ML030410038
| ML030410038 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/31/2003 |
| From: | Jury K Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-03-010 GE-NE-0000-0003-5526-02a, Rev 0 | |
| Download: ML030410038 (172) | |
Text
ATTACHMENT I Affidavit STATE OF ILLINOIS COUNTY OF DUPAGE IN THE MATTER OF:
EXELON GENERATION COMPANY (EGC), LLC LASALLE COUNTY STATION - UNIT I and UNIT 2
SUBJECT:
)
)
)
)
)
Docket Numbers 50-373 and 50-374 Request for Amendment to Technical Specification Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits" AFFIDAVIT I affirm that the content of this transmittal is true and correct to the best of my knowledge, information, and belief.
T. W. Simp"*n Manager-Licensing Mid-West Regional Operating Group Subscribed and sworn to before me, a Notary Public in and for the State above named, this ___'_-
day of
-. 0-
,A.,,.2003 "o yPu lc TIYMOH A. YM]
OYMMEISON MWm 12/0Wl A
(ýJ-
ATTACHMENT 2 Evaluation of Proposed Changes Page I of 7
1.0 INTRODUCTION
2.0 DESCRIPTION
OF PROPOSED AMENDMENT
3.0 BACKGROUND
4.0 REGULATORY REQUIREMENTS & GUIDANCE
5.0 TECHNICAL ANALYSIS
6.0 REGULATORY ANALYSIS
7.0 NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC)
8.0 ENVIRONMENTAL CONSIDERATION
9.0 PRECEDENT
10.0 REFERENCES
ATTACHMENT 2 Evaluation of Proposed Changes Page 2 of 7
1.0 INTRODUCTION
In accordance with 10 CFR 50.90, Exelon Generation Company (EGC), LLC, hereby requests the following amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-1 1 and NPF-18. Specifically, the proposed changes will revise TS 3.4.11, "RCS Pressure and Temperature (P/T) Limits," to incorporate revised P/T curves. The revised P/T curves are based on calculations performed in accordance with General Electric (GE) Topical Report NEDC-32983P, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation." The NEDC-32983P methodology is consistent with the guidance contained in Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated March 2001.
2.0 DESCRIPTION
OF PROPOSED AMENDMENT The proposed changes are as follows.
TS Surveillance Requirement (SR) 3.4.11.1 and SR 3.4.11.2 are modified to reference the revised P/T curves to be used to perform the surveillances. The revised surveillances with the modified wording in bolded text are as follows.
- a.
RCS pressure and RCS temperature are within the applicable limits specified in Figures 3.4.11-1, 3.4.11-2, and 3.4.11-3 for Unit 1 up to 20 EFPY, and Figures 3.4.11-4, 3.4.11-5, and 3.4.11-6 for Unit 2 up to 20 EFPY;
- b.
RCS heatup and cooldown rates are _ 100OF in any I hour period; and
- c.
RCS temperature change during system leakage and hydrostatic testing is < 20°F in any one hour period when the RCS pressure and RCS temperature are not within the limits of Figure 3.4.11-2 for Unit I up to 20 EFPY, and Figure 3.4.11-5 for Unit 2 up to 20 EFPY.
SR 3.4.11.2 Verify RCS pressure and RCS temperature are within the criticality limits specified in Figure 3.4.11-3 for Unit I up to 20 EFPY, and Figure 3.4.11-6 for Unit 2 up to 20 EFPY.
Replace the current TS Figures 3.4.11-1 through Figure 3.4.11-6 with revised TS Figures 3.4.11-1 through Figure 3.4.11-6. The revised TS Figures are applicable to 20 EFPY.
LaSalle County Station (LSCS) intends to submit updated TS curves applicable to 32 EFPY and receive NRC approval prior to either Unit I or 2 exceeding the enclosed 20
ATTACHMENT 2 Evaluation of Proposed Changes Page 3 of 7 EFPY P/T curves. LSCS Unit I is predicted to reach 20 EFPY in 2010 and Unit 2 in 2011.
3.0 BACKGROUND
In Reference 1, EGC requested changes to the P/T limits in the TS for LSCS Units 1 and
- 2. During teleconferences between members of the NRC and EGC in support of the review of the proposed TS changes, the NRC stated that the neutron fluence calculations used to develop the revised P/T limits were not consistent with the guidance contained in Draft Regulatory Guide DG-1053, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated September 1999.
In March 2001, DG-1053 was approved as RG 1.190. The NEDC-32983P methodology is consistent with the guidance contained in RG 1.190. In a letter dated September 14, 2001, the NRC approved NEDC-32983P for use by licensees.
4.0 REGULATORY REQUIREMENTS & GUIDANCE 10 CFR 50.36(c)(2)(ii)(B), "Criterion 2," requires that a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier, must be included in a licensee's TS.
5.0 TECHNICAL ANALYSIS
5.1 Design Bases All components of the reactor coolant system (RCS) are designed to withstand effects of cyclic loads due to system pressure and temperature changes.
These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. TS 3.4.11 limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
TS 3.4.11 contains P/T limit curves for heatup, cooldown, inservice leak testing, hydrostatic testing, criticality and also limits the maximum rate of change of reactor coolant temperature. Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and
ATTACHMENT 2 Evaluation of Proposed Changes Page 4 of 7 temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.
Attachments 5 and 6 to this submittal contain the GE proprietary and non proprietary reports, and Attachment 7 contains a GE non-proprietary letter.
The reports describe the analyses scope, assumptions, methodology and results. The P/T curves contained in Attachment 7 were used to revise TS Figures 3.4.11-1, 3.4.11-2, 3.4.11-4 and 3.4.11-5. The P/T curves contained in were used to revise TS Figures 3.4.11-3 and 3.4.11-6.
The TS P/T curves included in this submittal have been developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. The revised TS Figures are applicable to 20 EFPY.
LaSalle County Station (LSCS) intends to submit updated TS curves applicable to 32 EFPY and receive NRC approval prior to either Unit 1 or 2 exceeding the enclosed 20 EFPY P/T curves. LSCS Unit I is predicted to reach 20 EFPY in 2010 and Unit 2 in 2011.
5.2 Risk Information This submittal is not based on risk informed decision making.
6.0 REGULATORY ANALYSIS
The P/T limits are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the reactor coolant pressure boundary, a condition that is unanalyzed. Therefore, the P/T limits must be included in LSCS TS in accordance with 10 CFR 50.36(c)(2)(ii).
7.0 NO SIGNIFICANT HAZARDS CONSIDERATION Exelon Generation Company, LLC (EGC) has evaluated the proposed changes to the Technical Specifications (TS) for LaSalle County Station (LSCS), Unit I and Unit 2, and has determined that the proposed changes do not involve a significant hazards consideration and is providing the following information to support a finding of no significant hazards consideration.
ATTACHMENT 2 Evaluation of Proposed Changes Page 5 of 7 Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed changes request for LSCS, Units I and 2, that the pressure and temperature (PIT) limit curves in TS 3.4.11, "RCS Pressure and Temperature (P/T) Limits," and Surveillance Reqirement (SR) 3.4.11.1 and SR 3.4.11.2 be revised. The revised curves were developed using the methodology of General Electric (GE) Topical Report NEDC-32983P, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation."NEDC-32983P methodology has been previously approved by the NRC for use by licensees.
The P/T limits are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the reactor coolant pressure boundary, a condition that is unanalyzed. Thus, the proposed changes do not have any affect on the probability of an accident previously evaluated.
The P/T curves are used as operational limits during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. The PIT curves provide assurance that station operation is consistent with previously evaluated accidents. Thus, the radiological consequences of any accident previously evaluated are not increased.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed changes do not change the response of plant equipment to transient conditions. The proposed changes do not introduce any new equipment, modes of system operation or failure mechanisms.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.
Does the change involve a significant reduction in a margin of safety?
Response: No
ATTACHMENT 2 Evaluation of Proposed Changes Page 6 of 7 The proposed changes adopt P/T curves that have been developed using the methodology of GE Topical Report NEDC-32983P. The NEDC-32983P methodology is consistent with the guidance contained in Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated March 2001. In a letter dated September 14, 2001, the NRC approved NEDC-32983P for use by licensees.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based upon the above, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
8.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
9.0 PRECEDENT The proposed changes for LSCS, Units I and 2, provide revised P/T curves that have been developed using the methodology of GE Topical Report NEDC-32983P. The NEDC-32983P methodology is consistent with the guidance contained in RG 1.190 and in a letter dated September 14, 2001, the NRC approved NEDC-32983P for use by licensees.
10.0 REFERENCES
(1)
Letter from R.M. Krich (CoinEd) to U. S. NRC, "Application for Amendment to Appendix A, Technical Specifications, Section 3/4.4.6, "Pressure Temperature Limits, Reactor Coolant System," and Request for Exemption from 10 CFR 50.60, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation," dated February 29, 2000
ATTACHMENT 2 Evaluation of Proposed Changes Page 7 of 7 (2)
Letter from Charles G. Pardee (CornEd) to U. S. NRC, "Supplement to Application for Amendment to Appendix A, Technical Specifications, Section 3/4.4.6, "Pressure Temperature Limits, Reactor Coolant System," and Request for Exemption from 10 CFR 50.60, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation," dated August 18, 2000 (3)
Letter from Donna M. Skay (U. S. NRC) to 0. D. Kingsley (CornEd), "LaSalle County Station, Units I and 2 - Issuance of Amendments (TAC Nos. MA8403 and MA8404)," dated November 8, 2000 (4)
Letter from Charles G. Pardee (CornEd) to U. S. NRC, "Revised General Electric Nuclear Energy Reports, "Pressure Temperature Curves for ComEd LaSalle Unit 1" and "Pressure Temperature Curves for CoinEd LaSalle Unit 2" dated June 26, 2000 (5)
Letter from T. Simpkin (Exelon Generation Company, LLC) to USNRC, "Justification for the Continued Use of Technical Specifications, Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits," dated July 19, 2002 (6)
Letter from K. R. Jury (EGC) to U.S. NRC, "Request for Amendment to Technical Specifications Justification for the Continued Use of Technical Specifications, Section 3.4.11, 'RCS Pressure and Temperature (P/T) Limits,'"
dated October 21, 2002 (7)
Letter from P. R. Simpson (EGC) to U.S. NRC, "Revised Request for Amendment to Technical Specifications Justification for the Continued Use of Technical Specifications, Section 3.4.11, 'RCS Pressure and Temperature (P/T)
Limits,"' dated November 8, 2002
ATTACHMENT 3 MARKUP OF PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES Revised TS Pages
RCS P/T Limits 3.4.11 SURVEILLANCE REOUIREMENTS SURVEILLANCE SR 3.4.11.1
- ------------------- NOTE -------------------
Only required to be performed during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.
Veri fy:
- a.
RCS pressure and RCS temperature are within the applicable limits specified in Figures 3.4.11-1, 3.4.11-2, 3.4.11-3 for Unit I,*and Figures 3.4.11-4. 3.4.11-5, and 3.4.11-6 for Unit 2,
- b.
RCS heatup and cooldown rates are
< 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period; and
- c.
RCS temperature change during system leakage and hydrostatic testing is
< 20 0F in any one hour period when the RCS pressure and RCS temperature ar not within the limits of Figure 3.4.11-2 for Unit 1 and Figure 3.4.11-5 for Unit 2.
SR 3.4.11.2 Verify RCS pressure and RCS temperature are within the criticality limits specified in Figure 3.4.11-3 for Unit 1 and Figure 3.4.11-6 for Unit 2 J.
FREQUENCY 30 minutes Once within 15 minutes prior to control rod withdrawal for the purpose of achieving criticality (continued)
Amendment No.
147/133 SURVILLACE RQUIRMENT LaSalle I and 2 3.4.11-3
RCS P/T Limits 3.4.11
/
11400 1300e 1200 1100 9000 900 sow Ei 3~
00 400 U.
3w 20C 10C 0
50 75 100 125 150 175 200 225 250 5300 IMUM REACTOR VESSEL METAL TEMPERAT
('F)
P-T Curves for Hydrostatic or Leak Te ting Figure 3.4.11-1 (Page 1 of 1)
Unit 1 Minimum Reactor Vessel Metal Temperature vs.
Reactor Vessel Pressure (Valid until December 15, 2004) and 2 3.4.11-6 Amendment No.
156/142
/rO
\\HA
\\
- FLANG REGIO
\\-2'F F?.
INITIAL RTndt VALUES ARE
-30F FOR BELTUNE.
40OF FOR UPPER VESSEL, AND 47°F FOR BOTTOM HEAD I BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 32 102 HEATUPICOOLDOWN RATE OF COOLANT
< 20"FIHR UPPER VESSEL AND BELTUNE UMITS
....... BOTTOM HEAD CURVE I
a:
A C
4 U
0 LaSalle 1 4q I
RCS P/T Limits 3.4.11 0
25 so 75 100 15 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
INITIAL RTndt VALUES ARE
-30°F FOR BELTLINE, 40"F FOR UPPER VESSEL, AND 47"F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (OF) 20 114 HEATUP/COOLDOWN RATE OF COOLANT 5 20OFIHR UPPER VESSEL AND BELTLINE LIMITS BOTTOM HEAD CURVE m
Figure 3.4.11-1 (Page 1 of 1)
Unit 1 P-T Curves for Hydrostatic or Leak Testing up to 20 EFPY LaSalle I and 2 o~ oao IU U) 8O0 03 IU o700 SI I.
- s Ixl L0 mu A1.
0 Amendment No.
3.4.11-6
Limits 3.4.11 BELTLINE CURVES IAf r%
I"QT~fl AQ QLIAWRIs EFPY SHIFT (°F)
SW 32 102 w
/
co--
HEATUP/COOLDOWN
>u RATE OF COOLANT o
700
< 10--F/HR W WO n II UPPER VESSEL wI I
AND BELTUNE 3O//
LIUMITS
BOTTOM HEAD BOTTOM CURVE 200 HEAD i/
0 25 75 100 125 150 175 200 225 250275 MINIM REACTOR VESSEL METAL TEMPERATURE
(*F)
P-T Curve for Heatup by Non-Nuclear Means, Cooldo n Following A uclear Shutdown and Low Power Physics Tes ing Figure 3.4.11-2 (Page 1 of 1)
Unit 1 Minimum Reactor Vessel Metal Temperature vs.
eactor Vessel Pressure (Valid until December 15, 2004 0/
Amendment No.
156/142 LaSalle 1 and 2 3.4.11-7
RCS P/T Limits 3.4.11 INITIAL RTndt VALUES ARE
-30"F FOR BELTLINE, 40°F FOR UPPER VESSEL, AND 47°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 20 114 HEATUP/COOLDOWN RATE OF COOLANT
< 100°F/HR UPPER VESSEL AND BELTLINE LIMITS
- ------ BOTTOM HEAD CURVE 0
2 s0 7S 100 1t2 150 115 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
Figure 3.4.11-2 (Page 1 of 1)
Unit 1 P-T Curves for Heatup by Non-Nuclear Means, Cooldown Following a Nuclear Shutdown and Low Power Physics Testing up to 20 EFPY LaSalle 1 and 2
[woo a
Wow 0
n, tJ I
4W z_
a "ccm CJ 0
Amendment No.
3.4.11-7
RCS P/T Limits 3.4.11 "a
5 50 75 100 125 150 175 200 225 250 75 500 NIMUM REACTOR VESSEL METAL TEMPERATUR (F)
P-T Curves for Operation with a Core Crit other than Low Power Physics Testin Figure 3.4.11-3 (Page 1 of 1)
Unit 1 Minimum Reactor Vessel Metal Temperature 14\\00 1300 1200 1100 1000 U) goo zV i
0-700 00
@5 00 I
400 BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 32 102 HEATUPICOOLDOWN RATE OF COOLANT S0FIH JR
-BELTLINE AND NON-BELTLINE LIMITS
ýical Reactor Vessel Pressure (Valid until December 15, 2004)
Amendment No. 156/142 LaSalle 1 and 2 INITIAL RTndt VALUES ARE
-30°F FOR BELTLINE, 404F FOR UPPER VESSEL, AND 479F FOR BOTTOM HEAD 300 200 100 int VS.\\
I 3.4.11-8
RCS P/T Limits 3.4.11 I
INITIAL RTndt VALUES ARE
-30°F FOR BELTLINE, 40'F FOR UPPER
- VESSEL, AND 47°F FOR BOTTOM HEAD BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 20 114 HEATUP/COOLDOWN RATE OF COOLANT
< 100°F/HR BELTLINEAND NON-BELTLINE LIMITS 25 O
50 75 100 125 150 175 20 22 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE
('F)
Figure 3.4.11-3 (Page 1 of 1)
Unit 1 P-T Curves for Operation with a Core Critical other than Low Power Physics Testing up to 20 EFPY LaSalle 1 and 2 1100 Iw OJ a) co
.I U,
mu a.
0 Amendment No.
3.4.11-8
RCS P/T Limits 3.4.11 1400 1300 1200 1100 25 0/
7 100 125 150 175 200 225 250 5
MINIVIM REACTOR VESSEL METAL TEMPERATURE P-T Curves for Hydrostatic or Leak Tes ng Figure 3.4.11-4 (Page 1 of 1)
/
Unit 2 Minimum Reactor Vessel Metal Temperature vs.
Reactor Vessel Pressure (Valid until December 15, 2004)
Amendment No. 156/142 LaSalle 1 and 2 BELTLINE CURVES ADJUSTED AS SHOWN EFPY SHIFT ('F) 32 23 HEATUPICOOLDOWN RATE OF COOLANT
.20FIHR
-UPPER VESSEL AND BELTLINE LIMITS BOTTOM HEAD CURVE 1000
,=,
O. 300 0
70
$Do 5
500 us S400 a)
A.
I 3.4.11-9
RCS P/T Limits 3.4.11 1=0 1300 1200 1100 IO0 7WO 6wO SIX 4OO SIX 2wO l0 0
25 so 15 10 125 150 115 2W MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 3.4.11-4 (Page 1 of 1)
Unit 2 P-T Curves for Hydrostatic or Leak Testing up to 20 EFPY LaSalle 1 and 2 INITIAL RTndt VALUES ARE 52°F FOR BELTLINE.
40°F FOR UPPER VESSEL.
AND 49°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (=F) 20 25 HEATUP/COOLDOWN RATE OF COOLANT
< 20F/HR UPPER VESSEL AND BELTLINE LIMITS
........ BOTTOM HEAD CURVE
_z
- a.
40 I
w I
a.
Amendment No.
3.4.11-9
1400 1300 1200 1100 1000 too 700 600 soo 400 300 Soo 100 N
"HEA 68 FLANG
- 1/61E
/ 0 5
30 15 100 125 150 175 200 t25 t50 075 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)
Curves for Heatup by Non-Nuclear Means, Coo own Following A Nuclear Shutdown and Low Power Physics sting Figure 3.4.11-5 (Page 1 of 1)
Unit 2 Minimum Reactor Vessel Metal Temperature vs.
Reactor Vessel Pressure (Valid until December 15, 2004)
Amendment No. 156/142 LaSalle I and 2 NITIAL RTndt VALUES ARE 52"F FOR BELTLINE.
409F FOR UPPER VESSEL AND 49"F FOR BOTTOM HEAD "BELTLINE CURVES ADJUSTED AS SHOWN EFPY SHIFT ('F) 32 23 HEATUPICOOLDOWN RATE OF COOLANT c 100"FIHR
'-UPPER VESSEL AND BELTLINE LIMITS BOTTOM HEAD CURVE P
- a.
- a.
0 I
-I 0
IU
'U
- 0)
- 0)
U, a-I 3.4.11-10
RCS P/T Limits 3.4.11 INITIAL RTndt VALUES ARE 52"F FOR BELTLINE, 40°F FOR UPPER VESSEL, AND 49°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 20 25 HEATUP/COOLDOWN RATE OF COOLANT
< 100F/HR
-UPPER VESSEL AND BELTLINE LIMITS
....... BOTTOM HEAD CURVE 0
25 so 7S 100 125 I5 17S MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 3.4.11-5 (Page 1 of 1)
Unit 2 P-T Curves for Heatup by Non-Nuclear Means, Cooldown Following a Nuclear Shutdown and Low Power Physics Testing up to 20 EFPY LaSalle 1 and 2 1100 C. 0 8OO o==
0 I.
@29o 400 Inl a.
0 3.4.11-10 Amendment No.
RCS P/T Limits 3.4.11 1400 1300 1200 1100 1000 900 goo 100 600 g00 400 g00 "100 100 0
10 7S 100 123 130 17S 100 21S3 150 7S 100 NIMUM REACTOR VESSEL METAL TEMPERATURE \\F)
P-T Curves for Operation with a Co e Critical other than Low Power Physics T ting Figure 3.4.11-6 (Page 1 of 1)
Unit 2 Minimum Reactor Vessel Metal Temperature vs.
Reactor Vessel Pressure (Valid until December 15, 2004) and 2 3.4.11-11 Amendment No. 156/142 N1
~I t
I Y/
I A N I
/
31/I 11PS=
cdrM'inmum M
-Temperat a
U a
-a U'
0 I
0 0
LaSalle I INITIAL RTndt VALUES ARE 62"F FOR BELTLINE.
40"F FOR UPPER VESSEL, AND 49OF FOR BOTTOM HEAD BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT ('F) 32 23 HEATUPICOOLDOWN RATE OF COOLANT c 100*FIHR "BELTLINE AND NON-BELTLINE LIMITS I
RCS P/T Limits 3.4.11 I
I-.
I 4-
-4.
I I--
___4-4 I
I i# 4
- a.
-a in "10.
Uj 6OO Mi 0 amo
- a.
