ML20101M689
| ML20101M689 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 07/06/1992 |
| From: | Burski R ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-92-01, GL-92-1, W3F192-0094, W3F192-94, NUDOCS 9207090002 | |
| Download: ML20101M689 (10) | |
Text
. _ _ _ _ _ _ _ _ - _ - _ - - - _ _ - - _ _ _ - _ - -
4 Entsrpy Optrettann. Inc.
=== ENTERGY mes Wrvi L A 7(&C kt C' 4 739 C??4 R. F. Butski D%,
%.,- r... =,
s W3F192-0094 A4.05 QA July 6, 1992 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20535
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Generic Letter 92-01, Revision 1, Response Gentlemen:
In accordance with Generic Letter 92-01, Revision 1,
" Reactor Vessel Structural Integrity," attached is the Entergy Operations, Inc. Waterford 3 response.
Please contact me or Robert J.
Murillo should there be any questions regarding this response.
Very truly yours,
\\
wu -
RTR/RJM/ssf Actachment cc:
R.D. Martin, NRC Region IV D.L. Wigginton, NRC-NRR R.B. McGehee N.S. Reynolds NRC Resident Inspectors Office c[1001.0 sq 9207C70002 920706 Id lIl PDR._ADOCK 05000382 l
P PDR l
s
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the matter of
)
)
Entergy Operations, Incorporated
)
Docket No. 50-382 Waterford 3 Steam Electric Station )
AFFIDAVIT R.F.
Burski, being duly sworn, hereby deposes and says that he is Director Nuclear Safety - Waterford 3 of Entergy Operations, Incorporated; that he is duly authorized to sign ar.d file with the Nuclear Regulatory Commission the attached response to Revision 1 of Generic Letter 92-01; that he is familiar with the content thereof; and that the matters set forth therein are-true and correct to the best of his knowledge, information and belief.
/u R.F.
Bur' ski Director Nuclear Safety - Waterford 3 STATE OF LOUISIANA
)) ss PARISH OF ST. CHARLES
)
Subscribed and sworn to before me, a Notary Public in and.for the Parish and State above named this (31" day of du u f 1992.
-4
,N, o L la L%
Notary Public My Commission expires tv em ( ' 7 c v
Response to NRC Generio Letter 92-01 for Waterford 3 Question 1 Certain addressees are requested to provide the following information regarding Appendix H to 10 CFR Part 50:
1 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 of GL 92-01), are requested to describe actions taken or to be taken to ensure compliance with Appendix H to 10CFR 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 meet Appendix H to 10 CFR Part 50, addressees are' requested to indicate when they plan to request an exemption frsm Appendix-H to 10 CFR Part 50 under 10 CFR 50.60(b).
Resoonse to Ouestion 1 The Waterford 3 surveillance program was designed to meet the requirements of ASTM E-185-73.
Because the reactor vessel was fabricated to Section III of the ASME Boiler and Pressure Vessel Code 1971 Edition through the Summer 1971 Addenda, there were two exceptions:
(1) plate selection for the surveillance program was based on longitudinal data, and (2) the surveillance capsule assembly is attached to the vessel wall.
The NRC Staff cgcluded in its Supplemental Safety Evaluation Report (SSER) #1 that the surveillance program meets all the requirements of 10 CFR 50 Appendix H,
with the exception of Paragraph II.B.
The exception is that the limiting weld material, as determined by the Staff,.is not the weld metal included in the surveillance program. However, in SSER #1 the Staff also concluded that an exemption from Paragraph II.B was justified, as follows:
Waterford 3 would calculate the adjusted reference temptrature (ARTygy) for the weld-metal based on the greater of the measured shift in RT as determined by impact testing of the uoy surveillance weld metal and the predicted shift in RT as determined by Regulatory Guide 1.99 for the Staff-determkned uo limiting weld material.
Subsequent.to the Staff's SSER #1, Appendix H to 10 CFR 50 was revised, effective July 26, g83.
As a result of the changes, the Staff concluded in SSER #8 that the exemption was no longer required and'that.the Waterford.3 surveillance program complies 1
with the revised Appendix H requirements.
In addition to the above, it should be noted that Waterford 3 is committed to the 198'e revision of ASTM E-185, as documented in i
References (3) and (4).
Ouestion 2.a Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:
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 Paragraph 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 che end of their current license for the limiting beltline weld and the plate or forging and are requested to describe the actions taken pursuant to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.
Resnonse to Ouestion 2.a The Charpy upper shelf energy (USE), as predicted in accordance with Regulatory Gride 1.99 Revision 2, does not fall below 50 ft-lbs at the end of de ggn life.
In their SER related-to Generic Letter 88-11 response the NRC documented their conclusion that a conservative prediction of end-of-life Charpy USE is above the 50 ft-lb minimum required by Appendix G.
Therefore, no further action is required.
