactor Vessels 6.6-1 Report of Radioactive Effluents 6.6-8 vi Amendments Nos. lig,

• 119, & 116

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4 ATTACHMEtiTS TO LICENSE NiEriGiENTS

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NIEliL'iEliT fiO.11970 CPR-33 Ai!EitCitENT NO. i10TO CPR 47 NiEliGiENT NO.11670 CPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Replace the following pages o' the Appendix "A" Technical Specifications f

with the attached pages. The ' revised, ages are identified by amendment numbers and contain vertical lines indicating the area of change.

Remove Paues Insert Paaes vi Vi 3.1 -3 3.1-3 3.1-3d 3.1-3a 3.1-5 3.1-5 3.1-Sa

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3.1-5b 3.1-6 3.1-6 3.1-6a 3.1 -6a 3.1-6b 3.1-6b 3.1-7 3.1-7 3.1-7a 3.1-7a

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3.1-7b 3.1-7b 3.1-7c 3.1-7c 3.1-7d 3.1-7d 7

e' 3.1-7e 3.1-7e.

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STATEMENT OF PROFESSIONAL QUALIFICATIONS BARRY J. ELLIOT U.S. NUCLEAR REGULATORY Com ISSION MATERIALS ENGINEERING BRANCH, DIVISION 0F ENGINEERING, NUC' EAR REACTOR REGULATION s

I am currently employed as'a Materials Engineer in the Materials Engineering Branch, Division of Engineering, Office of Nuclear Reactor

, Regulation.

I am responsible for review and evaluation of safety analysis reports which are related to the material engineering aspects of components in nuclear power plant systems, and for providing technical assistance to Offices of NRR, I&E and RES in related reactor safety matters.

I have been a member of the Materials Engineering Branch since March 31, 1980.

I was a full-time student at Rensselaer Polytechnic Institute, whert in 1968, I received.a Bachelors Degree in Materials Engineering.. I attended evening classes at Fairleigh Dickenson University, where in 1971, l

l-I received a Masters Degree in Business Administration.

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I was employed by Curtiss Wright Corporation from 1968, when I grad-uated from Rensselaer Polytechnic Institute, to 1980, when I was employed I

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by the NRC. From 1968 to 1971.I worked in the Materials Develbpment Laboratory of the Aeronautical Division where I performed failure analysis of reciprocating and gas turbine engines, and developed test apparatus to evaluate material reliablity. Frod1971to1980IworkedintheNuclear Division where I was responsible for developing and implementihg non-destructive examination test procedures and fusion weld procedures for inspection and fabrication of Navy Nuclear Pressure Vessels.

I am a member of the American Society.of Metals.

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