IUl o
70 z
25 50 75 1oo IZ LW 175 2W 2
2m 0 MINIMUM REACTOR VESSEL METAL TEMPERATURE INITIAL RTndt VALUES ARE 52°F FOR BELTLINE, 40°F FOR UPPER
- VESSEL, AND 49°F FOR BOTTOM HEAD (OF)
Figure 3.4.11-6 (Page 1 of 1)
Unit 2 P-T Curves for Operation with a Core Critical other than Low Power Physics Testing up to 20 EFPY LaSalle 1 and 2 SMiimum Cri1icality Temperature 86F a
0 BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 20 25 HEATUP/COOLDOWN RATE OF COOLANT
< 100°F/HR BELTLINE AND NON-BELTLINE LIMITS 3.4.11-11 Amendment No.
ATTACHMENT 4 RETYPED PAGES FOR TECHNICAL SPECIFICATION CHANGES AND BASES CHANGES (FOR INFORMATION ONLY)
Retyped TS Pages Retyped Bases Pages
RCS P/T Limits 3.4.11 SURVEILLANCE REQUIREMENTS SURVILLACE RQUIRMENT SURVEILLANCE 4
NOTE -------------------
Only required to be performed during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.
Verify:
- a.
RCS pressure and RCS temperature are within the applicable limits specified in Figures 3.4.11-1, 3.4.11-2, 3.4.11-3 for Unit 1 up to 20 EFPY, and Figures 3.4.11-4, 3.4.11-5, and 3.4.11-6 for Unit 2 up to 20 EFPY;
- b.
RCS heatup and cooldown rates are
- IO0°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period; and
- c.
RCS temperature change during system leakage and hydrostatic testing is
- 20°F in any one hour period when the RCS pressure and RCS temperature are not within the limits of Figure 3.4.11-2 for Unit 1 up to 20 EFPY and Figure 3.4.11-5 for Unit 2 up to 20 EFPY.
FREQUENCY 30 minutes SR 3.4.11.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified in 15 minutes Figure 3.4.11-3 for Unit 1 up to 20 EFPY prior to and Figure 3.4.11-6 for Unit 2 up to 20 control rod EFPY.
withdrawal for the purpose of achieving criticality (continued)
LaSalle 1 and 2 I I 3.4.11-3 Amendment No.
RCS P/T Limits 3.4.11 INITIAL RTndt VALUES ARE
-30°F FOR BELTLINE, 40°F FOR UPPER VESSEL, AND 47°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT ('F) 20 114 HEATUP/COOLDOWN RATE OF COOLANT
< 20FIHR
-UPPER VESSEL AND BELTLINE LIMITS
- ------ BOTTOM HEAD CURVE 0
25 50 S
100 125 150 175 2O MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
Figure 3.4.11-1 (Page 1 of 1)
Unit 1 P-T Curves for Hydrostatic or Leak Testing up to 20 EFPY LaSalle 1 and 2 CL o
_z I
100 w
CL OW In lu,
O
'U
'U 0
7 100 0
3.4.11-6 Amendment No.
RCS P/T Limits 3.4.11 1100 1000 OW 0
I.
-J uJ o'uo
~0 O:
700 Sma 50 U)ow
'U U,
'U co 4
0 INITIAL RTndt VALUES ARE
-30°F FOR BELTLINE, 40°F FOR UPPER VESSEL, AND 47"F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 20 114 HEATUPICOOLDOWN RATE OF COOLANT
< 100°FIHR UPPER VESSEL AND BELTLINE LIMITS
-...... BOTTOM HEAD CURVE 0
5 o
75 100 15 150 17S MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
Figure 3.4.11-2 (Page 1 of 1)
Unit 1 P-T Curves for Heatup by Non-Nuclear Means, Cooldown Following a Nuclear Shutdown and Low Power Physics Testing up to 20 EFPY LaSalle 1 and 2 Amendment No.
3.4.11-7
RCS P/T Limits 3.4.11 INITIAL RTndt VALUES ARE
-30°F FOR BELTLINE, 40°F FOR UPPER
- VESSEL, AND 47°F FOR BOTTOM HEAD V/
Minimum Criticality Temperature 72°F7 0
5 50 7S 100 125 1
175 2s 22 0so MINIMUM REACTOR VESSEL METAL TEMPERATURE
('F)
Figure 3.4.11-3 (Page 1 of 1)
Unit 1 P-T Curves for Operation with a Core Critical other than Low Power Physics Testing up to 20 EFPY LaSalle 1 and 2 1100 CL a
0 40O W
I-
,-J
'I, uJ o, 700 z
lu a-0 BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (F) 20 114 HEATUP/COOLDOWN RATE OF COOLANT
< 100°FIHR
- BELTLINE AND NON-BELTLINE LIMITS Amendment No.
3.4.11-8
RCS P/T Limits 3.4.11 13W0 1200 1000 7W0 4[]
2WO 100 a
0 25 50 75 100 125 150 1715 2
MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 3.4.11-4 (Page 1 of 1)
Unit 2 P-T Curves for Hydrostatic or Leak Testing up to 20 EFPY LaSalle 1 and 2 0
I
-J il w
F (0
a.
INITIAL RTndt VALUES ARE 52°F FOR BELTLINE, 40°F FOR UPPER VESSEL.
AND 49°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 20 25 HEATUP/COOLDOWN RATE OF COOLANT
< 20°FIHR
-UPPER VESSEL AND BELTLINE LIMITS
...... BOTTOM HEAD CURVE Amendment No.
3.4.11-9
RCS P/T Limits 3.4.11 INITIAL RTndt VALUES ARE 52°F FOR BELTLINE, 40°F FOR UPPER VESSEL, AND 49°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (OF) 20 25 HEATUP/COOLDOWN RATE OF COOLANT
< 100"F/HR UPPER VESSEL AND BELTLINE LIMITS BOTTOM HEAD CURVE 0
2:
50 15 100 15 150 115 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 3.4.11-5 (Page 1 of 1)
Unit 2 P-T Curves for Heatup by Non-Nuclear Means, Cooldown Following a Nuclear Shutdown and Low Power Physics Testing up to 20 EFPY LaSalle 1 and 2 1100 CL Iw
_z 0
70 I-.
EM Z
o 0 Uj W60 CL50 0
Amendment No.
3.4.11-10
RCS P/T Limits 3.4.11 Iao0 11W v 0 0
w I
uJ u9
'U am t=
- 3 co W,
0 Minimum Critticality Temperature 86*F 46-1 INITIAL RTndt VALUES ARE 52°F FOR BELTLINE, 40°F FOR UPPER
- VESSEL, AND 49°F FOR BOTTOM HEAD BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 20 25 HEATUPICOOLDOWN RATE OF COOLANT
< IO0°FIHR
-BELTLINE AND NON-BELTLINE LIMITS O
25 9D 75 1W 125 1
175 200 290 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 3.4.11-6 (Page 1 of 1)
Unit 2 P-T Curves for Operation with a Core Critical other than Low Power Physics Testing up to 20 EFPY LaSalle 1 and 2 Amendment No.
3.4.11-11
RCS P/T Limits B 3.4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.11 RCS Pressure and Temperature (P/T)
Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.
These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips.
This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
The Specification contains P/T limit curves for heatup, cooldown, inservice leak and hydrostatic testing, and criticality and also limits the maximum rate of change of reactor coolant temperature.
The P/T limit curves are applicable for 20 effective full power years.
Each P/T limit curve defines an acceptable region for normal operation.
The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB).
The vessel is the component most subject to brittle failure.
Therefore, the LCO limits apply mainly to the vessel.
10 CFR 50, Appendix G (Ref.
1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.
Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.
It mandates the use of the American Society of Mechanical Engineers (ASME)
Code,Section III, Appendix G (Ref.
2).
The actual shift in the RTHDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref.
- 3) and 10 CFR 50, Appendix H (continued)
Revi si on LaSalle 1 and 2 B 3.4.11-1
RCS P/T Limits B 3.4.11 BASES BACKGROUND (Ref.
4).
The operating P/T limit curves will be adjusted, (continued) as necessary, based on the evaluation findings and the recommendations of Reference 5.
The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive.
At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit.
Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.
The non-nuclear heatup and cooldown curve applies during heatups with non-nuclear heat (e.g., recirculation pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram).
The curve provides the minimum reactor vessel metal temperatures based on the most limiting vessel stress.
The P/T criticality limits include the Reference 1 requirement that they be at least 40°F above the non-critical heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leak and hydrostatic testing.
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident.
In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.
The ASME Code,Section XI, Appendix E (Ref.
6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses.
They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a
condition that is unanalyzed.
Reference 5 establishes the methodology for determining the P/T limits.
Since the (continued)
Revision LaSalle 1 and 2 I
B 3.4.11-2
RCS P/T Limits B 3.4.11 BASES SURVEILLANCE SR 3.4,11.5. SR 3.4.11.6. and SR 3.4,11.7 (continued)
REQUIREMENTS Unit 1 and
- 1060F for Unit 2, monitoring of the flange temperature is required every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure the temperatures are within the specified limits.
The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on the rate of temperature change possible, at these temperatures.
SR 3.4.11.5 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs.
SR 3.4.11.6 is modified by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature
- 770F for Unit 1 and
< 91°F for Unit 2 in MODE 4, SR 3.4.11.7 is modified by a Note that requires the Surveillance to be initiated 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature
- 92 0F for Unit 1 and : 106 0F for Unit 2 in MODE 4.
The Notes contained in these SRs are necessary to specify when the reactor vessel flange and head flange temperatures are required to be verified to be within the specified limits.
REFERENCES
- 1.
- 2.
ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.
- 3.
ASTM E 185.
- 4.
- 5.
Regulatory Guide 1.99, Revision 2, May 1988.
- 6.
- ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
- 7.
UFSAR, Section 15.4.4.
Revi si on LaSalle 1 and 2 B 3.4.11-9
GE Nuclear Energy GE-NE-0000-0003-5526-LT04 1400 1300 1200 1100 all
- 1000 x
0L 900 0
U) 800 U.'
w o
700 I
0 k
600 I-S~
BoI-rC 3 500 HEAE w
668"F
) 400
'U 300 200 100 0
INITIAL RTndt VALUES ARE 52°F FOR BELTLINE.
40°F FOR UPPER VESSEL, AND 49°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (F) 20 25 HEATUP/COOLDOWN RATE OF COOLANT
< 100°FIHR UPPER VESSEL AND BELTLINE LIMITS
...... BOTTOM HEAD CURVE 0
25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)
LaSalle Unit 2 Composite Core Not Critical P-T Curves [Curve B] up to 20 EFPY
[100°F/hr or less coolant heatup/cooldown]
ATTACHMENT 6 Non-Proprietary General Electric Company Reports "Pressure-Temperature Curves for Exelon LaSalle Unit 1" and "Pressure-Temperature Curves for Exelon LaSalle Unit 2"
ATTACHMENT 7 Non-Proprietary General Electric Company Letter
General Electric Company 175 Curtner Avenue, San Jose, CA 95125 October 7, 2002 Mr. John Rommel Exelon Corporation 2601 North 21st Road Marseilles, IL 61341-9757
Subject:
References:
Transmittal of Composite 20 EFPY P-T Curves for LaSalle Units 1 and 2
- 1) Contract 833. Release 180; LS117, Fluence Analysis
- 2) Pressure Temperature Curves for Exelon LaSalle Unit 1, GE-NE-0000 0003-5526-02, Revision 0. July 2002
- 3) Pressure Temperature Curves for Exelon LaSalle Unit 2, GE-NE-0000 0003-5526-01, Revision 0, June 2002
Dear John,
The purpose of this letter is to transmit the subject Composite P-T Curves to supplement the Reference 1 and 2 reports.
If you have any questions regarding this letter do not hesitate to contact the undersigned. GE appreciates this opportunity to serve Exelon.
DJ Bouchie Technical Projects Manager Nuclear Services M/C 782 (408) 925-6213 FAX (408) 925-5210 c:\\documents and settingsvne24525Vocal settingsVemporary internetfiles\\o1k26\\isI I 7dxmtl.doc Date:
To:
w GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-LT04 Structural Assessment and Mitigation 175 Curtner Avenue M/C HME San Jose, CA 95125 (928) 301-2237 GE-N E-0000-0003-5526-LT04 Class II October 7, 2002 TO:
Daryl Bouchie cc: BJ Branlund FROM:
Lori Tilly
SUBJECT:
Composite 20 EFPY P-T Curves for LaSalle Units 1&2
REFERENCES:
- 1. LJ Tilly, "Pressure-Temperature Curves for Exelon LaSalle Unit 1",
GE-NE, San Jose, CA, July 2002 (GE-NE-0000-0003-5526-02, Revision 0)(GE Proprietary Information)
- 2. LJ Tilly, "Pressure-Temperature Curves for Exelon LaSalle Unit 2",
GE-NE, San Jose, CA, June 2002 (GE-NE-0000-0003-5526-01, Revision 0) (GE Proprietary Information)
Exelon has requested composite curves for 20 EFPY similar to Figures 5-10 and 5-11 provided in References 1&2 for 32 EFPY. The composite 20 EFPY Curve A provided in this letter is based upon the bottom head (Figure 5-1), upper vessel (Figure 5-2) and beltline curve (Figure 5-3) from Reference I for LaSalle Unit I and Reference 2 for LaSalle Unit 2. The composite 20 EFPY Curve B provided in this letter is based upon the bottom head (Figure 5-5), upper vessel (Figure 5-6) and beltline curve (Figure 5-7) from Reference I for LaSalle Unit 1 and Reference 2 for LaSalle Unit 2. These four (4) figures are consistent with Figures 5-10 and 5-11, where the bottom head curve is provided separate from the composite upper vessel plus beltline curve.
Please provide these curves to Exelon.
Should you have any questions with this transmittal, please contact me via e-mail or phone at (928) 301-2237.
Attachments:
[20°F/hr or less coolant heatup/cooldown]
[100°F/hr or less coolant heatup/cooldown]
[20°F/hr or less coolant heatup/cooldown]
[1 00°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy GE-NE-0000-0003-5526-LT04 1400 1300 I
i INITIAL RTndt VALUES ARE 120 30-F FOR BELTLINE,
/
/40'F FOR UPPER VESSEL, i
i ii I
iI AND 47*F FOR BOTTOM HEAD S1000
- 1 j
BELTLINE CURVES ADJUSTED AS SHOWN:
z ii EFPY SHIFT (OF)
CL 900 0
20 114 S800 I
w I
O 700 o-I HEATUP/COOLDOWN
- I IRATE OF COOLANT 0
600
<20F/HR
_z U..
- *500 BOTTOM 400 300 UPPER VESSEL AND BELTLINE 200 i
/
~REGION iLMT 72*F LIMITS S........BOTTOM HEAD 100
- CURVE 0
25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
LaSalle Unit 1 Composite Pressure Test P-T Curves [Curve A] up to 20 EFPY
[20°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy GE-NE-0000-0003-5526-LT04 FLANGE
- if'I" I
G10N 0
25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF) 1400 1300 1200 1100 CL 1000 w
900 0
all to 800 o
700 S600 z
S500 w
m 400 w
L3 300 200 LaSalle Unit 1 Composite Core Not Critical P-T Curves [Curve B] up to 20 EFPY
[100°F/hr or less coolant heatup/cooldown]
INITIAL RTndt VALUES ARE
-30*F FOR BELTLINE.
40*F FOR UPPER VESSEL, AND 47°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT ('F) 20 114 HEATUP/COOLDOWN RATE OF COOLANT
< 100°FIHR UPPER VESSEL AND BELTLINE LIMITS
...... BOTTOM HEAD CURVE 200 100 0
GE Nuclear Energy GE-NE-0000-0003-5526-LTO4 1400 1300 1200 1100 1000 0
IL 900 0
.j
-J w
0 800 w
O 700 w S600 500 BOrOt 0
400 U)
ILl 300 200 100 0
"" 1 25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE
('F)
INITIAL RTndt VALUES ARE 52°F FOR BELTLINE, 40°F FOR UPPER VESSEL, AND 49°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 20 25 HEATUPICOOLDOWN RATE OF COOLANT
< 20°F/HR SUPPER VESSEL AND BELTLINE LIMITS
.BOTTOM HEAD CURVE 200 LaSalle Unit 2 Composite Pressure Test P-T Curves [Curve A] up to 20 EFPY
[20 0F/hr or less coolant heatup/cooldown]
GE Nuclear Energy Engineering and Technology General Electric Company 175 Curtner Avenue San Jose, CA 95125 GE-NE-0000-0003-5526-02a Revision 0 Class I July 2002 Pressure-Temperature Curves For Exelon LaSalle Unit I Prepared by:
Structurl W
eniorc ' eer
-.Structural Mechanics and Materials Verified by:
D6>
c 5
4 B.D. Frew, Principal Engineer Structural Mechanics and Materials Approved-1y':
B.J. Branlund, Principal Engineer Structural Mechanics and Materials
GE-NE-0000-0003-5526-02a IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between Exelon and GE, Fluence Analysis, effective 11/14/01, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract.
The use of this information by anyone other than Exelon, or for any purpose other than that for which it is fumished by GE, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.
Copyright, General Electric Company, 2002
- iii -
GE Nuclear Energy
GE-NE-0000-0003-5526-02a EXECUTIVE
SUMMARY
This report provides the pressure-temperature curves (P-T curves) developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 2000 [1]. The P-T curve methodology includes the following: 1) The incorporation of ASME Code Case N-640. 2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of lc of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. This report incorporates a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14b], and is in compliance with Regulatory Guide 1.190.
CONCLUSIONS The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.
There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:
, Closure flange region (Region A)
- Core beltline region (Region B)
- Upper vessel (Regions A & B)
- Lower vessel (Regions B & C)
For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 1 00°F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [21 and the
- iv -
GE Nuclear Energy
GE-NE-0000-0003-5526-02a nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in this report. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20°F/hr or less must be maintained at all times.
The P-T curves apply for both heatup/cooldown and for both the 1/4T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, Kir, at 1/4T to be less than that at 3/4T for a given metal temperature.
Composite P-T curves were generated for each of the Pressure Test, Core Not Critical and Core Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate bottom head, beltline, upper vessel and closure assembly P-T limits. Separate P-T curves were developed for the upper vessel, beltline (at 20 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions. A composite P-T curve was also generated for the Core Critical condition at 20 EFPY.
GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE OF CONTENTS
1.0 INTRODUCTION
I 2.0 SCOPE OF THE ANALYSIS 3
3.0 ANALYSIS ASSUMPTIONS 5
4.0 ANALYSIS 6
4.1 INITIAL REFERENCE TEMPERATURE 6
4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 14 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 19
5.0 CONCLUSION
S AND RECOMMENDATIONS 49
6.0 REFERENCES
64
- vi-GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE OF APPENDICES APPENDIX A DESCRIPTION OF DISCONTINUITIES APPENDIX B PRESSURE-TEMPERATURE CURVE DATA TABULATION APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS APPENDIX D GE SIL 430 APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS APPENDIX F EVALUATION FOR UPPER SHELF ENERGY (USE)
APPENDIX G THICKNESS TRANSITION DISCONTINUITY EVALUATION
- vii-GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE OF FIGURES FIGURE 4-1:
FIGURE 4-2.
FIGURE 4-3.
FIGURE 5-1:
FIGURE 5-2:
FIGURE 5-3:
FIGURE 5-4:
FIGURE 5-5:
FIGURE 5-6:
FIGURE 5-7:
FIGURE 5-8:
FIGURE 5-9:
FIGURE 5-10:
FIGURE 5-11:
FIGURE 5-12:
SCHEMATIC OF REACTOR VESSEL ARRANGEMENT SHOWING PLATES AND WELDS 10 CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 30 FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 36 BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] [20-F/HR OR LESS COOLANT HEATUP/COOLDOWN]
52 UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] [200F/HR OR LESS COOLANT HEATUP/COOLDOWN]
53 BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 20 EFPY
[20°F/HR OR LESS COOLANT HEATUP/COOLDOWN]
54 BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 32 EFPY
[20'F/HR OR LESS COOLANT HEATUP/COOLDOWN]
55 BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [100'F/HR OR LESS COOLANT HEATUP/COOLDOWN]
56 UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [100°F/HR OR LESS COOLANT HEATUP/COOLDOWN]
57 BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 20 EFPY
[100*F/HR OR LESS COOLANT HEATUP/COOLDOWN]
58 BELTLINE P-T CURVES FOR CORE NOT CRITICAL [CURVE B] UP TO 32 EFPY
[10 0 F/IHR OR LESS COOLANT HEATUP/COOLDOWN]
59 COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 20 EFPY
[100*F/HR OR LESS COOLANT HEATUP/COOLDOWN]
60 COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 32 EFPY
[20*F/HR OR LESS COOLANT HEATUP/COOLDOWN]
61 COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 32 EFPY
[100*F/HR-OR LESS COOLANT HEATUP/COOLDOWN]
62 COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 32 EFPY
[100F/HR OR LESS COOLANT HEATUP/COOLDOWN]
63
- viii -
GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR LASALLE UNIT 1 VESSEL MATERIALS 11 TABLE 4-2: RTNDT VALUES FOR LASALLE UNIT 1 NOZZLE MATERIALS 12 TABLE 4-3: RTNT VALUES FOR LASALLE UNIT 1 WELD MATERIALS 13 TABLE 4-4: LASALLE UNIT 1 BELTLINE ART VALUES (20 EFPY) 17 TABLE 4-5: LASALLE UNIT 1 BELTLINE ART VALUES (32 EFPY) 18 TABLE 4-6:
SUMMARY
OF THE IOCFR50 APPENDIX G REQUIREMENTS 21 TABLE 4-7: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 23 TABLE 4-8: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH CRD (BOTTOM HEAD) CURVES A&B 23 TABLE 5-1: COMPOSITE AND INDIVIDUAL CURVES USED TO CONSTRUCT COMPOSITE P-T CURVES 51
- ix-
GE-NE-0000-0003-5526-02a
1.0 INTRODUCTION
The pressure-temperature (P-T) curves included in this report have been developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline.
