ML060040272

Technical Specifications Basis Change - Delete Tables for Reactor Vessel Toughness Data

ML060040272
Person / Time
Site:	Surry
Issue date:	08/23/1999
From:	Christian D Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
99-404
Download: ML060040272 (18)

Text

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8 %--.

VIRGINIA ELECTRIC AND POWER COMPANY RICHMoND, VIRGINIA 23261 August 23, 1999 U. S. Nuclear Regulatory Commission Serial No.99-404 Attention: Document Control Desk SPS-LIC/CGL, KLT RO Washington, D.C. 20555 Docket Nos.

50-280 50-281 License Nos.

DPR-32 DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS I AND 2 TECHNICAL SPECIFICATIONS BASIS CHANGE DELETE TABLES FOR REACTOR VESSEL TOUGHNESS DATA Virginia Electric and Power Company has approved a revision to the Basis for Technical Specification (TS) 3.11.B, "Heatup and Cooldown".

This change deletes reactor vessel toughness data that is duplicated in the UFSAR. In addition to deletion of the redundant information, a reference to the applicable UFSAR section is included in the TS Basis. The TS Basis revision is provided for your information.

A UFSAR change associated with this TS Basis revision is currently being processed and will be included in the next scheduled UFSAR update.

The UFSAR change updates the reactor vessel toughness data consistent with information previously provided to the NRC in a February 24, 1999 letter (Serial No.99-034).

The TS Basis revision has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee. It has been determined that this change does not involve an unreviewed safety question as defined in 10 CFR 50.59. A discussion of the TS Basis change, the mark-up of the TS Basis change, and the revised TS Basis are provided in Attachments 1, 2, and 3, respectively.

Should you have any questions or require additional information, please contact us.

Very truly yours, David A. Christian Vice President - Nuclear Operations

Attachments:

1. Discussion of Change

2. Mark-up of Technical Specification Basis

3. Revised Technical Specification Basis Commitments made in this letter:

1. A UFSAR change associated with the revised reactor vessel toughness data provided to the NRC in response to Generic Letter 92-01, Supplement 1, Revision 1, is currently being processed, and will be included in the next scheduled UFSAR update.

cc:

U.S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station Commissioner Department of Radiological Health Room 104A 1500 East Main Street Richmond, VA 23219

COMMONWEALTH OF VIRGINIA

)

)

COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by David A. Christian, who is Vice President - Nuclear Operations, of Virginia Electric and Power Company.

He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this3 day of f1 19 q9.

My Commission Expires: A 31 /J2Pf.Ua (SEAL)

ATTACHMENT I DISCUSSION OF CHANGE VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2

DISCUSSION OF CHANGE INTRODUCTION Virginia Electric and Power Company (Virginia Power) is revising the Surry Technical Specification (TS) 3.1.B Basis to delete Tables 3.1-1 and 3.1-2 and instead reference identical information that is contained in the Surry Updated Final Safety Analysis Report (UFSAR). The revised TS Basis is necessary to prevent potential consistency concerns between the Surry Technical Specifications and the UFSAR.

BACKGROUND TS 3.11.B Basis Tables 3.1-1 and 3.1-2 of the Surry Technical Specifications currently contain information regarding Reactor Vessel Toughness Data that is identical to the information included in Surry UFSAR Tables 4.1-14 and 4.1-15.

The data in these tables require update to reflect the latest information that was submitted to the NRC in a letter dated February 24,1999. Because the information is identical in both documents, and because it is referenced only by the Basis for Technical Specification (TS) 3.1.B, the Basis requires revision to delete the two tables and instead reference the updated tables in the UFSAR.

Licensing/Design Basis The Virginia Power response to Generic Letter 92-01, Revision 1, is documented in BAW-2222, 'Reactor Vessel Working Group Response to Closure Letters to NRC Generic Letter 92-01, Revision I," dated June 1994. A resultant TS change request was submitted to the NRC in a letter dated June 8, 1995 (Serial No.95-197) that included modifications to TS Tables 3.1-1 and 3.1-2 and was approved by the NRC in a letter dated December 28, 1995 (TS Amendments 207).

