ML20101N929

Forwards Response to Rev 1 to Generic Ltr 92-01 Re Reactor Vessel Structural Integrity.Table Listing Load Factor Vs Days Encl

ML20101N929
Person / Time
Site:	Brunswick
Issue date:	07/06/1992
From:	Starkey R CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, NLS-92-180, NUDOCS 9207100204
Download: ML20101N929 (14)

Text

__ _ _ _ _ _ _ _ _ _

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CD&L Carourta Power & Ught Company P O Ha,1!al

• Hsie@ N C 2 M.02 lJUL 001992 SEhlAL: NLS 92180 n a sunston 10 CFR 50.54(f)

No.[UO.T[e.<n United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50 325 & 50 324/ LICENSE NOS DPR 71 & DPR 62 nESPONSE TO GENERIC LETTER 92 01, REVISION 1 NEACTOR VESSD STRUCTURAL INTEGRITY Gentlemen:

The purpose of this letter is to provide Carolina Power & Light Company's response to NRC Generic Letter 92 01, ' Reactor Vessel Structural Integrity", Revision 1. The Generic Letter, which was issued March 0,1992, requests information needed to assess licensee compliance with te raments and commitments regarding reactor vesselintegrity in v%W of concerns raised in the staff's review of reactor vesselIntegrity for the Yankee Nucle.r Power Station. Enclosure 1 provides C?&L's responses to the NRC requests for the Brunswick Steam Electric Plant, Units 1 and 2.

Please refer any questions regarding this submittal to Mr. W. R. Murray at (919) 540-4661.

Yours very truly, bthbis R. B. Starkey, Jr.

WRM/wrm (gl92-01.wpf)

Enclosure R. B. Starkey, Jr., having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, contractors, and agents of Carolina Power & Light Company.

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ENCLOSURE 1 BPUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKET NOS. 50-325 & 50 324 OPERATING LICENSE NOS. DPR 71 & DPR 62 RESPONSE TO GENERIC LETTER 92 01, REVISION 1 REACTOR VESSEL STRUCTURAL INTEGRITY NRC REQUEST:

Certain addressees are requested to provide the following information regarding Appendix H to CFR Part 50:

Addressees who do not have a surveillance program meeting ASTM E185 73, 79, or -82 and who do not have an integrated surveillance program approved by the NRC (see Enclosure 2),

are requested to describe actions taken or to be taken to ensure compliance with Appendix H to 10 CFR Part 50. Addressees who plan to revise the surveillance program to meet Appendix H to 10 CFR Part 50 are requested to indicate when the revised program will be submitted to the NRC staff for review. If the surveillance program is not to be revised to eneet Appendix H to 10 CFR Part 50, addressees are requested to indicate when they plan to request an exemption frnm Appendix H to 10 CFR Part 50 ur. der 10 CFR 50.60(b).

CP&L RESPONSE:

The Brunswick Steam Electric Plant, Unit 1 and Unit 2 vessels were purchased to the 1965 Edition

o. aston lit of the ASME Boiler and Pressure Vessel Code and Addenda through Summer 1967; 11 ee, the E185 66 ASTM standard edition applies to the Surveillance Program conducted prior to ).. capsula removal. At that time, the ASTM standard was not mandatory, and General Electric used Section 111 of the ASME Code alone.

Unirradiated Charpies in a lot of three were tested only at 10'F and only two tensiles were pulled compared to the ASTM receirement of 15 Charpies and three tensiles. The surveillance capsules contain the ASTM recommended number of specimens except for the case of weld tensiles. The pair of specimens supplied is short of the recommended three (Reference 1).

Three surveillance capsules were supplied for each unit, and the withdrawal schedule incorporated in the Technical Specifications satisfied ASTM E185-66 except for removing one at the near end of design life neutron fluence (Reference 2).

Revision of the Technical Specifications was obtained in order to achieve more useful fast neutron exposures for the first capsule withdrawal and permit a subsequent decision as to the most appropriate time to remove 'he second and third capsules. The low capacity factor, low chemistry factors, and low fast neutron flux prompted the changes that were approved in Amendments 140 and 172 to the Brunswick Plant Technical Specifications (Reference 3). Carolina Power & Light Company believes that the revised schedule currently contained in the Brunswick Plant Technical Specifications meets the intent of ASTM E185 82. Although ASTM E185-82 requires the use of three capsules that are withdrawn at specific times, the Brunswick Plant uses three capsules that will be tested using withdrawal times that are varied in order to improve the information to be obtained.