Complete P-T curves were developed for 20 and 32 effective full power years (EFPY).
The P-T curves are provided in Section 5.0 and a tabulation of the curves is included in Appendix B. The P-T curves incorporate a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER [14b], and is in compliance with Regulatory Guide 1.190.
The methodology used to generate the P-T curves in this report is presented in Section 4.3 and is similar to the methodology used to generate the P-T curves in 2000 [1]. The P-T curve methodology includes the following: 1) The incorporation of ASME Code Case N-640 [4]. 2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 [6] for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of K1c of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. P-T curves are developed using geometry of the RPV shells and discontinuities, the initial RTNDT of the RPV materials, and the adjusted reference temperature (ART) for the beltline materials.
The initial RTNDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. The Charpy energy data used to determine the initial RTNDT values are tabulated from the Certified Material Test Report (CMTRs). The data and methodology used to determine initial RTNDT is documented in Section 4.1.
Adjusted Reference Temperature (ART) is the reference temperature when including irradiation shift and a margin term. Regulatory Guide 1.99, Rev. 2 [7] provides the methods for calculating ART. The value of ART is a function of RPV 1/4T fluence and beltline material chemistry. The ART calculation, methodology, and ART tables for 20 and 32 EFPY are included in Section 4.2. The 32 EFPY peak ID fluence value of GE Nuclear Energy
GE-NE-0000-0003-5526-02a 1.02 x 1018 n/cm 2 used in this report is discussed in Section 4.2.1.2. Beltline chemistry values are discussed in Section 4.2.1.1.
Comprehensive documentation of the RPV discontinuities that are considered in this report is included in Appendix A. This appendix also includes a table that documents which non-beltline discontinuity curves are used to protect the discontinuities.
Guidelines and requirements for operating and temperature monitoring are included in Appendix C. GE SIL 430, a GE service information letter regarding Reactor Pressure Vessel Temperature Monitoring is included in Appendix D. Appendix E demonstrates that all reactor vessel nozzles (other than the LPCI nozzle) are outside the beltline region. Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE). Finally, Appendix G contains an evaluation of the vessel wall thickness discontinuity in the beltline region. GE Nuclear Energy
GE-NE-0000-0003-5526-02a 2.0 SCOPE OF THE ANALYSIS The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 2000 [1]. The P-T curves in this report incorporate a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER [14b], and is in compliance with Regulatory Guide 1.190. A detailed description of the P-T curve bases is included in Section 4.3. The P-T curve methodology includes the following: 1) The incorporation of ASME Code Case N-640. 2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of Kic of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. Other features presented are:
", Generation of separate curves for the upper vessel in addition to those generated for the beltline, and bottom head.
"* Comprehensive description of discontinuities used to develop the non-beltline curves (see Appendix A).
The pressure-temperature (P-T) curves are established to the requirements of 10CFR50, Appendix G [8] to assure that brittle fracture of the reactor vessel is prevented. Part of the analysis involved in developing the P-T curves is to account for irradiation embrittlement effects in the core region, or beltline. The method used to account for irradiation embrittlement is described in Regulatory Guide 1.99, Rev. 2 [7].
In addition to beltline considerations, there are non-beltline discontinuity limits such as nozzles, penetrations, and flanges that influence the construction of P-T curves. The non-beltline limits are based on generic analyses that are adjusted to the maximum reference temperature of nil ductility transition (RTNDT) for the applicable LaSalle Unit 1 vessel components. The non-beltline limits are discussed in Section 4.3 and are also governed by requirements in [8].
Furthermore, curves are included to allow monitoring of the vessel bottom head and upper vessel regions separate from the beltline region. This refinement could minimize heating requirements prior to pressure testing. Operating and temperature monitoring GE Nuclear Energy
GE-NE-0000-0003-5526-02a requirements are found in Appendix C. Temperature monitoring requirements and methods are available in GE Services Information Letter (SIL) 430 contained in Appendix D. Appendix E demonstrates that all reactor vessel nozzles (other than the LPCI nozzle) are outside the beltline region. Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE). Finally, Appendix G contains an evaluation of the vessel wall thickness discontinuity in the beltline region. GE Nuclear Energy
GE-NE-0000-0003-5526-02a 3.0 ANALYSIS ASSUMPTIONS The following assumptions are made for this analysis:
For end-of-license (32 EFPY) fluence an 80% capacity factor is used to determine the EFPY for a 40-year plant life. The 80% capacity factor is based on the objective to have BWR's available for full power production 80% of the year (refueling outages, etc. -20%
of the year).
The shutdown margin is calculated for a water temperature of 680F, as defined in the LaSalle Unit 1 Technical Specification, Section 1.1. GE Nuclear Energy
GE-NE-0000-0003-5526-02a 4.0 ANALYSIS 4.1 INITIAL REFERENCE TEMPERATURE 4.1.1 Background The initial RTNDT values for all low alloy steel vessel components are needed to develop the vessel P-T limits. The requirements for establishing the vessel component toughness prior to 1972 were per the ASME Code Section III, Subsection NB-2300 and are summarized as follows:
- a.
Test specimens shall be longitudinally oriented CVN specimens.
- b.
At the qualification test temperature (specified in the vessel purchase specification), no impact test result shall be less than 25 ft-lb, and the average of three test results shall be at least 30 ft-lb
- c.
Pressure tests shall be conducted at a temperature at least 60°F above the qualification test temperature for the vessel materials.
The current requirements used to establish an initial RTNDT value are significantly different. For plants constructed according to the ASME Code after Summer 1972, the requirements per the ASME Code Section III, Subsection NB-2300 are as follows:
- a.
Test specimens shall be transversely oriented (normal to the rolling direction) CVN specimens.
- b.
RTNDT is defined as the higher of the dropweight NDT or 60OF below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion is met.
- c.
Bolt-up in preparation for a pressure test or normal operation shall be performed at or above the highest RTNDT of the materials in the closure flange region or lowest service temperature (LST) of the bolting material, whichever is greater.
1 OCFR50 Appendix G [8] states that for vessels constructed to a version of the ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses GE Nuclear Energy
GE-NE-0000-0003-5526-02a must be supplemented in an approved manner. GE developed methods for analytically converting fracture toughness data for vessels constructed before 1972 to comply with current requirements. These methods were developed from data in WRC Bulletin 217 [9] and from data collected to respond to NRC questions on FSAR submittals in the late 1970s. In 1994, these methods of estimating RTNDT were submitted for generic approval by the BWR Owners' Group [10], and approved by the NRC for generic use [11].
4.1.2 Values of Initial RTNDT and Lowest Service Temperature (LST)
To establish the initial RTNDT temperatures for the LaSalle Unit 1 vessel per the current requirements, calculations were performed in accordance with the GE method for determining RTNDT. Example RTNDT calculations for vessel plate, weld, HAZ, and forging, and bolting material LST are summarized in the remainder of this section.
For vessel plate material, the first step in calculating RTNDT is to establish the 50 ft-lb transverse test temperature from longitudinal test specimen data (obtained from certified material test reports, CMTRs [12]). For LaSalle Unit 1 CMTRs, typically six energy values were listed at a given test temperature, corresponding to two sets of Charpy tests. The lowest energy Charpy value is adjusted by adding 20F per ft-lb energy difference from 50 ft-lb.
For example, for the LaSalle Unit 1 beltline plate heat C5978-2 in the lower shell course, the lowest Charpy energy and test temperature from the CMTRs is 41 ft-lb at 400F. The estimated 50 ft-lb longitudinal test temperature is:
T5OL = 40°F + [(50 - 41) ft-lb
- 2°F/ft-lb] = 58OF The transition from longitudinal data to transverse data is made by adding 30°F to the 50 ft-lb transverse test temperature; thus, for this case above, TsoT = 580F + 30°F = 880F GE Nuclear Energy
GE-NE-0000-0003-5526-02a The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (TsoT-600F). Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is -10F. Thus, the initial RTNDT for plate heat C5978-2 would be 28°F; however, a semi curve-fit approach using CMTR data was performed [5] that resulted in a RTNDT for plate heat C5978-2 of 230F.
For the LaSalle Unit I beltline weld heat 1 P3571 with flux lot 3958 (contained in the middle shell), the CVN results are used to calculate the initial RTNDT. The 50 ft-lb test temperature is applicable to the weld material, but the 300F adjustment to convert longitudinal data to transverse data is not applicable to weld material. Heat 1 P3571 has a lowest Charpy energy of 40 ft-lb at 10F as recorded in weld qualification records.
Therefore, T5oT = 10°F + [(50 -40) ft-lb
- 2"F/ft-lb] = 30°F The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T50T - 600F). For LaSalle Unit 1, the dropweight testing to establish NDT was -30°F.
The value of (TsoT - 600F) in this example is -30°F; therefore, the initial RTNDT was -30°F.
For the vessel HAZ material, the RTNDT is assumed to be the same as for the base material, since ASME Code weld procedure qualification test requirements and post weld heat treat data indicate this assumption is valid.
For vessel forging material, such as nozzles and closure flanges, the method for establishing RTNDT is the same as for vessel plate material. For the feedwater nozzle at LaSalle Unit 1 (Heat Q2Q14VW-174W-1/6), the NDT is 40°F and the lowest CVN data is 48 ft-lb at 10°F. The corresponding value of (TwoT - 600F) is:
(Tso - 600F) = {[10 + (50 - 48) ft-lb "20F/ft-lb] + 300F} - 60°F = -160F. GE Nuclear Energy
GE-NE-0000-0003-5526-02a Therefore, the initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T50T-600F), which is 400F.
In the bottom head region of the vessel, the full Charpy longitudinal test data was fit using a hyperbolic tangent fit to determine the 50 ft-lb transition temperature. For the bottom dome plate of LaSalle Unit 1 (Heat C6003-3), the NDT is 40°F and the 50 ft-lb longitudinal transition temperature is 77°F. The corresponding value of (T5oT - 600F) was:
(T5oT - 601F) = {77°F + 301F} - 60*F = 47*F.
Therefore, the initial RTNDT was 470F.
For bolting material, the current ASME Code requirements define the lowest service temperature (LST) as the temperature at which transverse CVN energy of 45 ft-lb and 25 mils lateral expansion (MLE) were achieved. If the required Charpy results are not met, or are not reported, but the CVN energy reported is above 30 ft-lb, the requirements of the ASME Code Section III, Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 60°F (as discussed in Section 4.3.2.3) is the LST for the bolting materials. Charpy data for the LaSalle Unit I closure studs do not meet the 45 ft-lb, 25 MLE requirement at 10°F. Therefore, the LST for the bolting material is 700F. The highest RTNDT in the closure flange region is 120F, for the vessel shell flange materials. Thus, the higher of the LST and the RTNDT +60°F is 720F, the boltup limit in the closure flange region.
The initial RTNDT values for the LaSalle Unit 1 reactor vessel (refer to Figure 4-1 for LaSalle Unit 1 Schematic) materials are listed in Tables 4-1, 4-2, and 4-3. This tabulation includes beltline, closure flange, feedwater nozzle, and bottom head materials that are considered in generating the P-T curves. GE Nuclear Energy
GE Nuclear Energy TOP OF BELTUNE REGION 376.31" TOP OFACTIVE FUEL (TAF) 366.31" LPClI NOZZLE BOTTOM OF ACTIVE FUEL (aAF) 216.31 BOTTOM OF BELTLINE REGION 206.31" GE-NE-0000-0003-5526-02a 0
A GIRTH AXIAL WELDS 3-308 AXIAL WELDS 4-308 GIRTH SWELD
ý"T AXIUAL WEILES 2-307
/
'br77-7,7-WVZZIt TOP HEAD TOP HEAD FLANGE SHELL FLANGE SHELL 95 SHELL #4
, -SHELL
- 3 SHELL #2 SHELL#1 BOTTOM HEAD SUPPORT SKIRT Notes:
(1) Refer to Tables 4-1, 4-2, and 4-3 for reactor vessel components and theirj heat identifications.
(2) See Appendix E for the definition of the beitline region.
Figure 4-1: Schematic of the LaSalle Unit I RPV Showing Alfangement of Vessel Plates and Welds I I
I
GE Nuclear Energy GE-NE-0000-0003-5526-02a Table 4-1: RTNDT Values for LaSalle Unit 1 Vessel Materials ITEST CHARPY ENERGY (Trr-60) WEIGT COMPONENT HEAT TEMP.
(
(F)
I (°F)
I-L IN-PLATES & FORGINGS:
Top Head & Flange:
Vessel Flange, 308-02 Closure Flange, 319-02 Dome, 319-05 Upper Torus, 319-04 Lower Torus, 319-03 Shell Courses:
Upper Shell 305-04 Upper Int. Shell 305-04 Middle Shell 305-03 Low-Int. Shell 305-02 Lower Shell 305-01 Bottom Head:
Bottom Head Dome. 306-17 Lower Torus 306-18 Upper Torus 306-19 Support Skirt:
309-08 309-06 309-04 STUDS:
Closure Head Studs, 32-01 2V-659 ATF-1 12 ACT-USS-4P-1 997 Ser.1 18 C7434-1 C7434-1 C7376-2 C5987-1 C5987-2 C6003-2 C5996-2 C5979-2 C5996-1 A5333-1 B0078-1 C6123-2 C6345-1 C6318-1 C6345-2 C5978-1 C5978-2 C5979-1 C6003-3 C5540-1 C5328-1 C5328-2 C5505-2 C5445-3 5P2003 Ser.201 B1042-3 C7159-4 14716 24632 10 10 10 10 10 10 10 40 10 10 10 10 10 10 10 10 10 40 40 40 40 10 40 40 10 10 10 10 40 10 10 Value of RTNDT was obtained from semi curve-fit calculation using CMTR data.
Value of RT~oT is obtained from curve-fit of CMTR data.
70 92 65 65 65 63 76 65 62 64 65 56 73 77 109 80 93 53 62 73 36 54 64 55 63 70 81 70 28 68 110 76 76 74 55 79 49 71 63 60 67 49 60 88 66 94 48 60 92 39 78 51 62 96 67 74 61 25 97 91 67 67 73 35 51 50 66 49 77 53 70 73 77 72 67 48 41 65 40 82 51 59 73 70 103 68 34 45 143 1 43 38 I36 I39 -20
-20
-20
-20
-20 10
-20 12
-20
-18
-20
-20
-18
-20
-20
-20
-20 14 28 10 38
-20 10 10
-20
-20
-20
-20 60 LST 70 70 10 10
-10
-10
-10
-10
-10 10
-10
-10
-10
-10
-10
-10
-40
-20
-40 10
-10
-10 40
-10
-10
-10
-10
-10 40 10 60 10 10
-10
-10
-10 10
-10 12
-10
-10
-10
-10
-10
-10
-20
-20
-20 14 23*
10 47..
-10 10 10
-10
-10 40 10 60 Closure NutlWashers 326-02/03 38 36 39 70
-11
GE Nuclear Energy GE-NE-0000-0 03-5526-02a Table 4-2: RTNDT Values for LaSalle Unit I Nozz
!Materials TEST I CHARPY ENERGY I (Tsr-60) WEIGHTDRO COMPONENT HEAT TEMP.
(FTLB)
WEI H NDT (F)
I _______
_I_(-F)rT T
NOZZLES:
Recirc. Outlet Nozzle 314-02 Recirc. Inlet Nozzle 314-07 Steam Outlet Nozzles 316-07 Feedwater Nozzle, 316-02 Core Spray Nozzle 316-12 RHR/LPC1 Nozzles, 316-17 CDR Hydro Return Nozzle, 315-10 Jet Pump Nozzles, 314-12 Closure Head Inst. Nozzle, 318-07 Vent Nozzle. 318-02 Drain Nozzle, 315-14 Stabilizer Bracket 324-19 AV5840-OK9380 AV5840-OK9381 Q2014VW-175W Q206VW-175W AV4276-919074 AV4279-919236 AV4442-9J9176 AV4274-9H9176 Q2QI4VW-174W-1/6 AV4067-9H9168 AV4068-9H9169 02Q22W-569F-1/3 AV3142-9G9640 AV3138-9F-9231B/C Q2Q23W-346J-IA Q2024W-345J QIQ1VW-738T C4943-3 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 73 56 30 34 44 84 93 69 48 79 45 44 34 116 35 78 39 36 84 84 30 36 62 55 97 100 72 70 35 44 30 90 47 109 25 35 65 80 43 39 42 80 82 71 60 71 76 37 44 96 31 122 32 36
-20
-20 20 12
-4
-20
-20
-20
-16
-20 10 6
20
-20 18
-20 30 10 0
10 40 40 30 30 30 30 40 30 30 10 30 30 30 10 30 10 0
10 40 40 30 30 30 30 40 30 30 10 30 30 30 10 a
GE Nuclear Energy GE-NEE-0000-0003-5526-02a Table 4-3: RTNDT Values for LaSalle Unit 1 Weld Materials COMPONENT WELDS:
Vertical Welds:
2-307 Bottom Shell Long Seams 1-308 Upper Shell Long Seams 2-308 Upper Inter. Shell Long Seams 1-308 Upper Shell Long Seams 2-307 Bottom Shell Long Seams 1-308 Upper Shell Long Seams 3-308 Middle Shell Long Seams 3-308 Middle Shell Long Seams 4-308 Lower Inter. Shell Long Seams 4-308 Lower Inter. Shell Long Seams 1-319 Closure Head Seg. Lower Torus 2-319 Closure Head Seg. Upper Torus 1-319 Closure Head Seg. Lower Torus Girth Welds:
3-306 Bottom Hd. Build up for sup. Skirt 5-306 Bottom Hd. Dome to Side Seg, 6-306 Bottom Hd. Low. To Up Side Seg.
6-306 Bottom Hd. Low. To Up Side Seg.
4-307 Inlay in Bot. Sd for Core Sup Attch.
9-307 Bottom Head to Lower Shell 3-319 Close. Hd. Torus to Close. Hd. Fig.
9-307 Bottom Head to Lower Shell 3-319 Close. Hd. Torus to Close. Hd. Fig.
6-308 Upper Vessel Shell Girth Seam 9-307 Bottom Head to Lower Shell 6-308 Upper Vessel Shell Girth Seam 15-308 Flange to Upper Shell 6-308 Upper Vessel Shell Girth Seam 6-308 Upper Vessel Shell Girth Seam 6-308 Upper Vessel Shell Girth Seam 6-308 Upper Vessel Shell Girth Seam 15-308 Flange to Upper Shell 5-319 Closure Hd. Upper Torus to Dome 4-309 Support Skirt Forging to Bot. Hd.
4-309 Support Skirt Forging to Bot. Hd.
1-313 Up. Assy to Lower Closing Seams 1-313 Up. Assy to Lower Closing Seams 1-313 Up. Assy to Lower Closing Seams 4-319 Close. Hd. Upper Torus to Lower HEAT 21935-1092-3889 12008-1092-3889 305424-1092-3889 IP3571-1092-3958 305414-1092-3947 12008-1092-3947 FOAA EAIB 305414-1092-3951 305424-1092-3889 10120-0091-3458 51874-0091-3458 51912-0091-3490 10137-0091-3999 5P5622-0091-0831 2P5755-0091-0831 6329637-0091-3458 6329637-0091-3999 90099-0091-3977 90136-0091-3998 4P651 9-0091-0145 4P6519-0091-0842 4P6519-0091-0653 7TEST TEMP.
(*Fi 10 10 10 10 10 10 10 10 10 10 10 10 10 10
-20
-10 10 10 10 10 0
0
-40 606L40-0091-3489 1 10 1 96 1 95 CHARPY ENERGY (FT-LB) 97 97 90 90 83 83 82 1 87 1 92 40 82 92 125 46 66 91 124 46 80 92 130 118 129 1 107 66 1 61 1 62 82 1 87 1 92 124 130 1 122 89 64 87 93 84 92 101 108 107 95 81 103 101 96 110 98 46 57 87 80 65 108 97 109 101 59 63 86 82 88 103 89 107 102 48 73 77 (TwO-60)
(°F)
DROP WEIGH NDT
-50
-50
-50
-50
-50
-50
-50
-50
-50
-50
-50
-50
-50
-50
-80
-70
-50
-50
-50
-50
-50
-30
-50
-50
-50
-50
-50
-50
-50
-50
-50
-50
-80
-70
-50
-50 ITTRTum
(°F)
-50
-50
-50
-30
-50
-50
-50
-50
-50
-50
-50
-50
-50
-50
-80
-70
-50
-50
-50
-50
-60
-52
-60
-50
-50
-50
-60
-52
-100
-50 1
-50
-50
-50
-60
-80
-60
GE-NE-0000-0003-5526-02a 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (Rev 2) provides the methods for determining the ART. The Rev 2 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section. An evaluation of ART for all beltline plates and welds was made and summarized in Table 4-4 for 20 EFPY and Table 4-5 for 32 EFPY.
4.2.1 Regulatory Guide 1.99, Revision 2 (Rev 2) Methods The value of ART is computed by adding the SHIFT term for a given value of effective full power years (EFPY) to the initial RTNDT. For Rev 2, the SHIFT equation consists of two terms:
SHIFT = ARTNDT + Margin
- where, ARTNDT = [CF]*f (0.28 -0.log 01 Margin = 2(al2 + C2)0.5 CF = chemistry factor from Tables 1 or 2 of Rev. 2 f = YT fluence /1019 Margin = 2(ai2 + 0., )0.5 a, = standard deviation on initial RTNDT, which is taken to be 0°F.
= standard deviation on ARTNDT, 28°F for welds and 17°F for base material, except that 1,, need not exceed 0.50 times the ARTNDT value.
ART = Initial RTNDT + SHIFT The margin term cra has constant values in Rev 2 of 170F for plate and 28°F for weld.