Since the publication of BAW-2222 in June 1994, Framatome Technologies has published BAW-2313, Revision 1, 'B&WOG Reactor Vessel Working Group Reactor Vessel Materials and Surveillance Data Information, Volumes I and 2," dated December, 1998. After an evaluation of data contained in BAW-2313, Revision 1, Virginia Power submitted a supplemental letter to the NRC dated February 24, 1999, to incorporate the most recent reactor vessel materials information into our docketed response to Generic Letter 92-01, Revision 1, Supplement 1. This submittal included more recent values of "mean for the heat" chemical composition for Surry Units 1 and 2 reactor vessel beltline materials, which will be reflected in UFSAR Table 4.1-14 and Table 4.1-15. Therefore, deletion of Tables 3.1-1 and 3.1-2 from the TS 3.1.B Basis will eliminate inconsistencies between the two documents and simplify any future revisions to the data.

Page 1 of 3

SPECIFIC CHANGES As noted above, the revised TS Basis for TS 3.1.B deletes two tables that are redundant to existing UFSAR tables. The specific revisions are as follows:

In the Basis of TS 3.1.B, under item 6), reference to Tables 3.1-1 and 3.1-2 is changed to, "UFSAR Section 4.1."

At the end of the Basis for TS 3.11.B, a reference table is added to reflect UFSAR Section 4.1.

TS Basis 3.1 Tables 3.1-1 and 3.1-2 are deleted.

SAFETY SIGNIFICANCE The revision to the Surry Technical Specification 3.11.B Basis deletes two tables that are redundant to existing UFSAR tables. Reference to the TS Basis tables that are being deleted is changed to reference the UFSAR instead.

Deletion of the redundant information eliminates the possibility that the two sets of information could be updated inconsistently. The information will continue to be controlled and maintained up to date in the UFSAR, and any future changes to the UFSAR tables will be reviewed in accordance with the requirements of 10 CFR 50.59. Consequently, no change to the licensing basis results from this revision. There is no safety significance associated with this revised TS Basis, since it does not involve any plant modifications or changes in system operation, is consistent with the Surry licensing and design bases, and does not affect the existing accident analyses, which remain bounding.

Furthermore, no change to the TS limiting conditions for operation is required.

Page 2 of 3

REFERENCES

1.

BAW-2222, "Reactor Vessel Working Group Response to Closure Letters to NRC Generic Letter 92-01, Revision 1," dated June 1994.

2.

Letter from J. P. O'Hanlon to USNRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Request for Exemption - ASME Code Case N-514, Proposed Technical Specifications

Change, Revised Pressure/Temperature Limits and LTOPS Setpoint," dated June 8, 1995 (Serial No.95-197).

3.

Letter from B. C. Buckley (USNRC) to J. P. O'Hanlon, "Surry Units 1 and 2 -

Issuance of Amendments Re: Surry Units 1 and 2 Reactor Vessel Heatup and Cooldown Curves," dated December 28, 1995 (Serial No.96-020).

4.

BAW-2313, Revision 1, "B&WOG Reactor Vessel Working Group Reactor Vessel Materials and Surveillance Data Information, Volumes 1 and 2," dated December 1998.

5.

Letter from L. N. Hartz to USNRC, "Virginia Electric and Power Company, Surry Power Station Units I and 2, Supplement to Virginia Power Response to NRC Request for Additional Information (RAI) on Generic Letter 92-01 Revision 1, Supplement 1," dated February 24, 1999 (Serial No.99-034).

Page 3 of 3

ATTACHMENT 2 MARK-UP OF TECHNICAL SPECIFICATION BASIS VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS I AND 2

TS 3.1-9 Heatup and cooldown limit curves are calculated using the most limiting value of the niltductility reference temperature. RTNDT, at the end of 28.8 Effective Full Power Years (EFPY) and 29.4 EFPY for Units I and 2, respectively. The most limiting value of RTNDT (228.41F) occurs at the 1/4-T, 0° azimuthal location in the Unit 1 intermediate-to-lower shell circumferential weld. The limiting RTNDT at the 114-T location in the core region is greater than the RTNDT Of the limiting unirradiated material.

This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

4 fAP, 'SecAkonr.