Ouestion 2.h 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 considerations given to the following material properties in their evaluations performed ' pursuant to 10CFR 50.61 and Paragraph III.A of 10 CFR Part 50, Appendix G:
(1)
The results from all Charpy and drop weight tests for all unirradiated beltline materials, and the unirradiated reference temperature for each beltline material, and the method for determining the unirradiated reference temperature from the Charpy and drop weight tests; (2)
The heat treatment received by all beltline and surveillance materials; (3)
The heat number for each beltline plate or forging and the heat number of wire and flux lot number used to i
fabricate each beltline weld; 2
i (4)
Tho heat numbsr for occh curvoillcncs plato or forging and heat number of wire and flux lot number 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 heat number of the wire used for determining the weld and chemical composition if different than Item (3).
Response to Ouestion 2.b The Waterford 3 reactor vessel meets the fracture toughness requirements in effect through the Summer 1971 Addenda to the ASME Code.
The longitudinal Charpy impact tests were satisfactorily performed, and Charpy V-notch results were delivered to the NRC with Reference (6).
Materials to perform additional testintj required by the Summer 1972 Addenda and 10 CFR 50 Appendix G were not available.
The available test data were evaluated according to BTF MTEB 5-2,
" Fracture Toughness Requirements", as recommenwed by the NRC-Staf f in a December 1974 meeting.
The results of baseline testing of the surveillance material demonstrated wide margin of conservatism in the evaluation.
This information was also included in the Reference (6) submittal to the NRC.
The Staff summarized in its SSER #1, its evaluation of the degree-of compliance for Waterford 3
with the fracture toughness requirements of 10 CFR 50 Appendix G.
The Staff concluded in relevant part that all requirements of Appendix G were met, with the exception of Paragraphs II.B.1, III.C.1, III.C.2 and IV.A.1, f
for which sufficient information was supplied to grant exemptions.
Additional information was also supplied to the NRC which was sufficient. to - demonstrate compliance with the requiremegy3s of Paragraphs III.B.3, IV.A.3, and IV.B of Appendir G, Part 50 Subsequent to SSER #1, Appendix G was revised, effective July 26, 1983.
With that revision, the Staff - concluded that, because fracture _ toughness properties for the reactor-coolant pressure boundary materials had been demonstrated equivalent to the requirements of Appendix G, the exemptions were longer required.
This conclusion was documented in their SSER #8 2.b.1 The drop weight test results, unirradiated reference temperature for each beltline _ material, and the method-for. determining the unirradiated reference temperature from the Charpy and drop weight test, are noted in Table 1.
2.b.2 The nominal heat treatment procean received by test and surveillance program materials is summarized below:
1.
1600
- F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Water Quenched. (Austenitizing) 2.
1225*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
(Tempering) 3.
1150'F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace Cooled to 600*F.
(Post-weld stress relief) 3
\\
L 2.b.3 Beltline material source is as follows:-
' Plate or Weld No.
Hgat No.
flux Lot No.
M-1003-1 56488-1 N/A M-1003-2 56512-1 N/A M-1003-3 56484-1 N/A M-1004-1
- 57326-1 N/A M-1004-2 57286-1 N/A M-1004-3 57359-1 _
N/A 101-124 A,B,C Type 80!A N/A electrode CE Lot No.
BOLA, HODA i'
101-142 A,B,C MIL-B-4~ wire 0091 flux-Heat No.
Lot No. 3536 83653 101-171 MIL B-4 wire 0091 flux Heat No. 88114. Lot No. 0145 4
2.b.4 Surveillance-material source is as follows:
Base metal ~is from plate M-1004 _(lower shell),
adjacent-to ASME Code Section III test material and.
water-quenchededge.gghickness at_ least-one-plate distant from. any Weld metal Land HAZ material -were. produced by welding together sections: of- - the selected base -
metal and another lower shell' plate of the vessel.
The HAZ test material was manufactured. from a section of the'rame bane netal1 plate u0ed'for baso i
metal test material.-
The suctions used for weld i
metal and HAZ test - materialiare adjacent - to the -
ASME Code Section III test materials at a distance.
of at~ 'least og plata= thickness from _ any water--
-quenched edge The welding-materials were the'same as that usede for weld No.~ 101-171, i.e., MIL:B-4_ weld wire Hett L
No. 88114.with Linde 0091-flux 1 Lot No. 0145.
2.b.5 Chemical composition of beltline. plates'and welds are in l
Table 1.
i 2.b.6
_Not applicable to Waterford 3.
Question 3.a' Addressees are requested-to provide the following information regarding commitments made to respor.d to GL 8G-11:
l l
l l
4 l
-. i
How the temperature effects of operating at an irradiation temperature (cold leg or recirculation suction temperature) below 525'T 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.
ResDonse to Ouestion 3.a The reactor has not been operated with a cold leg temperature (T-cold) less than 525'F for a significant period of time.
The Waterford 3 Technical Specifications (3.2.6) require that, in Mode 1 above 30% of rated thermal power, T-cold be maintained between 544
- F and 558'F.
Technical Specification 3.1.1.4 requires that the lowest operating loop temperature (T-cold) be greater than or equal to 520*F with the reactor critical.
The Action. Statement allows 15 minutes to restore T-cold within limits should T-cold fall below 520'F, or be in Mode 3 within the next 15 minutes.