However, a,& need not be greater than 0.5* ARTNDT. Since the GE/BWROG method of estimating RTNDT operates on the lowest Charpy energy value (as described in Section 4.1.2) and provides a conservative adjustment to the 50 ft-lb level, the value of a, is taken to be 0°F for the vessel plate and weld materials. b.-
GE Nuclear Energy
GE-NE-0000-0003-5526-02a 4.2.1.1 Chemistry The vessel beltline chemistries were obtained from the LaSalle Unit 1 NRC RAI submittal [13]. The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of Rev 2, to determine a chemistry factor (CF) per Paragraph 1.1 of Rev 2 for welds and plates, respectively.
4.2.1.2 Fluence A LaSalle Unit 1 flux for the vessel ID wall [14a] is calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER [14b], and is in compliance with Regulatory Guide 1.190. The flux as documented in [14a] is determined for the currently licensed power of 3489 MWt using a conservative power distribution and is conservatively used from the beginning to the end of the licensing period (32 EFPY).
The peak fast flux for the RPV inner surface from Reference 14 is 1.01 e9 n/cm2-s. The peak fast flux for the RPV inner surface determined from surveillance capsule flux wires removed during the outage in Spring 1994 after Fuel Cycle 6 at a full power of 3323 MWI is 4.41e8 n/cm 2-s [5]. Linearly scaling the Reference 5 flux by 1.05 to the currently licensed power of 3489 MWt results in an estimated flux of 4.63e8 n/cm2-s.
Therefore, the Reference 14 flux bounds the flux determined from the surveillance capsule flux wire results by 218%.
The time period 32 EFPY is 1.01e9 sec, therefore the RPV ID surface fluence is as follows: RPV ID surface fluence = 1.01e9 n/cm2-s*1.01e9 s = 1.02e18 n/cm2. This fluence applies to the lower-intermediate and middle shells, the vertical welds for these shells, and the girth welds. The fluence is adjusted for the lower shell and the vertical welds for the lower shell based upon a peak / lower shell location ratio of 0.44 (at an elevation of approximately 230" above vessel "0"); hence the peak ID surface fluence used for these components is 4.49e17 n/cm2. Similarly, the fluence is adjusted for the LPCI nozzle based upon a peak / LPCI nozzle location ratio of 0.244 (at an elevation of GE Nuclear Energy
GE-NE-0000-0003-5526-02a approximately 372" and at 450, 1350, and 2250 azimuths); hence the peak ID surface fluence used for this component is 2.49e17 n/cm 2.
4.2.2 Limiting Beltline Material The limiting beltline material signifies the material that is estimated to receive the greatest embrittlement due to irradiation effects combined with initial RTNDT. Using initial RTNDT, chemistry, and fluence as inputs, Rev 2 was applied to compute ART. Table 4-4 lists values of beltline ART for 20 EFPY and Table 4-5 lists the values for 32 EFPY.
Sections 4.3.2.2.2 and 4.3.2.2.3 provide a discussion of the limiting material. GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a Table 4-4: LaSalle Unit 1 Beltline ART Values (20 EFPY)
Thidcess in inches =
Thickness in hidres=
Thickness in inches=
613 7.13 613 Middle & Lewer-laermedirne Plrai t..
Welds 3-30.4.3064-30 & 1-313 RatioPeak/Location= 1.00 32 EFPY Peak I.D. fluencer 1.02E818 rionA2 32 EFPY Peak 1/4 T fluence =
7.1E817 r/cm^2 20 EFPY Peak 1/4 T fluence = 4.4E.17 n/cmt2 Lner rlse satld Welds 2-1 Ratio Peak/Location = 0.44 32 EFPY Peak I.D. fluence = 4 49E+17 n/can2 LocAin - 229 7/r" Elvatei 32 EFPY Peak 1/4 T fluence =
2.9E+17 n/CtT2 20 EFPY Peak 1/4 T fluence =
1.8E+17
/cmnr2 LPOi Noile Ratio Peak/ Location = 0244 32 EFPY Peak L.O. fluence = 2.49E817 n/cn'2 Locaioo --1?" Eleesijos 32 EFPY Peak 114 T fluence =
1.7E-17 nCra2 20 EFPY Peak 1/4 Tfluence =
1.1E-17 nr/at2 Initial 1/4 T 20 EFPY 20 EFPY 20 EFPY COMPONENT HEAT OR HEATILOT
%Cu
%Ni CF RTý Fluenoe a RTr oa or Margin Shift ART
""F n/cm^2
-F
-F
- F
°F PLATES:
Lower Shell Assy 307-04 G--5603-1 C5978-1 0.110 0.580 74 14 1 8E-17 12 0
6 12 24 38 G-5603-2 C5978-2 0110 0.590 74 23 1,8E÷17 12 0
6 12 24 47 G.-5603-3 C5979-1 0120 0660 84 10 1.8E-17 14 0
7 14 27 37 Lower-Iitermedlata Shell Assy 308-06 G5604-1 C6345-1 0.150 0.490 104
-20 44E+17 28 0
14 28 57 37 G5604-2 C6318-1 0120 0510 81
-20 4.4E.17 22 0
11 22 44 24 G5604-3 C6345-2 0150 0510 105
-20 4.4E817 29 0
14 29 57 37 Middle Shell Any 308-05 G5605-1 A5333-1 0120 0.540 82
-10 44.E17 22 0
11 22 45 35 G5605-2 B0078-1 0 150 0.500 105
-10 4 4E+17 29 0
14 29 57 47 G5605-3 C6123-2 0 130 0+680 93
-10 4 4E-17 25 0
13 25 51 41 WELDS:
Middle 3-308A.B.C 305424/3889 0.273 0.629 189.5
.50 44E+17 52 0
26 52 104 54 1P357113958 0.283 0755 212
-30 44E+17 58 0
28 56 114 84 Lower.Intermedlate 4-308 A.B.C 30541413947 0+337 0.609 209
-50 4.4E+17 57 0
28 56 113 63 12008i3947 0.235 0.975 233
-50 4-4E+17 64 0
28 56 120 70 305414&12008 Tandem 0+286 0792 219
-50 4.4E+17 60 0
28 56 116 66 Lower 2-307 A.B.C 21935/3889 0 183 0-704 172
-50 1.8E+17 28 0
14 28 56 6
12008/3889 0235 0.975 233
-50 1.8E817 38 0
19 38 76 26 21935&12008 tandem 0213 0.867 209
-50 1.8E+17 34 0
17 34 68 18 Glirth 6-306 6329637 0205 0.105 98
-50 4 4E.17 27 0
13 27 54 4
1-313 4P6519 0131 0060 64
-52 4.4AE17 18 0
9 18 35
-17 FORGINGS:
LPC Nozzle 02022W 0100 0+820 67 10 1.1E17 8
0 4
8 15 25
GE Nuclear Energy GE-NE-0000-0003-5526-02a Table 4-5: LaSalle Unit 1 Beltline ART Values (32 EFPY)
Thickness in inches =
Thickness in inches=
Thickness in inches=
6.13 7.13 6.13 Middl & Lewer-teermedtfat Paes med Weds 3"3M, 4-,30 6318 & 1-313 Ratio Peak/ Location 1.00 32 EFPY Peak I.D. fence -
1.02E618 n/cmn2 32 EFPY Peak 114 T ftjence 7.16E+17 nicnm2 32 EFPY Peak 1/4 T thence 7.1E+17 n/CrA2 Le-er Nloe mod W ' 2.30M Ratio Peak? Location - 0.44 32 EFPY Peak I.D. Auence s 4.49E÷17 rIcn/A2 Locate. - 229 7/r* DEat.o 32 EFPY Peak 114 T tzuence z 2.9E+17 ftoit2 32 EFPY Peak 14Ttfsence 2.9E+17 n/A2 8161 Nook.
Ratio Peak Location
- 0.244 32 EFPY Peak LD. fluenoe 2.49E617 n/crt2 Lacon--~3
" D
- e.
32 EFPY Peak 1/4 T fluence
- 1.7E+17 ngrn2 32 EFPY Peak 1/4 T fluence 1.7E+17 n/cmn2 Initia 1/4 T 32 EFPY 32 EFPY 32 EFPY COMPONENT HEAT OR HEATILOT
%Cu
%Ni CF RTOWr Ruence
& RTmT a
Margin Shift ART "F
ncm-A2 7F
-F 7
F
- F PLATES:
Lower Shell Assy 307-04 G-5603-1 C5978-1 0.110 0.580 74 14 2.9E+17 16 0
8 16 32 46 G-5603-2 C5976-2 0.110 0.590 74 23 2.9E÷17 16 0
8 18 32 56 G-5603-3 C5979-1 0.120 0.660 84 10 2.9E+17 18 0
9 18 36 46 LoweF-lrtenmedlate Shell Asy 30806 G5604-1 C6345-1 0.150 0.490 104
-20 7.1E÷17 36 0
17 34 70 50 G5604-2 C6318-1 0.120 0.510 81
-20 7.1E+17 28 0
14 28 57 37 G5604-3 C6345-2 0.150 0.510 105
.20 7.1E+17 37 0
17 34 71 51 Middle Shell May 30845 G5605-1 A5333-1 0.120 0.540 82
-10 7.16E17 29 0
14 29 58 48 G5605-2 B0078-1 0.150 0.500 105
-10 7.1E+17 37 0
17 34 71 61 G5605-3 C6123-2 0.130 0.680 93
-10 7.1E-17 33 0
16 33 65 55 WELDS:
Middle 3-308 A.BC 305424)3889 0.273 0.629 189.5
-50 7.1E+17 67 0
28 56 123 73 1P357113958 0.283 0.755 212
-30 7.1E÷17 74 0
28 56 130 100 Lower-tntennedlate 4-308 A.B.C 305414/3947 0.337 0.609 209
-50 7.1E+17 73 0
28 56 129 79 12008/3947 0235 0.975 233
-50 7.16E17 82 0
28 56 138 88 305414112008 Tandem 0.28 0.792 219
-50 7.16E17 77 0
28 56 133 83 Lower 2-307 A.B.C 2193513889 0.183 0.704 172
-50 2.9E+17 37 0
19 37 74 24 12008)3889 0.235 0.975 233
-50 2.9E+17 50 0
25 50 101 51 21935&12008 tandem 0.213 0.867 209
-50 2.9E+17 45 0
23 45 90 40 Girth 6-308 6329637 0.205 0.105 98
-50 7.1E17 34 0
17 34 69 19 1-313 4P6519 0.131 0,060 64
-52 7.16.17 22 0
11 22 45
-7 FORGINGS:
LPCINoaOe 02022W 0.100 0.820 67 10 1.7E-17 10 0
5 10 21 31
GE-NE-0000-0003-5526-02a 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 4.3.1 Background Nuclear Regulatory Commission (NRC) 10CFR50 Appendix G [8] specifies fracture toughness requirements to provide adequate margins of safety during the operating conditions that a pressure-retaining component may be subjected to over its service lifetime. The ASME Code (Appendix G of Section Xl of the ASME Code [6]) forms the basis for the requirements of 1 OCFR50 Appendix G. The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.
There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:
"* Closure flange region (Region A)
"* Core beltline region (Region B)
"* Upper vessel (Regions A & B)
"* Lower vessel (Regions B & C)
The closure flange region includes the bolts, top head flange, and adjacent plates and welds. The core beltline is the vessel location adjacent to the active fuel, such that the neutron fluence is sufficient to cause a significant shift of RTNDT. The remaining portion of the vessel (i.e., upper vessel, lower vessel) include shells, components like the nozzles, the support skirt, and stabilizer brackets; these regions will also be called the non-beltline region.
For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 1OO0 F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also GE Nuclear Energy
GE-NE-0000-0003-5526-02a developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in the sections below. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20 °F/hr or less must be maintained at all times.
The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 314T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4T location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness, KIr, at 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.
The applicable temperature is the greater of the 10CFR50 Appendix G minimum temperature requirement or the ASME Appendix G limits. A summary of the requirements is as follows in Table 4-6: GE Nuclear Energy
GE-NE-0000-0003-5526-02a Table 4-6: Summary of the 1 OCFR50 Appendix G Requirements Operating Condition and Pressure Minimum Temperature Requirement
- 1. Hydrostatic Pressure Test & Leak Test (Core is Not Critical) - Curve A
- 1. At < 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 600F*
- 2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 90°F II. Normal operation (heatup and cooldown),
including anticipated operational occurrences
- a. Core not critical - Curve B
- 1. At < 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 600F*
- 2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 120°F
- b. Core critical - Curve C
- 1. At < 20% of preservice hydrotest Larger of ASME Limits + 40°F or of a. 1 pressure, with the water level within the normal range for power operation
- 2. At > 20% of preservice hydrotest Larger of ASME Limits + 40°F or of pressure a.2 + 40°F or the minimum permissible temperature for the inservice system hydrostatic pressure test 60°F adder is included by GE as an additional conservatism as discussed in Section 4.3.2.3 There are four vessel regions that affect the operating limits: the closure flange region, the core beltline region, and the two regions in the remainder of the vessel (i.e., the upper vessel and lower vessel non-beltline regions). The closure flange region limits are controlling at lower pressures primarily because of 1 OCFR50 Appendix G [8]
requirements. The non-beltline and beltline region operating limits are evaluated according to procedures in 1 OCFR50 Appendix G [8], ASME Code Appendix G [6], and Welding Research Council (WRC) Bulletin 175 [15]. The beltline region minimum temperature limits are adjusted to account for vessel irradiation. GE Nuclear Energy
GE-NE-0000-0003-5526-02a GE PROPRIETARY INFORMATION DELETED 4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beltline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the neutron fluence is not sufficient (<1.0e17 n/cm2) to cause any significant shift of RTNDT. Non-beltline components include nozzles (see Appendix E),
the closure flanges, some shell plates, the top and bottom head plates and the control rod drive (CRD) penetrations.
Detailed stress analyses of the non-beltline components were performed for the BWR/6 specifically for the purpose of fracture toughness analysis. The analyses took into account all mechanical loading and anticipated thermal transients. Transients considered include 1 00F/hr start-up and shutdown, SCRAM, loss of feedwater heaters or flow, loss of recirculation pump flow, and all transients involving emergency core cooling injections. Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used according to the ASME Code [6] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). Plots were developed for the limiting BWR/6 components: the feedwater nozzle (FW) and the CRD penetration (bottom GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a head). All other components in the non-beltline regions are categorized under one of these two components as described in Tables 4-7 and 4-8.
Table 4-7: Applicable BWR/5 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B Discontinuity Identification FW Nozzle LPCI Nozzle CRD HYD System Return Core Spray Nozzle Recirculation Inlet Nozzle Steam Outlet Nozzle Main Closure Flange Support Skirt Stabilizer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle Steam Water Interface Instrumentation Nozzle Shell CRD and Bottom Head (B only)
Top Head Nozzles (B only)
Recirculation Outlet Nozzle (B only)
Table 4-8: Applicable BWR/5 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B Discontinuity Identification CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Shell**
Support Skirt**
Shroud Support Attachments**
Core AP and Liquid Control Nozzle**
- These discontinuities are added to the bottom head curve discontinuity list to assure that the entire bottom head is covered, since separate bottom head P-T curves are provided to monitor the bottom head.
GE-NE-0000-0003-5526-02a GE Nuclear Energy The P-T curves for the non-beitline region were conservatively developed for a large BWR/6 (nominal inside diameter of 251 inches). The analysis is considered appropriate for LaSalle Unit 1 as the plant specific geometric values are bounded by the generic analysis for a large BWR/6, as determined in Section 4.3.2.1.1 through Section 4.3.2.1.4. The generic value was adapted to the conditions at LaSalle Unit I by using plant specific RTNDT values for the reactor pressure vessel (RPV). The presence of nozzles and CRD penetration holes of the upper vessel and bottom head, respectively, has made the analysis different from a shell analysis such as the beltline.
This was the result of the stress concentrations and higher thermal stress for certain transient conditions experienced by the upper vessel and the bottom head.
4.3.2.1.1 Pressure Test - Non-Beitline, Curve A (Using Bottom Head)
In a finite element analysis [ ], the CRD penetration region was modeled to compute the local stresses for determination of the stress intensity factor, KI. The evaluation was modified to consider the new requirement for Mm as discussed in ASME Code Section XI Appendix G [6] and shown below. The results of that computation were K, = 143.6 ksi-in"2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T - RTNDT)
GE Nuclear Energy GE-NE-0000-0003-5526-02a was 840F.
The limit for the coolant temperature change rate is 20 °Flhr or less.
The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]
was based on a thickness of 8.0 inches; hence, tV2 = 2.83. The resulting value obtained was:
Mm = 1.85 for It <2 Mm = 0.926 Vf for 2< ft <3.464 = 2.6206 Mm = 3.21 for ý >3.464 Kim is calculated from the equation in Paragraph G-2214.1 [6] and Kib is calculated from the equation in Paragraph G-2214.2 [6]:
Kim = Mm " ypm =
ksi-in" 2 Kib = (2/3) Mm - 0 pb =
ksi-in'2 The total K, is therefore:
GE Nuclear Energy GE-NE-0000-0003-5526-0 2a I* = 1.5 (Kim+ K~b) + Mm" (aml
+ (2/3)
) = 143.6 ksi-in 12 This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNDT) for a specific K, is based on the K, equation of Paragraph A-4200 in ASME Appendix A [17]:
(r - RTNDT) = In [(KI - 33.2)/ 20.734] / 0.02 (T - RTNDT) = In [(144 - 33.2) / 20.7341 / 0.02 (T - RTNDT) = 84°F The generic curve was generated by scaling 143.6 ksi-in"2 by the nominal pressures and calculating the associated (T - RTNDT):
The highest RTNDT for the bottom head plates and welds is 470F, as shown in Tables 4-1, 4-2, and 4-3.
GE Nuclear Energy GE-NE-0000-0003-5526-02a Second, the P-T curve is dependent on the calculated K, value, and the K. value is proportional to the stress and the crack depth as shown below:
K, x a (ca)"*2 (4-1)
The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t/4. Thus, K, is proportional to R/(t)"2. The generic curve value of R/(t)' 2, based on the generic BWR/6 bottom head dimensions, is:
Generic:
R / (t)"2 = 138 / (8)1/2 = 49 inch" 2 (4-2)
The LaSalle Unit 1 specific bottom head dimensions are R = 127.4 inches and t =7.38 inches minimum [19], resulting in:
LaSalle Unit 1 specific:
R / (t)"2 = 127.4 / (7.38)"/2 = 46.9 inch/ 2 (4-3)
GE-NE-0000-0003-5526-02a Since the generic value of R/(t)12 is larger, the generic P-T curve is conservative when applied to the LaSalle Unit I bottom head.
4.3.2.1.2 Core Not Critical Heatup/Cooldown - Non-Beltline Curve B (Using Bottom Head)
As discussed previously, the CRD penetration region limits were established primarily for consideration of bottom head discontinuity stresses during pressure testing.
Heatup/cooldown limits were calculated by increasing the safety factor in the pressure testing stresses (Section 4.3.2.1.1) from 1.5 to 2.0.
The calculated value of K, for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with KIR, the material fracture toughness. A safety factor of 2.0 is used for the core not critical. Therefore, the I* value for the core not critical condition is (143.6 / 1.5). 2.0 = 191.5 ksi-in12. GE Nuclear Energy
GE-NE-0000-0003-5526-02a Therefore, the method to solve for (T - RTNDT) for a specific K, is based on the Kjc equation of Paragraph A-4200 in ASME Appendix A [17] for the core not critical curve:
(T - RTNDT) = In [(K, - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 1020F The generic curve was generated by scaling 192 ksi-in' by the nominal pressures and calculating the associated (T - RTNDT):
Core Not Critical CRD Penetration K, and as a Function of Pressure (T - RTNDT)
Nominal Pressure K,
T - RTNDT (psig)
(ksi-in'2)
(*F) 1563 192 102 1400 172 95 1200 147 85 1000 123 73 800 98 57 600 74 33 400 49
-14 The highest RTNDT for the bottom head plates and welds is 47°F, as shown in Tables 4-1, 4-2, and 4-3.
As discussed in Section 4.3.2.1.1 an evaluation is performed to assure that the CRD discontinuity bounds the other discontinuities that are to be protected by the CRD curve with respect to pressure stresses (see Tables 4-7, 4-8, and Appendix A). With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe than those of the other components being bounded. Therefore, for heatup/cooldown conditions, the CRD penetration provides bounding limits. GE Nuclear Energy
GE-NE-0000-0003-5526-02a GE Nuclear Energy Figure 4-2.
CRD Penetration Fracture Toughness Limiting Transients GE Nuclear Energy GE-NE-0000-0003-5526-02a 4.3.2.1.3 Pressure Test - Non-Beitline Curve A (Using Feedwater Nozzle/Upper Vessel Region)
The stress intensity factor, K1, for the feedwater nozzle was computed using the methods from WRC 175 [15] together with the nozzle dimension for a generic 251-inch BWR/6 feedwater nozzle. The result of that computation was K, = 200 ksi-in" for an applied pressure of 1563 psig preservice hydrotest pressure.
The respective flaw depth and orientation used in this calculation is perpendicular to the maximum stress (hoop) at a depth of 1/4T through the corner thickness.
To evaluate the results, K, is calculated for the upper vessel nominal stress, PR/t, according to the methods in ASME Code Appendix G (Section III or Xl). The result is compared to that determined by CBIN in order to quantify the K magnification associated with the stress concentration created by the feedwater nozzles. A calculation of K, is shown below using the BWR/6, 251-inch dimensions:
Vessel Radius, R, 126.7 inches Vessel Thickness, t, 6.1875 inches Vessel Pressure, P, 1563 psig Pressure stress: a = PR / t = 1563 psig - 126.7 inches / (6.1875 inches) = 32,005 psi.
The Dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding a = 34.97 ksi. The factor F (a/rm) from Figure A5-1 of WRC-1 75 is 1.4 where:
a = % (tn 2 + tv 2)112
=2.36 inches t, = thickness of nozzle
= 7.125 inches t, = thickness of vessel
= 6.1875 inches rn = apparent radius of nozzle
= r, + 0.29 r,=7.09 inches ri = actual inner radius of nozzle
= 6.0 inches rc = nozzle radius (nozzle comer radius)
= 3.75 inches GE-NE-0000-0003-5526-02a Thus, air,, = 2.36 17.09 = 0.33. The value F(a/r,), taken from Figure A5-1 of WRC Bulletin 175 for an air, of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K,, is 1.5 a (na)"2'- F(a/rn):
Nominal K, = 1.5 34.97- (a-2.36)1/2. 1.4 = 200 ksi-in'12 The method to solve for (T - RTNDT) for a specific KI is based on the K1, equation of Paragraph A-4200 in ASME Appendix A [17] for the pressure test condition:
(T - RTNDT) = In [(KI - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(200 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 104.2°F The generic pressure test P-T curve was generated by scaling 200 ksi-in' 2 by the nominal pressures and calculating the associated (T - RTNDT), GE Nuclear Energy
GE-NE-0000-0003-5526-02a The highest RTNDT for the feedwater nozzle materials is 40°F as shown in Tables 4-1, 4-2, and 4-3. The generic pressure test P-T curve is applied to the LaSalle Unit 1 feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 400F. GE Nuclear Energy
GE-NE-0000-0003-5526-02a Second, the P-T curve is dependent on the K, value calculated. The LaSalle Unit 1 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and K, are shown below:
Vessel Radius, Rv 127 inches Vessel Thickness, t, 6.69 inches Vessel Pressure, P, 1563 psig Pressure stress: a = PR / t = 1563 psig-127 inches / (6.69 inches) = 29,671 psi. The Dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding a = 32.64 ksi. The factor F (a/rn) from Figure A5-1 of WRC-175 is determined where:
a=
(t, 2 + tv 2)1 2
2.31 inches tn
thickness of nozzle
= 6.38 inches tv = thickness of vessel
= 6.69 inches r, = apparent radius of nozzle
ri + 0.29 rc=7.29 inches ri
actual inner radius of nozzle
6.13 inches r,
nozzle radius (nozzle comer radius)
= 4.0 inches Thus, air, = 2.31 / 7.29 = 0.32. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an airn of 0.32, is 1.5. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (aa)'2. F(a/rn):
Nominal K2= 1.5 32.64. (n -2.31)1'2-1.5 = 197.9 ksi-in112 GE Nuclear Energy
GE-NE-0000-0003-5526-02a 4.3.2.1.4 Core Not Critical HeatuplCooldown - Non-Beltline Curve B (Using Feedwater Nozzle/Upper Vessel Region)
The feedwater nozzle was selected to represent non-beltline components for fracture toughness analyses because the stress conditions are the most severe experienced in the vessel. In addition to the pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow in hotter vessel coolant.
Stresses were taken from a finite element analysis done specifically for the purpose of fracture toughness analysis [
]. Analyses were performed for all feedwater nozzle transients that involved rapid temperature changes. The most severe of these was normal operation with cold 40°F feedwater injection, which is equivalent to hot standby, see Figure 4-3.
The non-beltline curves based on feedwater nozzle limits were calculated according to the methods for nozzles in Appendix 5 of the Welding Research Council (WRC)
Bulletin 175 [15].
The stress intensity factor for a nozzle flaw under primary stress conditions (Kip) is given in WRC Bulletin 175 Appendix 5 by the expression for a flaw at a hole in a flat plate:
Kip = SF -a (na)1/2 - F(a/r.)
(4-4) where SF is the safety factor applied per WRC Bulletin 175 recommended ranges, and F(a/r,) is the shape correction factor. GE Nuclear Energy
GE-NE-0000-0003-5526-02a Figure 4-3.
Feedwater Nozzle Fracture Toughness Limiting Transient Finite element analysis of a nozzle comer flaw was performed to determine appropriate values of F(a/rn) for Equation 4-4. These values are shown in Figure A5-1 of WRC Bulletin 175 [15].
The stresses used in Equation 4-4 were taken from design stress reports for the feedwater nozzle. The stresses considered are primary membrane, apn, and primary bending, apb. Secondary membrane, asm, and secondary bending, ad,, stresses are included in the total K, by using ASME Appendix G [6] methods for secondary portion, K,I:
K*s = Mm (asm + (2/3) -sb)
(4-5) GE Nuclear Energy
GE-NE-0000-0003-5526-02a In the case where the total stress exceeded yield stress, a plasticity correction factor was applied based on the recommendations of WRC Bulletin 175 Section 5.C.3 [15].
However, the correction was not applied to primary membrane stresses because primary stresses satisfy the laws of equilibrium and are not self-limiting. Kip and K1, are added to obtain the total value of stress intensity factor, K4. A safety factor of 2.0 is applied to primary stresses for core not critical heatup/cooldown conditions.
Once K, was calculated, the following relationship was used to determine (T - RTNDT).
The method to solve for (T - RTNDT) for a specific K, is based on the K1c equation of Paragraph A-4200 in ASME Appendix A [17]. The highest RTNDT for the appropriate non-beltline components was then used to establish the P-T curves.
(T - RTNDT) = In [(K, - 33.2) / 20.734] / 0.02 (4-6)
Example Core Not Critical HeatuplCooldown Calculation for Feedwater Nozzle/Upper Vessel Region The non-beltline core not critical heatup/cooldown curve was based on the feedwater nozzle analysis, where feedwater injection of 40°F into the vessel while at operating conditions (551.4°F and 1050 psig) was the limiting normal or upset condition from a brittle fracture perspective. The feedwater nozzle comer stresses were obtained from finite element analysis [
]. To produce conservative thermal stresses, a vessel and nozzle thickness of 7.5 inches was used in the evaluation.
However, a thickness of 7.5 inches is not conservative for the pressure stress evaluation. Therefore, the pressure stress (Gpm) was adjusted for the actual vessel thickness of 6.1875 inches (i.e., cpm = 20.49 ksi was revised to 20.49 ksi
- 7.5 inches/6.1875 inches = 24.84 ksi). These stresses, and other inputs used in the generic calculations, are shown below:
apm = 24.84 ksi asm = 16.19 ksi ay = 45.0 ksi t=
6.1875 inches apb = 0.22 ksi Gsb = 19.04 ksi a = 2.36 inches rr = 7.09 inches t, = 7.125 inches GE Nuclear Energy
GE-NE-0000-0003-5526-02a In this case the total stress, 60.29 ksi, exceeds the yield stress, ay, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to the following equation based on the assumptions and recommendation of WRC Bulletin 175 [15]. (The value of specified yield stress is for the material at the temperature under consideration. For conservatism, the temperature assumed for the crack root is the inside surface temperature.)
R = [ay, - apm + ((atotai - ays) / 30)] / (atot. - apm)
(4-7)
For the stresses given, the ratio, R = 0.583. Therefore, all the stresses are adjusted by the factor 0.583, except for apm. The resulting stresses are:
(pm = 24.84 ksi asm = 9.44 ksi apb = 0.13 ksi cs =11.10 ksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]
was based on the 4a thickness ; hence, t'2 = 3.072. The resulting value obtained was:
Mm = 1.85 for -J:52 Mm = 0.926 ft-for 2< ft: <3.464 = 2.845 Mm = 3.21 for ft >3.464 The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an aIft of 0.33, is therefore, F (a /r,) = 1.4 Kip is calculated from Equation 4-4:
Ip = 2.0- (24.84 + 0.13) * (n -2.36)1"7 1.4 Kip = 190.4 ksi-in'12 GE Nuclear Energy
GE-NE-0000-0003-5526-02a K,, is calculated from Equation 4-5:
Kis = 2.845 - (9.44 + 2/3 - 11.10)
Kis = 47.9 ksi-in"12 The total K, is, therefore, 238.3 ksi-in112.
The total K, is substituted into Equation 4-6 to solve for (T - RTNDT):
(T - RTNDT) = In [(238.3-33.2) / 20.734] / 0.02 (T-RTNDT) = 115°F The curve was generated by scaling the stresses used to determine the K,; this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 40°F water injected into the hot reactor vessel nozzle. In the base case that yielded a K, value of 238 ksi-in112, the pressure is 1050 psig and the hot reactor vessel temperature is 551.4 0F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (Tsaturation - 40) / (551.4 - 40). From K, the associated (T - RTNDT) can be calculated: GE Nuclear Energy
GE-NE-0000-0003-5526-02a Core Not Critical Feedwater Nozzle K, and (T - RTNDT) as a Function of Pressure Nominal Pressure Saturation Temp.
R i*
(T - RTNDT)
(psig)
(OF)
(ksi-in"2 )
(OF) 1563 604 0.23 303 128 1400 588 0.34 283 124 1200 557 0.48 257 119 1050 551 0.58 238 115 1000 546 0.62 232 113 800 520 0.79 206 106 600 489 1.0 181 98 400 448 1.0 138 81
- Note: For each change in condition, there is determination of KI.
stress for each pressure a corresponding change and saturation temperature to R that influences the The highest non-beltline RTNDT for the feedwater nozzle at LaSalle Unit I is 40°F as shown in Tables 4-1, 4-2, and 4-3. The generic curve is applied to the LaSalle Unit I upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 40°F as discussed in Section 4.3.2.1.3.
4.3.2.2 CORE BELTLINE REGION The pressure-temperature (P-T) operating limits for the beltline region are determined according to the ASME Code. As the beltline fluence increases with the increase in operating life, the P-T curves shift to a higher temperature. GE Nuclear Energy 6-
GE-NE-0000-0003-5526-02a The stress intensity factors (K1), calculated for the beltline region according to ASME Code Appendix G procedures [6], were based on a combination of pressure and thermal stresses for a 1/4T flaw in a flat plate. The pressure stresses were calculated using thin-walled cylinder equations. Thermal stresses were calculated assuming the through wall temperature distribution of a flat plate; values were calculated for 100°F/hr coolant thermal gradient. The shift value of the most limiting ART material was used to adjust the RTNDT values for the P-T limits.
An evaluation was performed [22] for the vessel wall thickness transition discontinuity located between the lower and lower-intermediate shells in the beitline region.
Appendix G of this report contains an update of the evaluation.
4.3.2.2.1 Beltline Region - Pressure Test The methods of ASME Code Section Xl, Appendix G [6] are used to calculate the pressure test beltline limits. The vessel shell, with an inside radius (R) to minimum thickness (tmim) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:
am = PR / tmi, (4-8)
The stress intensity factor, Kim, is calculated using Paragraph G-2214.1 of the ASME Code.
The calculated value of Kim for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with K~c, the material fracture toughness. A safety factor of 2.0 is used for the core not critical and core critical conditions.
The relationship between Kic and temperature relative to reference temperature (T - RTNDT) is based on the Kic equation of Paragraph A-4200 in ASME Appendix A [17]
for the pressure test condition: GE Nuclear Energy
GE-NE-0000-0003-5526-02a GE Nuclear Energy Kim" SF = ic = 20.734 exp[0.02 (T - RTNDT )] + 33.2 (4-9)
This relationship provides values of pressure versus temperature (from KIR and (T-RTNDT), respectively).
GE's current practice for the pressure test curve is to add a stress intensity factor, K*,
for a coolant heatup/cooldown rate of 20°F/hr to provide operating flexibility. For the core not critical and core critical condition curves, a stress intensity factor is added for a coolant heatup/cooldown rate of 100°F/hr. The Kf calculation for a coolant heatup/cooldown rate of 100OF/hr is described in Section 4.3.2.2.3 below.
4.3.2.2.2 Calculations for the Beitline Region - Pressure Test This sample calculation is for a pressure test pressure of 1105 psig at 32 EFPY. The following inputs were used in the beltline limit calculation:
Adjusted RTNDT = Initial RTNDT + Shift A = -30 + 130 = 100=F (Based on ART values in Section 4.2)
Vessel Height H = 863.3 inches Bottom of Active Fuel Height B = 216 inches Vessel Radius (to inside of clad)
R = 126.7 inches Minimum Vessel Thickness (without clad) t = 6.13 inches Pressure is calculated to include hydrostatic pressure for a full vessel:
P = 1105 psi + (H - B) 0.0361 psi/inch = P psig
= 1105 + (863.3 - 216) 0.0361 = 1128 psig Pressure stress:
a PR/t
= 1.128 - 126.7 / 6.13 = 23.3 ksi (4-10)
(4-11)
GE Nuclear Energy GE-NE-0000-0003-5526-02a The value of Mm. for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]
was based on a thickness of 6.13 inches (the minimum thickness without cladding);
hence, t"2 = 2.48. The resulting value obtained was:
Mm = 1.85 for f_<2 Mm = 0.926 Vft for 2< -J <3.464 = 2.29 Mm = 3.21 for ft- >3.464 The stress intensity factor for the pressure stress is Kim = Mm - a. The stress intensity factor for the thermal stress, Kit, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 20°F/hr instead of 100°F/hr.
Equation 4-9 can be rearranged, and 1.5 Kim substituted for Kic, to solve for (T - RTNDT).
Using the Kic equation of Paragraph A-4200 in ASME Appendix A [17], Kim = 53.4, and Kt = 3.01 for a 20°F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:
(T - RTNDT)
= ln[(1.5 - Kim + Kit - 33.2) 120.734] / 0.02 (4-12)
= ln[(1.5 - 53.4 + 3.01 - 33.2) / 20.734] 1 0.02
= 43.90F T can be calculated by adding the adjusted RTNDT:
T = 43.9 + 100 = 143.90F for P = 1105 psig 4.3.2.2.3 Beltline Region - Core Not Critical Heatup/Cooldown The beltline curves for core not critical heatup/cooldown conditions are influenced by pressure stresses and thermal stresses, according to the relationship in ASME Section Xl Appendix G [6]:
Kic = 2.0- Kim +Kit (4-13)
GE-NE-0000-0003-5526-02a where K(im is primary membrane K due to pressure and Kit is radial thermal gradient K due to heatup/cooldown.
The pressure stress intensity factor Kim is calculated by the method described above, the only difference being the larger safety factor applied. The thermal gradient stress intensity factor calculation is described below.
The thermal stresses in the vessel wall are caused by a radial thermal gradient that is created by changes in the adjacent reactor coolant temperature in heatup or cooldown conditions. The stress intensity factor is computed by multiplying the coefficient Mt from Figure G-2214-1 of ASME Appendix G [6] by the through-wall temperature gradient AT,,
given that the temperature gradient has a through-wall shape similar to that shown in Figure G-2214-2 of ASME Appendix G [6]. The relationship used to compute the through-wall AT, is based on one-dimensional heat conduction through an insulated flat plate:
a 2T(x,t) / 8 x2 = I / 13 (aT(x,t) / OR)
(4-14) where T(x,t) is temperature of the plate at depth x and time t, and P3 is the thermal diffusivity.
The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that aT(x,t) / t = dT(t) / dt = G, where G is the coolant heatup/cooldown rate, normally 100°F/hr. The differential equation is integrated over x for the following boundary conditions:
- 1. Vessel inside surface (x = 0) temperature is the same as coolant temperature, To.
- 2. Vessel outside surface (x = C) is perfectly insulated; the thermal gradient dT/dx = 0.
The integrated solution results in the following relationship for wall temperature: GE Nuclear Energy
GE-NE-0000-0003-5526-02a T = Gx 2 / 2P - GCx / P + To (4-15)
This equation is normalized to plot (T - TO) / ATw versus x / C.
The resulting through-wall gradient compares very closely with Figure G-2214-2 of ASME Appendix G [6]. Therefore, ATw calculated from Equation 4-15 is used with the appropriate Mt of Figure G-2214-1 of ASME Appendix G [6] to compute Kit for heatup and cooldown.
The Mt relationships were derived in the Welding Research Council (WRC)
Bulletin 175 [15] for infinitely long cracks of 1/4T and 1/8T. For the flat plate geometry and radial thermal gradient, orientation of the crack is not important.
4.3.2.2.4 Calculations for the Beltline Region Core Not Critical Heatup/Cooldown This sample calculation is for a pressure of 1105 psig for 32 EFPY. The core not critical heatup/cooldown curve at 1105 psig uses the same Krm as the pressure test curve, but with a safety factor of 2.0 instead of 1.5. The increased safety factor is used because the heatup/cooldown cycle represents an operational rather than test condition that necessitates a higher safety factor. In addition, there is a Kit term for the thermal stress.
The additional inputs used to calculate Kit are:
Coolant heatup/cooldown rate, normally 100°F/hr G = 100 °F/hr Minimum vessel thickness, including clad thickness C = 0.588 ft (7.06 inches)
(the maximum vessel thickness is conservatively used)
Thermal diffusivity at 550°F (most conservative value) f3 = 0.354 ff/ hr [21]
Equation 4-15 can be solved for the through-wall temperature (x = C), resulting in the absolute value of AT for heatup or cooldown of: GE Nuclear Energy
GE-NE-0000-0003-5526-02a AT = GC2 / 23 (4-16)
= 100- (0.588)2/ (2 0.354) = 49°F The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.308) can be interpolated from ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, KK = Mt" AT = 15.1, can be calculated. KIm has the same value as that calculated in Section 4.3.2.2.2.
The pressure and thermal stress terms are substituted into Equation 4-9 to solve for (T - RTNDT):
(T - RTNDT)
=
ln[((2 - Kim + K4t) - 33.2) / 20.734] /0.2 (4-17)
=
ln[(2 53.4 + 15.1 -33.2)/
20.734] /0.02
=
72.7 *F T can be calculated by adding the adjusted RTNDT:
T = 72.7 + 100 = 172.7 °F for P = 1105 psig 4.3.2.3 CLOSURE FLANGE REGION 10CFR50 Appendix G [8] sets several minimum requirements for pressure and temperature in addition to those outlined in the ASME Code, based on the closure flange region RTNDT. In some cases, the results of analysis for other regions exceed these requirements and closure flange limits do not affect the shape of the P-T curves.
However, some closure flange requirements do impact the curves, as is true with LaSalle Unit I at low pressures.
The approach used for LaSalle Unit I for the bolt-up temperature was based on a conservative value of (RTNDT + 60), or the LST of the bolting materials, whichever is greater. The 60°F adder is included by GE for two reasons: 1) the pre-1971 GE Nuclear Energy
GE-NE-0000-0003-5526-02a requirements of the ASME Code Section III, Subsection NA, Appendix G included the 60°F adder, and 2) inclusion of the additional 60°F requirement above the RTNDT provides the additional assurance that a flaw size between 0.1 and 0.24 inches is acceptable. As shown in Tables 4-1 and 4-3, the limiting initial RTNDT for the closure flange region is represented by both the top head and vessel shell flange materials at 12°F, and the LST of the closure studs is 70"F; therefore, the bolt-up temperature value used is 72°F. This conservatism is appropriate because bolt-up is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.
10CFR50 Appendix G, paragraph IV.A.2 [8] including Table 1, sets minimum temperature requirements for pressure above 20% hydrotest pressure based on the RTNDT of the closure region. Curve A temperature must be no less than (RTNDT + 900F) and Curve B temperature no less than (RTNDT + 1200F).
For pressures below 20% of preservice hydrostatic test pressure (312 psig) and with full bolt preload, the closure flange region metal temperature is required to be at RTNDT or greater as described above. At low pressure, the ASME Code [6] allows the bottom head regions to experience even lower metal temperatures than the flange region RTNDT. However, temperatures should not be permitted to be lower than 680F for the reason discussed below.
The shutdown margin, provided in the LaSalle Unit 1 Technical Specification, is calculated for a water temperature of 68 0F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod fully withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 680F limit, further extensive calculations would be required to justify a lower temperature. The 72°F limit for the upper vessel and beltline region and the 68°F limit for the bottom head curve apply when the head is on and tensioned and when the head is off while fuel is in the vessel.
When the head is not tensioned and fuel is not in the vessel, the requirements of 10CFR50 Appendix G [8] do not apply, and there are no limits on the vessel temperatures. GE Nuclear Energy
GE-NE-0000-0003-5526-02a 4.3.2.4 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of I OCFR50 Appendix G [8], Table 1. Table 1 of [8] requires that core critical P-T limits be 40°F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 40°F for pressures above 312 psig.
Table 1 of 1 OCFR50 Appendix G [8] indicates that for a BWR with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT + 60°F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 720F, based on an RTNDT of 120F. In addition, above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 160°F or the temperature required for the hydrostatic pressure test (Curve A at 1105 psig). The requirement of closure region RTNDT + 160°F does cause a temperature shift in Curve C at 312 psig. GE Nuclear Energy
GE-NE-0000-0003-5526-02a
5.0 CONCLUSION
S AND RECOMMENDATIONS The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.
There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:
Closure flange region (Region A)
Core beltline region (Region B)
Upper vessel (Regions A & B)
Lower vessel (Regions B & C)
For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 1 00F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20OF/hr or less must be maintained at all times.
The P-T curves apply for both heatup/cooldown and for both the 1/4T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 114T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, Kir, at 1/4T to be less than that at 3/4T for a given metal temperature. GE Nuclear Energy
GE-NE-0000-0003-5526-02a The following P-T curves were generated for LaSalle Unit 1.
" Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel and closure assembly P-T limits. A separate Bottom Head Limits (CRD Nozzle) curve is also individually included with the composite curve for the Pressure Test and Core Not Critical condition.
"* Separate P-T curves were developed for the upper vessel, beltline (at 20 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.
A composite P-T curve was also generated for the Core Critical condition at 20 and 32 EFPY. The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel, bottom head, and closure assembly P-T limits.
Using the flux from Reference 14 the P-T curves are beltline limited above 1040 psig for curve A and 1090 psig for curve B for 20 EFPY. The P-T curves are beltline limited above 710 psig for curve A and 660 psig for curve B for 32 EFPY.