The reactor vessel materials have been tested to de RTNDT; the results of thesetst; are presented in JZ "RespIns to Glesure 6ftters to NRC Gonorc Letto 02f01.

R1vicsln 1.'

-dated juno, 104 94and ar roproducd in Tables 3. 1 and 3. 1 X2.

Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2 'Effects of Residual Elements on Predicted Radiation Damnage to Reactor Vessel Materials.' The heatup and cooldown limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RTNDT at the end of 28.8 EFPY and 29.4 EFPY for Units 1 and 2, respectively (as well as adjustments for location of the pressure sensing instrument).

Surveillance capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule Is shown in the UFSAR. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the calculated ARTNDT for the equivalent capsule radiation exposure, or when the service period exceeds 28.8 EFPY or 29.4 EFPY for Units I and 2, respectively, prior to a scheduled refueling outage.

Amendment Nos. 207 and 207 A.

TS 3.1-12

-t2&695 r ~ce-s (J) (4PSARI -Sec-6'on -44.1 s

THIS PAGE HAS BEEN iTENTIONAY DEL D

I Amendment Nos. A-nd -

TABLE 3.1-1 UNIT I REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)(d)

MA1EML ibw head dome HEAT OR CODE NO.

C4315-2 FV-1894 FV-1870 Head flangeN Vessel flange Inlet nozule Inlet nozzle MATERIAL S9PEC. NO-A53313 Cl. 1 A508 Cl. 2 A508 Cf. 2 A508 CI. 2 A508 Cl. 2 A508 Cl. 2 Cu LvD

.14

.13

.10 PA

.59

.64

.65

.87

.84 TNDT 0

10(a) 60(a)

RTNJrT 0

10 10 NMWD(b)

UPPER SHELF ENERGY 75 Inlel nozzle Outlet nozzle Outlet nozzle Outlet nozzle X,

Upper shel 3fCL Intermediate shell f

Intermediate shell C..

o Lower shell Lower nom head rg 9-4787 9-4762 9-4788 9-4825

.83

.84

.85

.74 en(a) 60(a) 60(a) 60(a) 40 10 60 60 60 60 60 40 10 0

20 74 64 68 64 85 72 68 83 1 I5(c) 94 103(0) 83 8B FIQi 122V109 A508 CI. 2 C432 A533B Cl. 1 4326-2 A533B Cl. 1 C4415-1 A533B Cl. 1 C4415-2 A5338 Cl. I 123T338 A50B Cl. 2 C4315-3 A5338 Cl. I 8T1554 & Lhds 80 tiux

.11

.11

.11

.11

.14

.18

.55

.50

.50

.69

.59

.63 20 0

50 0

0(a) 0 Bottom dorm Inter. & lower sthel vertical weld seam L1, L3.

UL4 50 0

-5 r%

t_

In I'

toI

TABLE 3.1-1 (Contrinued)

UNIT 1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)(d)

MATERIAL

'-t'sbe vertical HEAT OR MATERIAL CODE No. '

SPEG NM 2991.44 & Litl 80 flux Cu

.M

.35

.68 TNDT L

NMWD(b)

UPPER SHELF ENERGY r

70/EMA(e)

-7 weld sea Inter. to lower shell girth seam Upper shell to Inter.

shell girth seam 72445 & Linde 80 fhlx

.21 0(a)

-5 T7(8)/EMA(e)

I I-4-.4-89 flux

.10 0(a) 0 EMA(e) 0.m 0

a, m

5 0

0

-4 NOTES:

(a) Estimated per NRC standard review plan. NURE

00. Section MTEB i

(b) Nommal to major working dredbn - estimat per NRC standard review plan, NUR Section MTEB 5-2 (c) Actual values (d) Reactor Vessel Fabricator C Test Reports (e) The approved equiva imargins analysis In the Topical Reports BAW-2192PA and BAW-2178PA demonstrates 10 CFR 50, Appe 0.

with the requirements

TABLE 3.1-2 UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)

NMWD/b)