Surveillance Requirement 4.1.1.4 requires that T-cold be verified greater than or equal to 520*F at least once per 30 minutes'when the reactor is critical and T-cold is less than 530'F.
A review of LER logs showed that no violation of T-cold limits for criticality has ever been reported.
Operating procedures OP-10-001 and OP-903-001 provide controls _to ensure that limits on T-cold are maintained.
With reactor power greater than 30%, procedures require that T-cold is maintained between 544*F and 558'F.
During startup, procedure Op-10-001 requires that the lowest T-cold be verified greater than or equal to 520*F every 15 minutes until criticality is achieved, and at least every 30 minutes with the reactor critical and RCS T-cold less than 530*F.
Between critical and 30%,
the Technical Specification Surveillance Logs for Modes 1-4 note limits for T-ccid; 544'F-558'F ir Mode 1 above 30%, and greater than 530*F in Modes 1 and 2.
The plant 'is thus controlled by the operators such that. core-critical operation with T-culd less than 525'F is an abnormal or transient condition, and therefore a temporary condition.. The reactor vessel has thus not been operated
_ ith a cold leg w
temperature less than 525'F for a significant-period of time.
Conservatism is built into the Adjusted Reference Temperature because the projected fluence of 3.68E19-in the FSAR is based en a high-leakage core, and the plant has operated since the first refueling with a low-leakage core.
The first surveillance capsule was_ withdrawn at the last refueling and is. presently. being evaluated.
The results of the capsule evaluation should provide clarification as to whether-there~have been any^ adverse effects from normal' plant operating practices.
Capsule evaluation results will be submitted to the NRC by November-30, 1992 in accordance with Waterford 3 letter W3F192-0052.
5
attestion 3.b Utilities are requested to provide information on how their surveillance results have been used in response to GL 88-11 and how the predicted amount of ci'.brittlement has been considered.
Resoonse to Ouestion 3.b Credible surveillance measurements are not yet available using Regulatory Position 2.1 of Regulatory Guide 1.99 Revision 2.
Testing of the first capsule withdrawal has not yet been completed.
Waterford 3 requested an extension for submittal of the summary report with Reference (9).
Ouestion 3.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 Ragulatory 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 parerial as predicted for December 16, 1991 and for its current license, t
Epoconse to Question 3.c See response to questien 3.b.
d 1
6
. _. _. ~. - -
~._
b..
't t
References-(1)
Supplemental Safety Evaluation Report No.- 1, by the FRU Staff, i
page 5-9, NUREG-0787, dated' October 1981.
(2) - Supplemental Safety Evaluation Report No. 8, by tne NRC Staff, i
page 5-1-NUREG-0787,. dated December 1984..
I (J)
LP&L letter to NRC, W3P87-1578,. dated July 31, 1987.
(4)
NRC Staf f SER Supporting Change to Bases Section of Technical Specifications, dated 8/20/87.
i (5)
NRC Staff SER Related to Generic Letter 88-11 Response, dated-l 11/27/89.
f j
(6)
LP&L letter to NRC, LPL8254, dated February-24, 1978.
(7)
Waterford 3 FSAR, Section 5.3.1.6.1.1 1
(8)
Waterford-3 FSAR, sect.*an 5.3.1.6.1.2' (9)
Entergy letter W3F192-0052, to Mr. Thomas Murley, _ NRC, dated 4
April 2, 1992.
4 i
l 1
i' l
t
!c e
i s
l l
i 7
. _. -... -. -. ~.. -....... -... -
TABLE'1 WATERFORD UNIT 3 REACTOR VESSEL MATERIALS Charpy Chemical Content (%)
Product Material Drop Weight Initial Specimen _
Nickel Cooper Phosphorous Sulfur.
Form
. Identification.
NDDT (*F)
% 7(*F)
Orientation-Plate M-1003-1
-30
-30 longitudinal 0.71 0.02 0.004 0.010
. Plate.
M-1003-2
-50
-50 longitudinal O.67 0.02 0.006 0.007 Plate M-1003-3
-50
-42 longitudinal 0.70 0.02 0.007
'O.009 Plate
.M-1004-1 -15 longitudinal 0.62 0.03 0.006 0.008 Plate
'M-1004-2
-20 22 longitudinal 0.58 0.03 0.005 0.005 Plate M-1004-3'
-50
-10 longitudinal 0.62 0.03 0.007 0.007 Wald
.101-1g.
-60
-60 N/A
-0.96 0.02 0.010 0.016-A,B,C Wald 1101-1g
-80
-80 N/A
<0.20 0.03 0.007 0.009 A,B,C
. Wald 101-171
-70
-70 N/A 0.16 0.05 0.008 0.008
-(a)
Intermediate shell course longitudinal seam weld
-(b)
Lower she31 course longitudinal' seam weld (c)
Intermediate -.. lower shell girth weld (d). Plate RT determined:using Branch Technical Position MTEB 5-2; g7 Weld RT,7 determined in accordance with ASME Code,Section III, NB-2300 8
-