Table 5-1 shows the figure numbers for each P-T curve. A tabulation of the curves is presented in Appendix B. GE Nuclear Energy
GE-NE-0000-0003-5526-02a Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves Figure Table Numbers Curve Curve Description Numbers for for Presentation of Presentation of the P-T Curves the P-T Curves Curve A Bottom Head Limits (CRD Nozzle)
Figure 5-1 B-1 & B-3 Upper Vessel Limits (FW Nozzle)
Figure 5-2 B-1 & B-3 Beltline Limits for 20 EFPY Figure 5-3 B-3 Beltline Limits for 32 EFPY Figure 5-4 B-1 Curve B Bottom Head Limits (CRD Nozzle)
Figure 5-5 B-1 & B-3 Upper Vessel Limits (FW Nozzle)
Figure 5-6 B-1 & B-3 Beltline Limits for 20 EFPY Figure 5-7 B-3 Beltline Limits for 32 EFPY Figure 5-8 B-1 Curve C Composite Curve for 20 EFPY**
Figure 5-9 B-4 A, B, & C Composite Curves for 32 EFPY Bottom Head and Composite Curve A Figure 5-10 B-2 for 32 EFPY*
Bottom Head and Composite Curve B Figure 5-11 B-2 for 32 EFPY*
Composite Curve C for 32 EFPY*
Figure 5-12 B-2 The Composite Curve A & B curve is the more limiting of three limits: 1 0CFR50 Bolt up Limits, Upper Vessel Limits (FW Nozzle), and Beltline Limits. A separate Bottom Head Limits (CRD Nozzle) curve is individually included on this figure.
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1400 1300 1200 1100 09 S700 S600 500 Uj 400 w
IL 300 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)
Figure 5-1: Bottom Head P-T Curve for Pressure Test [Curve A]
[20 0F/hr or less coolant heatup/cooldown] SINITIAL RTndt VALUE IS 47°F FOR BOTTOM HEAD HEATUP/COOLDOWN RATE OF COOLANT
< 20°F/HR 200 100 0
0
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1100
"- 1000 w
0.900 0
Lu
" 800 o
700 Ir 600 z
S500
~400 IL 1400 1300 1200 I
I INITIAL RTndt VALUE IS
/40F FOR UPPER VESSELI HEATUP/COOLDOWN RATE OF COOLANT
< 20°F/HR UPPER VESSEL LIMITS (Including F
FLANGE REGION 72F Flange and FW Nozze Limits) 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-2: Upper Vessel P-T Curve for Pressure Test [Curve A]
[20°F/hr or less coolant heatup/cooldown]
300 200 100 0
kin.=
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1400 1300 1200 1100 1000
.900 0
I
-O 800 w
o 700 I
U 40 LU re60 Iz I
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 20 EFPY
[2O0FIhr or less coolant heatup/cooldown] BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT ('F) 20 114 HEATUP/COOLDOWN RATE OF COOLANT
< 20°F/HR BELTLINE LIMITS 300 200 100 0
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1400 1300 1200 1100 Isl 1000 w
x 9L 900 0
I 800 UJ 0
800 00 mu a,
o 700 I
Uj
,u 4
600 z
3 500 o*
400 mu 300 200 100 0
0 25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
INITIAL RTndt VALUE IS
-30"F FOR BELTLINE BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 32 130 HEATUPICOOLDOWN RATE OF COOLANT
< 20"F/HR BELTLINE LIMITS 200 Figure 5-4: Beltline P-T Curve for Pressure Test [Curve A] up to 32 EFPY
[20°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy
-a
- 0.
- a.
LU
- 0.
Z I
tg l
cU C4 0
LU 0
IL LUJ 0.
1400 1300 1200 1100 1000 900 800 700 600 500 400 300 200 100 0
GE-NE-0000-0003-5526-02a SINITIAL RTndt VALUE IS 47°F FOR BOTTOM HEAD HEATUP/COOLDOWN RATE OF COOLANT
<100OF/HR 0
25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (-F)
Figure 5-5: Bottom Head P-T Curve for Core Not Critical [Curve B]
[100°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1400 1300 1200 1100 fA 1000 U) 800 z
ul 0
7 00 600 500
.10 ul U)40 o0 0 w
3: 00 2 00 1400 300 0
25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(°F)
SINITIAL RTndt VALUE IS 40°F FOR UPPER VESSEL HEATUP/COOLDOWN RATE OF COOLANT
< IOO°F/HR
-UPPER VESSEL LIMITS (Including Flange and FW Nozzle Umits) 200 Figure 5-6: Upper Vessel P-T Curve for Core Not Critical [Curve B]
[1 00°F/hr or less coolant heatup/cooldown]
GE Ncler EnrgyGE-NE-OOOO-0003-5526-02a 1400 1300 1200 1100 LU
- 0.
ca 0
U)
Ua 0:
CL 1000 BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT ('F) 900 20 114 800 700I I
HEATUPICOOLDOWN I
RATE OF COOLANT 600
< 100*F/HR 500 400 300 200 I10CFR50
-BELTLINE LIMITS IBOLTUPI 100 I
72*F 0
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
Figure 5-7: Beltline P-T Curve for Core Not Critical [Curve B] up to 20 EFPY
[1 OO0F/hr or less coolant heatup/cooldown] GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1400 1300 1200oo 1 J INITIAL RTndt VALUE IS
-30°F FOR BELTLINE 1100 1000 BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (-F) 900 32 130 800 700 HEATUPICOOLDOWN RATE OF COOLANT 600 T<
100"FIHR 500 400 I
300 3
200 IOCFR50 BELTLINE LIMITS BOLTUP 100 72F 0
1 0
25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-8: Beltline P-T Curves for Core Not Critical [Curve B] up to 32 EFPY
[1 00°F/hr or less coolant heatup/cooldown] _;
- 0.
a U1 x
- n.
0 I
U)
I.J Ul it lu UL U,
0:
ILU
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1400 1INITIAL RTndt VALUES ARE 1300
-30°F FOR BELTLINE, 40°F FOR UPPER
- VESSEL, 1200 I
AND 47°F FOR BOTTOM HEAD 1100 "1000 BELTLINE CURVE ioo I iADJUSTED AS SHOWN:
I'EFPY SHIFT (-F) 9L 900 20 114 U9 800 HEATUPICOOLDOWN RATE OF COOLANT 0
700;
< 100°F/HR S600 2
I I-I
- 500 DiI I
400 1312 PSIG 300 I
200 I
BELTLINE AND NON-BELTLINE 100 Minimum Criticality LIMITS 0
Temperature 72°F 0
0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(°F)
Figure 5-9: Composite Core Critical P-T Curves [Curve C] up to 20 EFPY
[1000 F/hr or less coolant heatup/cooldown]
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1400 1300 1200 1100 S
""1000 0 1000 mu 9 900 0
"'800 o
700 S600 fi 500 U) 400 w
IL 300 200 100 0
0 25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
INITIAL RTndt VALUES ARE I
-30°F FOR BELTLINE, 40°F FOR UPPER VESSEL, I AND 47°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (F) 32 130 HEATUP/COOLDOWN RATE OF COOLANT
< 20F/HR
-UPPER VESSEL AND BELTLINE LIMITS BOTTOM HEAD CURVE 200 Figure 5-10: Composite Pressure Test P-T Curves [Curve A] up to 32 EFPY
[20°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1400 1300 I
1200 INITIAL RTndt VALUES ARE
-30°F FOR BELTLINE, 40F FOR UPPER VESSEL, 1100 AND 47°F FOR BOTTOM HEAD 1000
- BELTLINE CURVES ADJUSTED AS SHOWN
- 9.
900 EFPY SHIFT ('F) 0 -
,32 130 800 w >
[.
o 700 S"HEATUP/COOLDOWN 600 RATE OF COOLANT
_z L
_<100°F/HR I--
° 500 BOTTOM HEAD 68F aI 400 300 200
-UPPER VESSEL AND BELTLINE LIMITS 100 REGIN.....
BOTTOM HEAD 72*F CURVE 0
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-11: Composite Core Not Critical P-T Curves [Curve B] up to 32 EFPY
[1 00°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy GE-NE-0000-0003-5526-02a 1400 1300 1200 1100 0, "1000 w
x a-900 0
Go I
-.I 6800 O
700 I-.
U 4 600 It 500 cg 400 uJ a.
ii li i
Minimum Criticality Temperature 72°F 4
I t
0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
Figure 5-12: Composite Core Critical P-T Curves [Curve C] up to 32 EFPY
[100°F/hr or less coolant heatup/cooldown] INITIAL RTndt VALUES ARE
-30°F FOR BELTLINE, 40°F FOR UPPER
- VESSEL, AND 47"F FOR BOTTOM HEAD BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 32 130 HEATUP/COOLDOWN RATE OF COOLANT 5 10O0FIHR
-BELTLINE AND NON-BELTLINE LIMITS 300 200 100 0
GE-NE-0000-0003-5526-02a
6.0 REFERENCES
- 1.
B.J. Branlund, "Pressure-Temperature Curves for CornEd LaSalle Unit 1," GE-NE, San Jose, CA, May 2000, (GE-NE-B1 3-02057-00-06R1, Revision 1).
- 2.
GE Drawing Number 731 E776, "Reactor Vessel Thermal Cycles," GE-NED, San Jose, CA, Revision 3 (GE Proprietary).
- 3.
GE Drawing Number 158B8136, "Reactor Vessel Nozzle Thermal Cycles,"
GE-NED, San Jose, CA, Revision 7 (GE Proprietary).
- 4.
"Alternative Reference Fracture Toughness for Development of P-T Limit CurvesSection XI, Division 1," Code Case N-640 of the ASME Boiler & Pressure Vessel Code, Approval Date February 26, 1999.
- 5.
T. A. Caine, "LaSalle County Station Units 1 and 2 Fracture Toughness Analysis per 10CFR50 Appendix G," GE-NE, San Jose, CA, March 1988, (SASR 88-10).
- 6.
"Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section III or XI of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.
- 7.
"Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.
- 8. "Fracture Toughness Requirements," Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.
- 9.
Hodge, J. M., "Properties of Heavy Section Nuclear Reactor Steels," Welding Research Council Bulletin 217, July 1976.
- 10. GE Nuclear Energy, NEDC-32399-P, "Basis for GE RTNDT Estimation Method,"
Report for BWR Owners' Group, San Jose, California, September 1994 (GE Proprietary). GE Nuclear Energy
GE-NE-0000-0003-5526-02a
- 11. Letter from B. Sheron to R.A. Pinelli,"Safety Assessment of Report NEDC-32399-P, Basis for GE RTNDT Estimation Method, September 1994, " USNRC, December 16, 1994.
LaSalle Unit 1 -QA Records & RPV CMTR's LaSalle Unit 1 GE PO# 205-AK104, Manufactured by CE.
- 13. Letter from R. M. Krich to the NRC, "Response to Request for Additional Information Regarding Reactor Pressure Vessel Integrity - Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 - LaSalle County Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50 373 and 50-374 - Quad Cities Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50 265," Commonwealth Edison Company, Downers Grove, IL., July 30, 1998.
- 14. a) Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation," GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Rev. 0)(GE Proprietary Information).
b) Letter, S.A. Richards, USNRC to J.F. Klapproth, GE-NE, "Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)", MFN 01-050, September 14, 2001.
- 15. "PVRC Recommendations on Toughness Requirements for Ferritic Materials,"
Welding Research Council Bulletin 175, August 1972.
- 16.
- 17. "Analysis of Flaws," Appendix A to Section Xl of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996. GE Nuclear Energy
GE-NE-0000-0003-5526-02a
- 18.
- 19. Bottom Head and Feedwater Nozzle Dimensions:
a) CE Drawing # E232-842, Rev. 2, "Bottom Head Machining and Welding for 251" ID BWR," (GE VPF # 2029-107, Rev. 4).
b) CE Drawing # E-232-863, Rev. 4, "Nozzle Details for 251" ID BWR,"
(GE VPF 2029-099, Rev. 7).
- 20.
- 21. "Materials - Properties," Part D to Section II of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.
- 22. B.J. Branlund, "Plant LaSalle Units I and 2 RPV Shell Thickness Transition and Other Geometric Discontinuities", (GE-NE-B1301869-01), June 1998. GE Nuclear Energy
GE-NE-0000-0003-5526-0 2a GE Nuclear Energy APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1
GE-NE-0000-0003-5526-02a A-2 GE Nuclear Energy
GE-NE-0000-0003-5526-02a Table A Geometric Discontinuities Not Requiring Fracture Toughness Evaluations Per ASME Code Appendix G, Section G2223 (c), fracture toughness analysis to demonstrate protection against non-ductile failure is not required for portions of nozzles and appurtenances having a thickness of 2.5" or less provided the lowest service temperature is not lower than RTNDT plus 600F. Nozzles and appurtenances made from Alloy 600 (Inconel) do not require fracture toughness analysis. Components that do not require a fracture toughness evaluation are listed below:
Nozzle or Appurtenance Nozzle or Appurtenance Material Reference Remarks Identification 317-01 Core Differential Pressure SB 166 1.5.12 &
Thickness is < 2.5" and made
& Liquid Poison -
1.6 of Alloy 600; therefore, no Penetration < 2.5" further fracture toughness Bottom Head evaluation is required.
315-14 Drain-Penetration < 2.5" SA-508 Cl. 1 1.5.1, The discontinuity of the CRD
- Bottom Head 1.5.15 &
nozzle listed in Table A-1 1.6 bounds this discontinuity;,
therefore, no further fracture toughness evaluation is required.
321-05 Seal Leak Detection* -
1.5.1 Not a pressure boundary Penetration -1" component; therefore, Flange requires no fracture toughness evaluation.
319-06 Top Head Lifting Lugs SA-533 GR. B 1.5.1 &
Not a pressure boundary Attachment to Top Head CL. 1 1.5.13, 1.6 component and loads only occur on this component when the reactor is shutdown during an outage. Therefore, no fracture toughness evaluation is required.
- The high/low pressure leak detector, and the seal leak detector are the same nozzle, these nozzles are the closure flange leak detection nozzles.
A-3 GE Nuclear Energy
GE-NE-0000-0003-5526-02a APPENDIX A
REFERENCES:
1.5.
RPV Drawings 1.5.1. CE Drawing # 232-788, Rev. 3, "General Arrangement Elevation for 251" I.D. BWR," (GE VPF #2029-117, Rev. 4).
1.5.2. CE Drawing # 232-790, Rev. 8, "Lower Vessel Shell Assembly Machining & Welding for 251" 1.D. BWR" (GE VPF #2029-036, Rev. 8).
1.5.3. CE Drawing # 232-791, Rev. 15, "Upper Vessel Shell Assembly Machining & Welding for 251" I.D. BWR," (GE VPF #2029-037, Rev. 14).
1.5.4. CE Drawing # 232-792, Rev. 7, "Vessel Machining for 251" 1.D.
BWR," (GE VPF #2029-054, Rev. 8).
1.5.5. CE Drawing # 232-796, Rev. 9", Vessel External Attachments for 251" 1.D. BWR," (GE VPF #2029-085, Rev. 10).
1.5.6. CE Drawing # 232-801, Rev. 0, "Closure Head Final Machining for 251" I.D. BWR," (GE VPF #2029-114, Rev. 2).
1.5.7. CE Drawing # 232-839, Rev. 4, "Closure Head Nozzle Details for 251" I.D. BWR," (GE VPF #2029-108, Rev. 6).
1.5.8. CE Drawing # 232-842, Rev. 2, "Bottom Head Machining & Welding for 251" I.D. BWR," (GE VPF #2029-107, Rev. 4).
1.5.9. CE Drawing # 232-861, Rev. 0, "Vessel Support Skirt Assembly and Details for 251" I.D. BWR," (GE VPF #2029-121, Rev. 2).
1.5.10. CE Drawing # 232-862, Rev. 0, "Bottom Head Penetrations for 251" I.D. BWR," (GE VPF #2029-120, Rev. 2).
1.5.11. CE Drawing # 232-863, Rev. 4, "Nozzle Details for 251" I.D. BWR,"
(GE VPF #2029-099, Rev. 7).
1.5.12. CE Drawing # 232-880, Rev. 1, "Nozzle Details for 251" I.D. BWR,"
(GE VPF #2029-115, Rev. 3).
1.5.13. CE Drawing # 232-911, Rev. 4, "Closure Head Machining & Welding for 251" I.D. BWR," (GE VPF #2029-083, Rev. 6).
1.5.14. CE Drawing # 232-937, Rev. 3, "Shroud Support Details and Assembly for 251" I.D. BWR," (GE VPF #2029-082, Rev. 5).
1.5.15. CE Drawing # 232-938, Rev. 6, "Nozzle Details for 251" 1.D. BWR,"
(GE VPF #2029-084, Rev. 8) 1.6.
CE Stress Report, "Analytical Report for LaSalle County Station Unit 1 for Commonwealth Edison Company," CE Power Systems, Combustion Engineering, Inc, Chattanooga, TN, (Report No CENC-1250.)
A-4 GE Nuclear Energy
GE-NE-0000-0003-5526-02a 1.7.
Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation", GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Revision 0)(GE Proprietary).