El 1

_ ~ _ _ -.

NRC REQUEST:

Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:

a.

Addressees of plants for which the Charpy upper shelf energy is predicted to be less than 50 foot pounds at the end of their licenses using the guidance in Paragraphs C.1.2 or C.2.2 in Regulatory Guide 1.99, Revision 2, are requested to provide to the NRC the Charpy upper shelf energy predicted for December 16,1991, and for the end of their current license for the limiting beltline wold and the plate or forging and are requested to descJN the actions taken pursuant to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR wn 50.

b.

Addressees whose reactor vessels were constructed to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the following material proporties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph Ill.A of 10 CFR Part 50, Appendix G:

(1) the results from all Charpy and drop weight tests for all unirradiated beltline materials, the unitradiated reference temperature for ea.:h beltline material, and the method of determining the unitradiated reference tempeidture from the Charpy and drop weight test; (2) the heat treatment received by all beltline and surveillance materials; (3) the heat number for each beltline plate (.T forging and the heat number of wire and flux lot number used to fabricate each beltline weld; (4) the hs:t number for each surveillance plate or forging and the heat number of wire and flux lo; iumber used to fabricate the surveillance weld; (5) the chemical composition, in particular the weight in percent of copper, nickel, phosphorous, and sulfur for each beltline and surveillance material; and (6) the hcat number of the wire used for determining the weld metal chemical composition if different than item (3) above, CP&L RESPONSE:

ltem a:

Section lll of the ASME Boiler and Pressure Vessel Code as of 1967 did not require that the upper shelf energy be determined. When the NRC made inquiry into shelf properties on February 13, 1989, CP&L responded that the Branch Technical Position MTEB 5 2, paragraph B.1.2 would be followed !. Reference 5); to wit, conservative estimates will be made using results from the first surveillance capsule. The first surveillance capsule is scheduled to be removed from Unit 1 during the next refueling outage (i.e., the Reload 8 outage). The NRC will be informed of those results as required by Appendix H to 10 CFR Part 50, item b:

'n response to a May 23,1977 NRC request (R*'"ence 6), Carolina Power & Light Company provided the bulk of this information (Referen-Much of the transmittal was added to the Final Srfety Analysis Report as NEDO 24157 and 2

,1 (Reference 1).

El 2 l

Tho submittal did not include the chemical analyses of surveillance welds. These will be obtained as the first capsule is removed from each unit.

NRC REQUEST:

Addressees are requested to provide the following information regarding commitments made to respond to GL 8811:

a.

How the embrittlement effects of operating at an irradiation temperature (cold leg or recirculation suction temperature) below 625'F were considered. In particular licensees are requested to describe consideration given to determining the effect of lower irradiation temperature on the reference temperature and on the Charpy upper shelf energy, b.

How their surveillance results on the predicted amount of embrittlement were considered.

c. If a measured increase in reference temperature exceeds the mean plus two standard deviations predicted by Regulatory Guide 1.99, Revision 2, or if a measured decrease in Charpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C.1.2 in Regulatory Guide 1.99, Revision 2, the licensee is requested to report the information and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December 10,1991, and for the end of its current license.

CP&L RESPONSE:

110m3:

The Brunswick Plant, Units 1 and 2 operate below 525'F during ascension to full power and at times when abnormal feedwater flow conditions exist; for example, when a feedwater heater is out of service or the curbine trips without reactor scram. For purposes of this response, graphs and figures are used to provide an accounting for the time and neutron exposure to temperature less than 525'F.

Figures 1 and 2 give the number of days the two plants were expused to fractional thermal power over plant lifetime; load factor correlates well with average daily power. Figures 3 and 4 give the same data for the first year of operation, and Figures 5 and 6 give the same data for 1991. Note the significant reduction of time spent at less than 10 percent load factor to less than 90 percent load factor as unit operation became more stable with time.

The relationship between thermal power and recirculation inlet temperature is given in Figure 7. In a study of operations since 1985, only two significant perioos of abnormal operation were found.

Uo;t 2 operated at 517' to 523'F from November '3,1987 to November 20,1987 because of feedwater system problems. Unit 1 operated at 520* to 524'F from January 7,1992 to March 23,1992; this was followed by 5 days at 512' to 514'F. Again, feedwater system problems were responsible; two feedwater heaters were out of service for the latter period.

As power is reduced, neutron exposure of the vessel declines proportionately. This is quantified in Figures 8 and 9.