UPPER SHE HEAT OR MATERIAL Cu NI TNDT RTNDT ENERGY CODE=A SPEG

ieNO, MI MI ca E

COsur dome C4361-2 A5338 Cl. I

• 15

.52 20 7

He fta ZV-6475 A508 Cl. 2

.11

.60

,10(a) 129 Vessel flange ZV-3476 A508 Cl. 2

.10

.64 6 5

.65 129 In9le norz 1

A508 Cl. 2

.87 60 66 Inlet norzzb 9-151104

\\

A508 Cl. 2 0

()60 73 Inl nozzl 9-5205 08 Cl. 2

/

.86 60$a) 60 66 Outlet nozzle 9-4825 A50

.2

.85 60(a) 60 74 t

Outlel nozzle 9-5086 A508 C

.86 60(s) 60 79 O utlet no~zzl 9-5086 A

. 2

\\.87 60(s) 60 73 Uppershell 123V303

/A08 Cl. 2

.0 \\

73 30 30 104 ltemddlte shellC436 3

A533B Cl. 1

.12

-10

-10 84 Internediaeshe f3ll 72 AU330 Cl. I

.11

.59 5

-20 83 v

Lower she/l C4208-2 A533B Cf.

o

.15 855 530 39 94 Lower(shel C4339-1 A5331 C(.

0

.11

.54

-t0 105(C) 1 Bo(1D50%

123T321 A508 Cl. 2

-892

.55 1

10 EMA~d)

°CD dome C4361-3 A5338 Cl. I

.15

.52

-20

-15 80

\\

Irntemoi~edashenl 72445tA Linde 80 fx

.21

.59 77(a)/EMA(d ver~tial wellsearns Lot 8579.

o L3 (i00%).LAI(01350Y.)*d L4 (D51/6t) 8T17B2 & Lince80 flux8597

.20

.55

-S EMA{d)

TABLE 3.1-2 (Continued)

UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)

UPPER SHELF HEAT OR MATERIAL Cu N

TNDT RTN ENERGY vertical K

San L2 (I

)

T1782 & Lhde 80 flx 8597

.20

.55

-5 EMA(d)

Seam Li (1001%

8T1782 & Lhie 80 lux 8597

.20

.55

.5 EMA(d)

Seam L2 (0T7%)

811782 & Lhde 80 flx 862

.20

-5 EMA(d)

Inter. to lower 0nG o

to FkxLW320

.19

.56 0(a) 0 90(c)E~MA(d)

t stNll Mt seam Upper shetlto Inter.

4275 &

F89/

.35

.10 0(a) 0 EMA(d) shiel ginl seam NOTES:

(a) Estimated per NRC standard review plan REG-0800. Section UTEB 5-2 (b) Normal to major workhi dredion stted per NRC standard review plan. NUR E

ion MTEB 5-2 1

(c) Actual value based on tance ests normal to the major workg dkirction 0

?

(d) The approved equ nt margins analysis In the Topical Reports BAW-2192PA and BAW-2178PA demo es compfiance with the requiremenis of 0t CFR 50, di G.\\

0D

/

1S.-.

CLI

/-\\0

/',

ATTACHMENT 3 REVISED TECHNICAL SPECIFICATION BASIS VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS I AND 2

TS 3.1-9 Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 28.8 Effective Full Power Years (EFPY) and 29.4 EFPY for Units I and 2, respectively. The most limiting value of RTNDT (228.40F) occurs at the 1/4-T, 00 azimuthal location in the Unit I intermediate-to-lower shell circumferential weld. The limiting RTNDT at the 1/4-T location in the core region is greater than the RTNDT of the limiting unirradiated material. This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNDT; the results are presented in UFSAR Section 4.1. Reactor operation and resultant fast neutron (E greater than I MEV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2 "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RTNDT at the end of 28.8 EFPY and 29.4 EFPY for Units I and 2, respectively (as well as adjustments for location of the pressure sensing instrument).

Surveillance capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the UFSAR. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the calculated ARTNDT for the equivalent capsule radiation exposure, or when the service period exceeds 28.8 EFPY or 29.4 EFPY for Units I and 2, respectively, prior to a scheduled refueling outage.

Amendment Nos.

TS 3.1-12 References (1) UFSAR, Section 4.1, Design Bases Amendment Nos.

TS 3.1-26 Pages TS 3.1-26 through TS 3.1-29 have been deleted.

Amendment Nos.