A-5 GE Nuclear Energy
GE-NE-0000-0003-5526-02a GE Nuclear Energy APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1
GE-NE-0000-0003-5526-02a TABLE B-1. LaSalle Unit 1 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTILINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(°F)
(OF)
(OF) 0 68.0 72.0 72.0 68.0 72.0 72.0 10 68.0 72.0 72.0 68.0 72.0 72.0 20 68.0 72.0 72.0 68.0 72.0 72.0 30 68.0 72.0 72.0 68.0 72.0 72.0 40 68.0 72.0 72.0 68.0 72.0 72.0 50 68.0 72.0 72.0 68.0 72.0 72.0 60 68.0 72.0 72.0 68.0 72.0 72.0 70 68.0 72.0 72.0 68.0 72.0 72.0 80 68.0 72.0 72.0 68.0 72.0 72.0 90 68.0 72.0 72.0 68.0 72.0 72.0 100 68.0 72.0 72.0 68.0 72.0 72.0 110 68.0 72.0 72.0 68.0 72.0 72.0 120 68.0 72.0 72.0 68.0 72.0 72.0 130 68.0 72.0 72.0 68.0 74.2 72.0 140 68.0 72.0 72.0 68.0 77.4 72.0 150 68.0 72.0 72.0 68.0 80.2 72.0 160 68.0 72.0 72.0 68.0 82.9 73.9 170 68.0 72.0 72.0 68.0 85.5 76.5 180 68.0 72.0 72.0 68.0 87.9 78.9 190 68.0 72.0 72.0 68.0 90.2 81.2 200 68.0 72.0 72.0 68.0 92.3 83.3 210 68.0 72.0 72.0 68.0 94.3 85.3 220 68.0 72.0 72.0 68.0 96.3 87.3 230 68.0 72.0 72.0 68.0 98.1 89.1 240 68.0 72.0 72.0 68.0 99.9 90.9 B-2 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-1. LaSalle Unit I P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(OF)
(OF)
(OF) 250 68.0 72.0 72.0 68.0 101.6 92.6 260 68.0 72.0 72.0 68.0 103.2 94.2 270 68.0 72.0 72.0 68.0 104.8 95.8 280 68.0 72.0 72.0 68.0 106.3 97.3 290 68.0 72.0 72.0 68.0 107.8 98.8 300 68.0 72.0 72.0
.68.0 109.2 100.2 310 68.0 72.0 72.0 68.0 110.5 101.5 312.5 68.0 72.0 72.0 68.0 110.9 101.9 312.5 68.0 102.0 102.0 68.0 132.0 132.0 320 68.0 102.0 102.0 68.0 132.0 132.0 330 68.0 102.0 102.0 68.0 132.0 132.0 340 68.0 102.0 102.0 68.0 132.0 132.0 350 68.0 102.0 102.0 68.0 132.0 132.0 360 68.0 102.0 102.0 68.0 132.0 132.0 370 68.0 102.0 102.0 68.0 132.0 132.0 380 68.0 102.0 102.0 68.0 132.0 132.0 390 68.0 102.0 102.0 68.0 132.0 132.0 400 68.0 102.0 102.0 68.0 132.0 132.0 410 68.0 102.0 102.0 68.0 132.0 132.0 420 68.0 102.0 102.0 68.0 132.0 132.0 430 68.0 102.0 102.0 68.0 132.0 132.0 440 68.0 102.0 102.0 68.0 132.0 132.0 450 68.0 102.0 102.0 68.0 132.0 132.0 460 68.0 102.0 102.0 68.0 132.0 132.0 470 68.0 102.0 102.0 68.0 132.0 132.0 480 68.0 102.0 102.0 68.0 132.0 132.0 490 68.0 102.0 102.0 68.0 132.0 132.0 B-3 GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a TABLE B-1. LaSalle Unit I P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-4, 5-5, 5-6, & 5-8 PRESSURE (PSIG) 500 510 520 530 540 550 560 570 580 590 600 610 620 630 640 650 660 670 680 690 700 710 720 730 740 750 760 BOTTOM HEAD CURVE A (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.7 70.1 71.5 72.8 74.1 75.4 UPPER VESSEL CURVE A (OF) 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.3 103.1 104.0 104.8 32 EFPY BELTLINE CURVE A (OF) 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.8 104.4 106.0 107.5 108.9 110.3 BOTTOM HEAD CURVE B (OF) 68.0 68.0 68.0 68.2 70.1 71.9 73.7 75.4 77.0 78.6 80.2 81.7 83.1 84.5 85.9 87.2 88.5 89.8 91.1 92.3 93.4 94.6 95.7 96.8 97.9 99.0 100.0 UPPER VESSEL CURVE B (OF) 132.0 132.0 132.2 133.0 133.8 134.6 135.4 136.1 136.9 137.6 138.1 138.6 139.0 139.4 139.8 140.2 140.7 141.1 141.5 141.9 142.3 142.7 143.1 143.5 143.9 144.2 144.6 32 EFPY BELTLINE CURVE B
(°F) 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.9 134.0 135.2 136.3 137.4 138.5 139.5 140.5 141.5 142.5 143.5 144.4 145.4 146.3 147.2 148.1 148.9 149.8 B-4
GE-NE-0000-0003-5526-02a TABLE B-1. LaSalle Unit I P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(OF)
(°F)
(OF) 770 76.6 105.6 111.7 101.0 145.0 150.6 780 77.8 106.3 113.0 102.0 145.4 151.4 790 79.0 107.1 114.4 103.0 145.8 152.2 800 80.2 107.9 115.6 103.9 146.1 153.0 810 81.3 108.6 116.9 104.9 146.5 153.8 820 82.4 109.4 118.1 105.8 146.9 154.6 830 83.5 110.1 119.3 106.7 147.2 155.4 840 84.5 110.8 120.4 107.6 147.6 156.1 850 85.6 111.5 121.5 108.4 147.9 156.9 860 86.6 112.2 122.6 109.3 148.3 157.6 870 87.6 112.9 123.7 110.1 148.6 158.3 880 88.5 113.6 124.8 111.0 149.0 159.0 890 89.5 114.3 125.8 111.8 149.3 159.7 900 90.4 114.9 126.8 112.6 149.7 160.4 910 91.4 115.6 127.8 113.4 150.0 161.1 920 92.3 116.2 128.8 114.1 150.4 161.7 930 93.1 116.9 129.7 114.9 150.7 162.4 940 94.0 117.5 130.7 115.7 151.0 163.0 950 94.9 118.1 131.6 116.4 151.4 163.7 960 95.7 118.7 132.5 117.1 151.7 164.3 970 96.6 119.3 133.4 117.9 152.0 165.0 980 97.4 119.9 134.3 118.6 152.4 165.6 990 98.2 120.5 135.1 119.3 152.7 166.2 1000 99.0 121.1 136.0 120.0 153.0 166.8 1010 99.7 121.7 136.8 120.6 153.3 167.4 1020 100.5 122.2 137.6 121.3 153.6 168.0 1030 101.3 122.8 138.4 122.0 154.0 168.6 B-5 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-1. LaSalle Unit 1 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2. 5-3, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(OF)
(OF)
(OF) 1040 102.0 123.4 139.2 122.6 154.3 169.1 1050 102.7 123.9 140.0 123.3 154.6 169.7 1060 103.4 124.5 140.7 123.9 154.9 170.3 1070 104.2 125.0 141.5 124.5 155.2 170.8 1080 104.9 125.5 142.2 125.2 155.5 171.4 1090 105.6 126.1 143.0 125.8 155.8 171.9 1100 106.2 126.6 143.7 126.4 156.1 172.5 1105 106.6 126.8 144.0 126.7 156.3 172.7 1110 106.9 127.1 144.4 127.0 156.4 173.0 1120 107.6 127.6 145.1 127.6 156.7 173.5 1130 108.2 128.1 145.8 128.2 157.0 174.1 1140 108.9 128.6 146.5 128.7 157.3 174.6 1150 109.5 129.1 147.1 129.3 157.6 175.1 1160 110.1 129.6 147.8 129.9 157.9 175.6 1170 110.8 130.1 148.4 130.4 158.2 176.1 1180 111.4 130.6 149.1 131.0 158.5 176.6 1190 112.0 131.1 149.7 131.5 158.7 177.1 1200 112.6 131.5 150.4 132.1 159.0 177.6 1210 113.2 132.0 151.0 132.6 159.3 178.0 1220 113.8 132.5 151.6 133.2 159.6 178.5 1230 114.3 132.9 152.2 133.7 159.9 179.0 1240 114.9 133.4 152.8 134.2 160.2 179.5 1250 115.5 133.8 153.4 134.7 160.4 179.9 1260 116.0 134.3 154.0 135.2 160.7 180.4 1270 116.6 134.7 154.6 135.7 161.0 180.8 1280 117.1 135.2 155.1 136.2 161.2 181.3 1290 117.7 135.6 155.7 136.7 161.5 181.7 B-6 GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-0 2 a TABLE B-1. LaSalle Unit I P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °Flhr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(°F)
(OF)
(OF)
(OF)
(°F) 1300 118.2 136.0 156.3 137.2 161.8 182.2 1310 118.7 136.5 156.8 137.7 162.1 182.6 1320 119.3 136.9 157.4 138.2 162.3 183.1 1330 119.8 137.3 157.9 138.6 162.6 183.5 1340 120.3 137.7 158.4 139.1 162.8 183.9 1350 120.8 138.1 159.0 139.6 163.1 184.3 1360 121.3 138.6 159.5 140.0 163.4 184.8 1370 121.8 139.0 160.0 140.5 163.6 185.2 1380 122.3 139.4 160.5 140.9 163.9 185.6 1390 122.8 139.8 161.0 141.4 164.1 186.0 1400 123.3 140.2 161.5 141.8 164.4 186.4 B-7
GE-NE-0000-0003-5526-02a TABLE B-2. LaSalle Unit 1 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM HEAD CURVE A (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 UPPER RPV &
BELTLINE AT 32 EFPY CURVE A (OF) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 BOTTOM HEAD CURVE B (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 UPPER RPV &
BELTLINE AT 32 EFPY CURVE B (OF) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 74.2 77.4 80.2 82.9 85.5 87.9 90.2 92.3 94.3 96.3 98.1 B-8 NONBELTLINE AND BELTLINE AT 32 EFPY CURVE C
(°F) 72.0 72.0 72.0 72.0 72.0 72.0 80.0 87.2 93.2 98.3 102.8 106.9 110.7 114.2 117.4 120.2 122.9 125.5 127.9 130.2 132.3 134.3 136.3 138.1 PRESSURE (PSIG) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-2. LaSalle Unit I Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 PRESSURE (PSIG) 240 250 260 270 280 290 300 310 312.5 312.5 320 330 340 350 360 370 380 390 400 410 420 430 440 450 460 470 BOTTOM HEAD CURVE A (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 UPPER RPV &
BELTLINE AT 32 EFPY CURVE A (OF) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 BOTTOM HEAD CURVE B (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 UPPER RPV &
BELTLINE AT 32 EFPY CURVE B
(°F) 99.9 101.6 103.2 104.8 106.3 107.8 109.2 110.5 110.9 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 NONBELTLINE AND BELTLINE AT 32 EFPY CURVE C (OF) 139.9 141.6 143.2 144.8 146.3 147.8 149.2 150.5 150.9 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 B-9 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-2. LaSalle Unit 1 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM HEAD CURVE A (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.7 70.1 71.5 UPPER RPV &
BELTLINE AT 32 EFPY CURVE A (OF) 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.8 104.4 106.0 BOTTOM HEAD CURVE B (OF) 68.0 68.0 68.0 68.0 68.0 68.2 70.1 71.9 73.7 75.4 77.0 78.6 80.2 81.7 83.1 84.5 85.9 87.2 88.5 89.8 91.1 92.3 93.4 94.6 95.7 96.8 UPPER RPV &
BELTLINE AT 32 EFPY CURVE B (OF) 132.0 132.0 132.0 132.0 132.2 133.0 133.8 134.6 135.4 136.1 136.9 137.6 138.1 138.6 139.0 139.4 139.8 140.2 140.7 141.5 142.5 143.5 144.4 145.4 146.3 147.2 B-10 NONBELTLINE AND BELTLINE AT 32 EFPY CURVE C (OF) 172.0 172.0 172.0 172.0 172.2 173.0 173.8 174.6 175.4 176.1 176.9 177.6 178.1 178.6 179.0 179.4 179.8 180.2 180.7 181.5 182.5 183.5 184.4 185.4 186.3 187.2 PRESSURE (PSIG) 480 490 500 510 520 530 540 550 560 570 580 590 600 610 620 630 640 650 660 670 680 690 700 710 720 730 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-2. LaSalle Unit I Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 PRESSURE (PSIG) 740 750 760 770 780 790 800 810 820 830 840 850 860 870 880 890 900 910 920 930 940 950 960 970 980 990 BOTTOM HEAD CURVE A (OF) 72.8 74.1 75.4 76.6 77.8 79.0 80.2 81.3 82.4 83.5 84.5 85.6 86.6 87.6 88.5 89.5 90.4 91.4 92.3 93.1 94.0 94.9 95.7 96.6 97.4 98.2 UPPER RPV &
BELTLINE AT 32 EFPY CURVE A
(°F) 107.5 108.9 110.3 111.7 113.0 114.4 115.6 116.9 118.1 119.3 120.4 121.5 122.6 123.7 124.8 125.8 126.8 127.8 128.8 129.7 130.7 131.6 132.5 133.4 134.3 135.1 BOTTOM HEAD CURVE B (OF) 97.9 99.0 100.0 101.0 102.0 103.0 103.9 104.9 105.8 106.7 107.6 108.4 109.3 110.1 111.0 111.8 112.6 113.4 114.1 114.9 115.7 116.4 117.1 117.9 118.6 119.3 UPPER RPV &
BELTLINE AT 32 EFPY CURVE B (OF) 148.1 148.9 149.8 150.6 151.4 152.2 153.0 153.8 154.6 155.4 156.1 156.9 157.6 158.3 159.0 159.7 160.4 161.1 161.7 162.4 163.0 163.7 164.3 165.0 165.6 166.2 NONBELTLINE AND BELTINE AT 32 EFPY CURVE C
('F) 188.1 188.9 189.8 190.6 191.4 192.2 193.0 193.8 194.6 195.4 196.1 196.9 197.6 198.3 199.0 199.7 200.4 201.1 201.7 202.4 203.0 203.7 204.3 205.0 205.6 206.2 B-11 GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a TABLE B-2. LaSalle Unit 1 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV &
BOTTOM UPPER RPV &
NONBELTLINE HEAD BELTLINE AT HEAD BELTLINE AT AND BELTLINE 32 EFPY 32 EFPY AT 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG)
(OF)
(OF)
(OF)
(OF)
(°F) 1000 99.0 136.0 120.0 166.8 206.8 1010 99.7 136.8 120.6 167.4 207.4 1020 100.5 137.6 121.3 168.0 208.0 1030 101.3 138.4 122.0 168.6 208.6 1040 102.0 139.2 122.6 169.1 209.1 1050 102.7 140.0 123.3 169.7 209.7 1060 103.4 140.7 123.9 170.3 210.3 1070 104.2 141.5 124.5 170.8 210.8 1080 104.9 142.2 125.2 171.4 211.4 1090 105.6 143.0 125.8 171.9 211.9 1100 106.2 143.7 126.4 172.5 212.5 1105 106.6 144.0 126.7 172.7 212.7 1110 106.9 144.4 127.0 173.0 213.0 1120 107.6 145.1 127.6 173.5 213.5 1130 108.2 145.8 128.2 174.1 214.1 1140 108.9 146.5 128.7 174.6 214.6 1150 109.5 147.1 129.3 175.1 215.1 1160 110.1 147.8 129.9 175.6 215.6 1170 110.8 148.4 130.4 176.1 216.1 1180 111.4 149.1 131.0 176.6 216.6 1190 112.0 149.7 131.5 177.1 217.1 1200 112.6 150.4 132.1 177.6 217.6 1210 113.2 151.0 132.6 178.0 218.0 1220 113.8 151.6 133.2 178.5 218.5 1230 114.3 152.2 133.7 179.0 219.0 1240 114.9 152.8 134.2 179.5 219.5 B-1 2
GE-NE-0000-0003-5526-02a TABLE B-2. LaSalle Unit I Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 UPPER RPV &
BELTLINE AT 32 EFPY CURVE A (OF) 153.4 154.0 154.6 155.1 155.7 156.3 156.8 157.4 157.9 158.4 159.0 159.5 160.0 160.5 161.0 161.5 BOTTOM HEAD CURVE B (OF)
BOTTOM HEAD CURVE A (OF) 115.5 116.0 116.6 117.1 117.7 118.2 118.7 119.3 119.8 120.3 120.8 121.3 121.8 122.3 122.8 123.3 UPPER RPV &
BELTLINE AT 32 EFPY CURVE B (OF) 179.9 180.4 180.8 181.3 181.7 182.2 182.6 183.1 183.5 183.9 184.3 184.8 185.2 185.6 186.0 186.4 NONBELTLINE AND BELTLINE AT 32 EFPY CURVE C
(°F) 219.9 220.4 220.8 221.3 221.7 222.2 222.6 223.1 223.5 223.9 224.3 224.8 225.2 225.6 226.0 226.4 B-13 134.7 135.2 135.7 136.2 136.7 137.2 137.7 138.2 138.6 139.1 139.6 140.0 140.5 140.9 141.4 141.8 PRESSURE (PSIG) 1250 1260 1270 1280 1290 1300 1310 1320 1330 1340 1350 1360 1370 1380 1390 1400 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-3. LaSalle Unit I P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6. & 5-7 PRESSURE (PSIG) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 BOTTOM HEAD CURVE A
(°F) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 UPPER VESSEL CURVE A (OF) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 20 EFPY BELTLINE CURVE A (OF) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 BOTTOM HEAD CURVE B (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 UPPER VESSEL CURVE B (OF) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 74.2 77.4 80.2 82.9 85.5 87.9 90.2 92.3 94.3 96.3 98.1 99.9 20 EFPY BELTLINE CURVE B
(°F) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.9 75.2 77.3 79.3 81.3 83.1 84.9 B-14 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-3. LaSalle Unit I P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 PRESSURE (PSIG) 250 260 270 280 290 300 310 312.5 312.5 320 330 340 350 360 370 380 390 400 410 420 430 440 450 460 470 480 490 BOTTOM HEAD CURVE A
(°F) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 UPPER VESSEL CURVE A
(°F) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 20 EFPY BELTLINE CURVE A
(°F) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 BOTTOM HEAD CURVE B (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 UPPER VESSEL CURVE B (OF) 101.6 103.2 104.8 106.3 107.8 109.2 110.5 110.9 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 20 EFPY BELTLINE CURVE B (OF) 86.6 88.2 89.8 91.3 92.8 94.2 95.5 95.9 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 B-1 5 GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a TABLE B-3. LaSalle Unit I P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 PRESSURE (PSIG) 500 510 520 530 540 550 560 570 580 590 600 610 620 630 640 650 660 670 680 690 700 710 720 730 740 750 760 BOTTOM HEAD CURVE A (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.7 70.1 71.5 72.8 74.1 75.4 UPPER VESSEL CURVE A (OF) 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.3 103.1 104.0 104.8 20 EFPY BELTLINE CURVE A (OF) 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 102.0 BOTTOM HEAD CURVE B (OF) 68.0 68.0 68.0 68.2 70.1 71.9 73.7 75.4 77.0 78.6 80.2 81.7 83.1 84.5 85.9 87.2 88.5 89.8 91.1 92.3 93.4 94.6 95.7 96.8 97.9 99.0 100.0 UPPER VESSEL CURVE B (F) 132.0 132.0 132.2 133.0 133.8 134.6 135.4 136.1 136.9 137.6 138.1 138.6 139.0 139.4 139.8 140.2 140.7 141.1 141.5 141.9 142.3 142.7 143.1 143.5 143.9 144.2 144.6 20 EFPY BELTLINE CURVE B (OF) 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.0 132.1 132.9 133.8 B-16
GE-NE-0000-0003-5526-02a TABLE B-3. LaSalle Unit I P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM UPPER 20 EFPY BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(°F)
(OF)
(OF) 770 76.6 105.6 102.0 101.0 145.0 134.6 780 77.8 106.3 102.0 102.0 145.4 135.4 790 79.0 107.1 102.0 103.0 145.8 136.2 800 80.2 107.9 102.0 103.9 146.1 137.0 810 81.3 108.6 102.0 104.9 146.5 137.8 820 82.4 109.4 102.1 105.8 146.9 138.6 830 83.5 110.1 103.3 106.7 147.2 139.4 840 84.5 110.8 104.4 107.6 147.6 140.1 850 85.6 111.5 105.5 108.4 147.9 140.9 860 86.6 112.2 106.6 109.3 148.3 141.6 870 87.6 112.9 107.7 110.1 148.6 142.3 S880 88.5 113.6 108.8 111.0 149.0 143.0 890 89.5 114.3 109.8 111.8 149.3 143.7 900 90.4 114.9 110.8 112.6 149.7 144.4 910 91.4 115.6 111.8 113.4 150.0 145.1 920 92.3 116.2 112.8 114.1 150.4 145.7 "930 93.1 116.9 113.7 114.9 150.7 146.4 940 94.0 117.5 114.7 115.7 151.0 147.0 950 94.9 118.1 115.6 116.4 151.4 147.7 960 95.7 118.7 116.5 117.1 151.7 148.3 970 96.6 119.3 117.4 117.9 152.0 149.0 980 97.4 119.9 118.3 118.6 152.4 149.6 990 98.2 120.5 119.1 119.3 152.7 150.2 "1000 99.0 121.1 120.0 120.0 153.0 150.8 1010 99.7 121.7 120.8 120.6 153.3 151.4 1020 100.5 122.2 121.6 121.3 153.6 152.0 1030 101.3 122.8 122.4 122.0 154.0 152.6 B-17 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-3. LaSalle Unit 1 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1. 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM UPPER 20 EFPY BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(°F)
(OF)
(°F)
(°F)
(OF) 1040 102.0 123.4 123.2 122.6 154.3 153.1 1050 102.7 123.9 124.0 123.3 154.6 153.7 1060 103.4 124.5 124.7 123.9 154.9 154.3 1070 104.2 125.0 125.5 124.5 155.2 154.8 1080 104.9 125.5 126.2 125.2 155.5 155.4 1090 105.6 126.1 127.0 125.8 155.8 155.9 1100 106.2 126.6 127.7 126.4 156.1 156.5 1105 106.6 126.8 128.0 126.7 156.3 156.7 1110 106.9 127.1 128.4 127.0 156.4 157.0 1120 107.6 127.6 129.1 127.6 156.7 157.5 1130 108.2 128.1 129.8 128.2 157.0 158.1 1140 108.9 128.6 130.5 128.7 157.3 158.6 1150 109.5 129.1 131.1 129.3 157.6 159.1 1160 110.1 129.6 131.8 129.9 157.9 159.6 1170 110.8 130.1 132.4 130.4 158.2 160.1 1180 111.4 130.6 133.1 131.0 158.5 160.6 1190 112.0 131.1 133.7 131.5 158.7 161.1 1200 112.6 131.5 134.4 132.1 159.0 161.6 1210 113.2 132.0 135.0 132.6 159.3 162.0 1220 113.8 132.5 135.6 133.2 159.6 162.5 1230 114.3 132.9 136.2 133.7 159.9 163.0 1240 114.9 133.4 136.8 134.2 160.2 163.5 1250 115.5 133.8 137.4 134.7 160.4 163.9 1260 116.0 134.3 138.0 135.2 160.7 164.4 1270 116.6 134.7 138.6 135.7 161.0 164.8 1280 117.1 135.2 139.1 136.2 161.2 165.3 1290 117.7 135.6 139.7 136.7 161.5 165.7 B-18 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-3. LaSalle Unit I P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6. & 5-7 BOTTOM UPPER 20 EFPY BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(°F)
(OF)
(OF)
(°F)
(OF)
(OF) 1300 118.2 136.0 140.3 137.2 161.8 166.2 1310 118.7 136.5 140.8 137.7 162.1 166.6 1320 119.3 136.9 141.4 138.2 162.3 167.1 1330 119.8 137.3 141.9 138.6 162.6 167.5 1340 120.3 137.7 142.4 139.1 162.8 167.9 1350 120.8 138.1 143.0 139.6 163.1 168.3 1360 121.3 138.6 143.5 140.0 163.4 168.8 1370 121.8 139.0 144.0 140.5 163.6 169.2 1380 122.3 139.4 144.5 140.9 163.9 169.6 1390 122.8 139.8 145.0 141.4 164.1 170.0 1400 123.3 140.2 145.5 141.8 164.4 170.4 B-1 9 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-4. LaSalle Unit 1 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 UPPER PRESSURE VESSEL CURVE C (PSIG)
(OF) 0 72.0 10 72.0 20 72.0 30 72.0 40 72.0 50 72.0 60 80.0 70 87.2 80 93.2 90 98.3 100 102.8 110 106.9 120 110.7 130 114.2 140 117.4 150 120.2 160 122.9 170 125.5 180 127.9 190 130.2 200 132.3 210 134.3 220 136.3 230 138.1 240 139.9 250 141.6 260 143.2 BOTTOM HEAD CURVE C (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 20 EFPY BELTLINE CURVE C (OF) 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 72.0 75.0 80.7 B-20 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-4. LaSalle Unit I P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 PRESSURE (PSIG) 270 280 290 300 310 312.5 312.5 320 330 340 350 360 370 380 390 400 410 420 430 440 450 460 470 480 490 500 510 520 UPPER VESSEL CURVE C
(°F) 144.8 146.3 147.8 149.2 150.5 150.9 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.2 BOTTOM HEAD CURVE C
(°F) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 69.3 73.3 77.0 80.5 83.8 86.8 89.7 92.4 95.0 97.5 99.8 102.0 104.2 106.2 B-21 20 EFPY BELTLINE CURVE C (OF) 85.9 90.5 94.8 98.7 102.4 103.3 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-4. LaSalle Unit 1 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 PRESSURE (PSIG)
UPPER VESSEL CURVE C (OF) 530 540 550 560 570 580 590 600 610 620 630 640 650 660 670 680 690 700 710 720 730 740 750 760 770 780 790 800 BOTTOM HEAD CURVE C (OF) 108.2 110.1 111.9 113.7 115.4 117.0 118.6 120.2 121.7 123.1 124.5 125.9 127.2 128.5 129.8 131.1 132.3 133.4 134.6 135.7 136.8 137.9 139.0 140.0 141.0 142.0 143.0 143.9 20 EFPY BELTLINE CURVE C (OF) 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.0 172.1 172.9 173.8 174.6 175.4 176.2 177.0 B-22 173.0 173.8 174.6 175.4 176.1 176.9 177.6 178.1 178.6 179.0 179.4 179.8 180.2 180.7 181.1 181.5 181.9 182.3 182.7 183.1 183.5 183.9 184.2 184.6 185.0 185.4 185.8 186.1 GE Nuclear Energy
-W
GE-NE-0000-0003-5526-02a TABLE B-4. LaSalle Unit 1 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 PRESSURE (PSIG) 810 820 830 840 850 860 870 880 890 900 910 920 930 940 950 960 970 980 990 1000 1010 1020 1030 1040 1050 1060 1070 1080 UPPER VESSEL CURVE C (OF) 186.5 186.9 187.2 187.6 187.9 188.3 188.6 189.0 189.3 189.7 190.0 190.4 190.7 191.0 191.4 191.7 192.0 192.4 192.7 193.0 193.3 193.6 194.0 194.3 194.6 194.9 195.2 195.5 BOTTOM HEAD CURVE C (OF) 144.9 145.8 146.7 147.6 148.4 149.3 150.1 151.0 151.8 152.6 153.4 154.1 154.9 155.7 156.4 157.1 157.9 158.6 159.3 160.0 160.6 161.3 162.0 162.6 163.3 163.9 164.5 165.2 B-23 20 EFPY BELTLINE CURVE C (OF) 177.8 178.6 179.4 180.1 180.9 181.6 182.3 183.0 183.7 184.4 185.1 185.7 186.4 187.0 187.7 188.3 189.0 189.6 190.2 190.8 191.4 192.0 192.6 193.1 193.7 194.3 194.8 195.4 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-4. LaSalle Unit I P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 PRESSURE (PSIG) 1090 1100 1105 1110 1120 1130 1140 1150 1160 1170 1180 1190 1200 1210 1220 1230 1240 1250 1260 1270 1280 1290 1300 1310 1320 1330 1340 1350 UPPER VESSEL CURVE C (OF) 195.8 196.1 196.3 196.4 196.7 197.0 197.3 197.6 197.9 198.2 198.5 198.7 199.0 199.3 199.6 199.9 200.2 200.4 200.7 201.0 201.2 201.5 201.8 202.1 202.3 202.6 202.8 203.1 BOTTOM HEAD CURVE C (OF) 165.8 166.4 166.7 167.0 167.6 168.2 168.7 169.3 169.9 170.4 171.0 171.5 172.1 172.6 173.2 173.7 174.2 174.7 175.2 175.7 176.2 176.7 177.2 177.7 178.2 178.6 179.1 179.6 20 EFPY BELTLINE CURVE C
(°F) 195.9 196.5 196.7 197.0 197.5 198.1 198.6 199.1 199.6 200.1 200.6 201.1 201.6 202.0 202.5 203.0 203.5 203.9 204.4 204.8 205.3 205.7 206.2 206.6 207.1 207.5 207.9 208.3 B-24 GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE B-4. LaSalle Unit 1 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 UPPER PRESSURE VESSEL CURVE C (PSIG)
(°F) 1360 203.4 1370 203.6 1380 203.9 1390 204.1 1400 204.4 BO'-rOM HEAD CURVE C (OF) 180.0 180.5 180.9 181.4 181.8 20 EFPY BELTLINE CURVE C (OF) 208.8 209.2 209.6 210.0 210.4 B-25 GE Nuclear Energy
GE-NE-0000-0003-5526-02a APPENDIX C Operating And Temperature Monitoring Requirements C-1 GE Nuclear Energy
GE-NE-0000-0003-5526-02a C.1 NON-BELTLINE MONITORING DURING PRESSURE TESTS It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the recirculation pumps are operating at low speed, or are off, and injection through the control rod drives is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.