The effect of this operation below 525 F on the fracture toughness properties of the reactor vessel steels and welds used for the Brunswick Plant, Units 1 and 2 will be evaluated by the specimens in El 3

4 the first surveillance capsule to be removed from the units. The first capsule is scheduled to be removed from Unit i during the Reload 8 outage (currently expected to begin in the second quarter of 1993). Removal of the first Unit 2 capsule will follow by approximately two effective full power years of operation. The Unit I capsule 11 expected to have accumulated a neutron dose of approximately 4 x 10" n/cm' (E > 1 Mev). (A predictive calculational approach to quantify the effect on fracture toughness is unavailable due to the shortage of operational or research data on the subject.)

As shown in this siction, operation in future years should occur minimally at a temperature below 525'F.

Real.h:

Carolina Power & Light Company will remove the first capsule from the Brunswick Plant, Unit I during the upcoming Reload 8 outage (currently expected to begin in the second quarter of 1993).

Currently, embrittlement is predicted by use of methods in Regulatory Guide 1.99, Revision 2.

j Regulatory Guide 1.99, Revision 2 was used in all responses to NRC requests and for heatup-cooldown calculations.

Item c:

Carolina Power & Light Company does m ;. %ve surve:ilance results from the Brunswick Plant, Unit 1 or Unit 2.

REFERENCES:

1. Brunswick Steam Electric Plant Updated Final Safety Analysis Report for Units 1 and 2, Appendix 5.3.B, though Amendment 10.

2. Brunswick Steam Electric Plant Technical Specifications, Ame:,oment 117 (Unit Il and Amendment 147 (Unit 2).

3. Brunswick Steam Electric Plant Technical Specifications, Amendment 140 (Unit 1) and Amendment 172 (Unit 2).

4. Letter from E. D. Sylvester (USNRC) to E. E. Utley (CP&L) dated April 4,1988.

5. Letter from L. l Loflin (CP&L) to USNRC Document Control Desk dated March 30,1989, Serial Number: NLS-89 059, " Supplement to Request for License Amendment, Pressure / Temperature Limits."

6. Letter from A. Schwencer (USNRC) to J. A. Jones (CP&L) dated May 23,1977.

7. Letter from E, E. Utley (CP&L) to T. A. Ippolito (USNRC) dated January 8,1979, Serial Number: GD-79 060, ' Reactor Vessel Material Surveillance Program Data."

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BRUNSWICK UNIT 1 LOAD FACTOR vs DAYS LIFETIME i

DAYS l

7,000--

6,500

=

6,000 5.82s

=

i 5,500

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L 5,000

=

4,500

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2,084 2,000 1,500 1,000 476 500

- 23e-3c3 3es 0

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~ ^ ^ M-S/D <10 <20 <30 <40 <50 <60 <70 <80 <90 290 TOTAL LOAD FACTOR (%)

FIGURE 1

s BRUNSWICK UNIT 2 LOAD FACTOR vs DAYS l

LIFETIME l

7,000 6,500 e.282

=

6,000

=-

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=

5,000 :-

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l 1,973 2,000 1,500 1,000 454 111 94 73 88 126 197 215'

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FIGURE 2

BRUNSWICK UNIT 1

_OAD =AC-~OR vs JAYS YEAR: 1976 DAYS f

184 175 [-

150 150 125 100 75 50 25

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4 2

o o

o o

o o

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FIGURE 3

BRUNSWICK UNIT 2 LOAD =AC~~OR vs DAYS YEAR: 1975 DAYS 300

~

275 275 ?

250 1 225 r-4 200 1

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r 175 ;-

150 ?

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LOAD FACTOR (%)

FIGURE 4

BRUNSWICK UNIT 1

LOAD FACTOR vs JAYS YEAR: 1991 i

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365 1

350 '

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230' 225 [

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FIGURE 5

BRUNSWICK UNIT 2

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365 350 7 325 t 300 275 r 250 '

225.-

1g4 200 175 155 150 125 100 75 50 25

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FIGURE 6

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es Relation Between Reactor Power and Recirculation inlet Temperature I

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Equivalent Neutron Exposure Com aared to Time at Fractional Power (Brunswic< L ni: -)

l Figure 8 60 :'

A % Time at Power Equival tf4 uron 2

59.4 50 40

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Equivalent Neutron Exposure Com aared to Time at Fractional Power i Brunswic< Uni ~: 2) s 62 60 -

A % Time at Power Equival t t4 u ron 50 =

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