One condition on monitoring the bottom head separately is that it must be demonstrated that the vessel beltline temperature can be accurately monitored during pressure testing. An experiment has been conducted at a BWR-4 that showed that thermocouples on the vessel near the feedwater nozzles, or temperature measurements of water in the recirculation loops provide good estimates of the beltine temperature during pressure testing. Thermocouples on the RPV flange to shell junction outside surface should be used to monitor compliance with upper vessel curve. Thermocouples on the bottom head outside surface should be used to monitor compliance with bottom head curves. A description of these measurements is given in GE SIL 430, attached in Appendix D. First, however, it should be determined whether there are significant temperature differences between the beltline region and the bottom head region.
C.2 DETERMINING WHICH CURVE TO FOLLOW The following subsections outline the criteria needed for determining which curve is governing during different situations. The application of the P-T curves and some of the assumptions inherent in the curves to plant operation is dependent on the proper monitoring of vessel temperatures.
C-2 GE Nuclear Energy
GE-NE-0000-0003-5526-02a C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the coolant temperature is changing by *20 0F per hour. If the coolant is experiencing a higher heating or cooling rate in preparation for or following a pressure test, Curve B applies.
C.2.2 Curve B: Non-Nuclear HeatuplCooldown Curve B should be used whenever Curve A or Curve C do not apply. In other words, the operator must follow this curve during times when the coolant is heating or cooling faster than 20°F per hour during a hydrotest and when the core is not critical.
C.2.3 Curve C: Core Critical Operation The operator must comply with this curve whenever the core is critical. An exception to this principle is for low-level physics tests; Curve B must be followed during these situations.
C.3 REACTOR OPERATION VERSUS OPERATING LIMITS For most reactor operating conditions, coolant pressure and temperature are at saturation conditions, which are well into the acceptable operating area (to the right of the P-T curves). The operations where P-T curve compliance is typically monitored closely are planned events, such as vessel boltup, leakage testing and startup/shutdown operations, where operator actions can directly influence vessel pressures and temperatures.
The most severe unplanned transients relative to the P-T curves are those that result from SCRAMs, which sometimes include recirculation pump trips. Depending on operator responses following pump trip, there can be cases where stratification of colder water in the bottom head occurs while the vessel pressure is still relatively high.
Experience with such events has shown that operator action is necessary to avoid P-T curve exceedance, but there is adequate time for operators to respond.
C-3 GE Nuclear Energy
GE-NE-0000-0003-5526-02a In summary, there are several operating conditions where careful monitoring of P-T conditions against the curves is needed:
"* Head flange boltup
"* Leakage test (Curve A compliance)
"* Startup (coolant temperature change of less than or equal to 100°F in one hour period heatup)
"* Shutdown (coolant temperature change of less than or equal to 100OF in one hour period cooldown)
"* Recirculation pump trip, bottom head stratification (Curve B compliance)
C-4 GE Nuclear Energy
GE-NE-0000-0003-5526-02a GE Nuclear Energy APPENDIX D GE SIL 430 D-1
GE-NE-0000-0003-5526-02a September 27, 1985 SIL No. 430 REACTOR PRESSURE VESSEL TEMPERATURE MONITORING Recently, several BWR owners with plants in initial startup have had questions concerning primary and alternate reactor pressure vessel (RPV) temperature monitoring measurements for complying with RPV brittle fracture and thermal stress requirements.
As such, the purpose of this Service Information Letter is to provide a summary of RPV temperature monitoring measurements, their primary and alternate uses and their limitations (See the attached table). Of basic concern is temperature monitoring to comply with brittle fracture temperature limits and for vessel thermal stresses during RPV heatup and cooldown. General Electric recommends that BWR owners/operators review this table against their current practices and evaluate any inconsistencies.
TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (Typical)
Measurement Use Limitations Steam dome saturation temperature as determined from main steam instrument line pressure Recirc suction line coolant temperature.
Primary measurement above 212°F for Tech Spec I OOOF/hr heatup and cooldown rate.
Primary measurement below 212°F for Tech Spec 100°F/hr heatup and cooldown rate.
Alternate measurement above 2120F.
Must convert saturated steam pressure to temperature.
Must have recirc flow.
Must comply with SIL 251 to avoid vessel stratification.
When above 212°F need to allow for temperature variations (up to 10-15oF lower than steam dome saturation temperature) caused primarily by FW flow variations.
D-2
-. d GE Nuclear Energy
GE-NE-0000-0003-5526-02a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)
(Typical)
Measurement RHR heat exchanger inlet coolant temperature RPV drain line coolant temperature Use Alternate measurement for RPV drain line temperature (can use to comply with delta T limit between steam dome saturation temperature and bottom head drain line temperature).
Alternate measurement for Tech Spec IOOOF/hr cooldown rate when in shutdown cooling mode.
Primary measurement to comply with Tech Spec delta T limit between steam dome saturated temp and drain line coolant temperature.
Primary measurement to comply with Tech Spec brittle fracture limits during cooldown.
Alternate information only measurement for bottom head inside/
outside metal surface temperatures.
Limitations Must have previously correlated RHR inlet coolant temperature versus RPV coolant temperature.
Must have drain line flow. Otherwise, lower than actual temperature and higher delta T's will be indicated Delta T limit is 1000F for BWR/6s and 145°F for earlier BWRs.
Must have drain line flow. Use to verify compliance with Tech Spec minimum metal temperature/reactor pressure curves (using drain line temperature to represent bottom head metal temperature).
Must compensate for outside metal temperature lag during heatup/cooldown.
Should have drain line flow.
D-3 GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)
(Typical)
Measurement Closure head flanges outside surface T/Cs RPV flange-to-shell junction outside surface T/Cs RPV shell outside surface T/Cs Top head outside surface T/Cs Use Primary measurement for BWR/6s to comply with Tech Spec brittle fracture metal temperature limit for head boltup.
One of two primary measure ments for BWR/6s for hydro test.
Primary measurement for BWRs earlier than 6s to comply with Tech Spec brittle fracture metal temperature limit for head boltup.
One of two primary measurements for BWRs earlier than 6s for hydro test. Preferred in lieu of closure head flange T/Cs if available.
Information only.
Information only.
Limitations Use for metal (not coolant) temperature. Install temporary T/Cs for alternate measurement, if required.
Use for metal (not coolant) temperature. Response faster than closure head flange T/Cs.
Use RPV closure head flange outside surface as alternate measurement.
Slow to respond to RPV coolant changes. Not available on BWR/6s.
Very slow to respond to RPV coolant changes. Not avail able on BWR/6s.
D-4
GE-NE-0000-0003-5526-02a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)
(Typical)
Measurement Bottom head outside surface T/Cs Use 1 of 2 primary measurements to comply with Tech Spec brittle fracture metal temperature limit for hydro test.
Primary measurement to comply with Tech Spec brittle fracture metal temperature limits during heatup.
Limitations Should verify that vessel stratification is not present for vessel hydro.
(see SEL No. 251).
Use during heatup to verify compliance with Tech Spec metal temperature/reactor pressure curves.
Note: RPV vendor specified metal T limits for vessel heatup and cooldown should be checked during initial plant startup tests when initial RPV vessel heatup and cooldown tests are run.
D-5 GE Nuclear Energy
GE-NE-0000-0003-5526-02a Product
Reference:
B21 Nuclear Boiler Prepared By: A.C. Tsang Approved for Issue:
Issued By:
B.H. Eldridge, Mgr.
D.L. Alired, Manager Service Information Customer Service Information and Analysis Notice:
SILs pertain only to GE BWRs. GE prepares SILs exclusively as a service to owners of GE BWRs. GE does not consider or evaluate the applicability, if any, of information contained in SILs to any plant or facility other than GE BWRs as designed and furnished by GE. Determination of applicability of information contained in any SIL to a specific GE BWR and implementation of recommended action are responsibilities of the owner of that GE BWR.SILs are part of GE s continuing service to GE BWR owners. Each GE BWR is operated by and is under the control of its owner. Such operation involves activities of which GE has no knowledge and over which GE has no control. Therefore, GE makes no warranty or representation expressed or implied with respect to the accuracy, completeness or usefulness of information contained in SILs. GE assumes no responsibility for liability or damage, which may result from the use of information contained in SILs.
D-6 GE Nuclear Energy
GE-NE-0000-0003-5526-02a APPENDIX E Determination of Beitline Region and Impact on Fracture Toughness E-1 GE Nuclear Energy
GE-NE-0000-0003-5526-02a 10CFR50, Appendix G defines the beltline region of the reactor vessel as follows:
"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 radiation damage" To establish the value of peak fluence for identification of beltline materials (as discussed above), the 1 OCFR50 Appendix H fluence value used to determine the need for a surveillance program was used; the value specified is a peak fluence (E>1 MEV) of 1.0e17 n/cm2. Therefore, if it can be shown that no nozzles are located where the peak neutron fluence is expected to exceed or equal 1.0e17 n/cm 2, then it can be concluded that all reactor vessel nozzles are outside the beltline region of the reactor vessel, and do not need to be considered in the P-T curve evaluation.
The following dimensions are obtained from the referenced drawings:
Shell # 3 - Top of Active Fuel (TAF): 366.31" (from vessel 0) [1]
Shell # 1 - Bottom of Active Fuel (BAF): 216.31" (from vessel 0) [1]
Bottom of LPCI Nozzle in Shell # 3: 355.6" (from vessel 0) [2]
Center line of LPCI Nozzle in Shell # 3: 372.5" (from vessel 0) [3]
Top of Recirculation Outlet Nozzle in Shell # 1: 197.188" (from vessel 0) [4]
Center line of Recirculation Outlet Nozzle in Shell # 1: 172.5" (from vessel 0) [3]
Top of Recirculation Inlet Nozzle in Shell # 1: 197.688" (from vessel 0) [4]
Center line of Recirculation Inlet Nozzle in Shell # 1: 181" (from vessel 0) [3]
As shown above, the LPCI nozzle is within the core beltline region. This nozzle is bounded by the feedwater pressure-temperature curve as stated in Appendix A.
From [3], it is obvious that the recirculation inlet and outlet nozzles are closest to the beltline region (the top of the recirculation inlet nozzle is -18" from BAF and the top of the recirculation outlet nozzle is -19" from BAF), and no other nozzles are within the BAF-TAF region of the reactor vessel. Therefore, if it can be shown that the peak E-2 GE Nuclear Energy
GE-NE-0000-0003-5526-02a fluence at this location is less than 1.0e17 n/cm2, it can be safely concluded that all nozzles (other than the LPCI nozzle) are outside the beltline region of the reactor vessel.
Based on the axial flux profile [5], the RPV flux level at -10" below the BAF dropped to less than 0.1 of the peak flux at the same radius. Likewise, the RPV flux level at -10" above the TAF dropped to less than 0.1 of the peak flux at the same radius. Therefore, if the RPV fluence is 1.02e18 n/cm 2 [5], fluence at -10" below BAF and -10" above TAF are expected to be less than 1.00e7 n/cm2 at 32 EFPY. The beltline region considered in the development of the P-T curves is adjusted to include the additional 10" above and below the active fuel region. The adjusted beltline region extends from 206.31" to 376.31" above reactor vessel "0".
Based on the above, it is concluded that none of the LaSalle Unit I reactor vessel nozzles, other than the LPCI nozzle which is considered in the P-T curve evaluation, are in the beltline region.
E-3 GE Nuclear Energy
GE-NE-0000-0003-5526-02a APPENDIX E
REFERENCES:
- 1.
ComEd Nuclear Design Information Transmittal (NDIT) No. LS-1 169, "Pressure-Temperature Curves", 12/10/99.
- 2.
CE Drawing #232-863, Revision 4, "Nozzle Details for 251" I.D. BWR", (GE VPF #2029-099, Revision 7).
- 3.
CE Drawing #232-788, Revision 3, "General Arrangement Elevation for 251" I.D. BWR" (GE VPF #2029-117, Revision 4).
- 4.
CE Drawing #232-879, Revision 3, "Nozzle Details for 251" I.D. BWR", (GE VPF #2029-092, Revision 6).
- 5.
Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation", GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Rev. 0)(GE Proprietary Information).
E-4 GE Nuclear Energy
GE-NE-0000-0003-5526-02a APPENDIX F EVALUATION FOR UPPER SHELF ENERGY (USE)
F-1 GE Nuclear Energy
GE-NE-0000-0003-5526-02a Paragraph IV.B of 10CFR50 Appendix G [1] sets limits on the upper shelf energy (USE) of the beltline materials. The USE must remain above 50 ft-lb at all times during plant operation, assumed here to be up to 32 EFPY. Calculations of 32 EFPY USE, using Reg. Guide 1.99, Rev. 2 [2] methods, are summarized in Table F-I.
The USE decrease prediction values from Reg. Guide 1.99, Rev. 2 [2] were used for the beltline plates and welds in Table F-1. These calculations are based on the peak 1/4T fluence for all materials other than the LPCI nozzle, for conservatism. Because the Charpy data available for the LPCI nozzle consists of shear energy of 70-80%, this conservatism is not applied to the 32 EFPY USE calculation for this component; the 1/4T fluence for the LPCI nozzle as provided in Table 4-4 is used. Based on these results, the beltline materials will have USE values above 50 ft-lb at 32 EFPY, as required in 10CFR50 Appendix G [1]. The lowest USE predicted for 32 EFPY is 60 ft-lb (for vertical weld heat 1 P3571).
F-2 GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a Table F-I: Upper Shelf Energy Evaluation for LaSalle Unit I Beltline Materials Initial mina 32 EFPY Test Longitudinal Transverse 114T
% Decrease 32 EFPY Location Heat Temperature USE USE"
%Cu Fluence USEb Transverse USES (oF)
(ft4b)
(ft4b)
(nn')
(lb)
Plates:
Lower C5978-1 160 136 88.4 0.11 7.1E+17 11 79 C5978-2 160 120 78 0.11 7.1E+17 11 69 C5979-1 160 136 88.4 0.12 7.1E+17 11.5 78 Lower Intermediate C6345-1d 160 165 107.3 0.15 7.1E+17 13 93 C6318-1 160 140 91 0.12 7.1E+17 11.5 81 C6345-2 160 161 104.7 0.15 7.1E+17 13 91 Middle A5333-1 160 155 100.8 0.12 7.1E+17 11.5 89 B0078-1 160 151 98.2 0.15 7.1E+17 13 85 C6123-2 160 151 98.2 0.13 7.1E+17 12 86 Welds:
Vertical:
3-308 305424 10 92 0.273 7.1E+17 23 71 1P3571 10 79 0.283 7.1E+17 23.5 60 4-308 3054140 10 92 0.337 7.1E+17 26.5 68 305414' 10 92 0.286 7.11E+17 23.5 70 120086 10 92 0.235 7.1E+17 21 73 1T2008' 10 92 0.286 7.1E+17 23.5 70 2-307 21935*
10 97 0.183 7.1E+17 18 80 2719_35 10 97 0.213 7.1E+17 19.5 78 12008*
10 97 0.235 7.1E+17 21 77 12008e 10 97 0.213 7.1E+17 19.5 78 Girth:
6-308 16329637 1 10 1103 0.205 7.1E+17 19 83 1-313 14P6519 10 116 0.131 7.1E+17 15 99 Forgings:
LPC, Nozzle 0Q2Q2 Q
10 73 0.10 I17E+17 7.5 68 a Values obtained from 13]
b Values obtained from Figure 2 of [2] for 32 EFPY 114T fluence c 32 EFPY Transverse USE = Initial Transverse USE * [1 - (% Decrease USE /100)]
d The initial transverse USE value is 65% of the highest 160"F data from CMTRs e Single Wire f Tandem Wire g Average of Charpy V-Notch data for %Shear >= 70 F-3
GE-NE-0000-0003-5526-02a APPENDIX F
REFERENCES:
"1. "Fracture Toughness Requirements", Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.
- 2.
"Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1199, Revision 2, May 1988.
- 3. T.A. Caine, "Upper Shelf Energy Evaluation for LaSalle Units 1 and 2", GENE, San Jose, CA, June 1990 (GE Report SASR 90-07).
F-4 GE Nuclear Energy
GE-NE-0000-0003-5526-02a APPENDIX G THICKNESS TRANSITION DISCONTINUITY EVALUATION G-1 GE Nuclear Energy
GE Nuclear Energy GE-NE-0000-0003-5526-02a Objectives:
The purpose of the following evaluations is to determine the hydrotest and the heat-up/cool down temperature (T) for the shell thickness transition discontinuity and to demonstrate that the temperature is bounded by the beltline hydrotest and heat-up/cool-down temperature, Methods and Assumptions:
An ANSYS finite element analysis was performed for the thickness discontinuity in the beltline region of LaSalle Unit 1. The purpose of this evaluation was to determine the RPV discontinuity stresses (hoop and axial) that result from a thickness transition discontinuity in the beltline region. The transition is modeled as a transition from 6 1/8" minimum thickness to 7 1/8" minimum thickness [1].
Three load cases were evaluated for the beltline shell discontinuity: 1) hydrostatic test pressure at 1563 psig, 2) a cool-down transient of 100OF/hr, starting at 550OF and decreasing to 70°F on the inside surface wall [2] and with an initial operating pressure of 1050 psig, and 3) a heat-up transient of 100°F/hr, starting at 70°F and increasing to 550°F on the inside surface wall [2] and with a final operating pressure of 1050 psig. For both transient cases it was assumed that the outside RPV wall surface is insulated with a heat transfer coefficient of 0.2 BTU/hr-ft2 OF [3] and that the ambient temperature is 100°F.
These are the bounding beltline transients of those described in Table 5.2-4 of the LaSalle Unit 1 and 2 UFSAR and Region B of the thermal cycle diagram [2] at temperatures for which brittle fracture could occur. Material properties were used from the Code of construction for the RPV Materials: Shell Plate Materials are ASME SA533, Grade B, Class 1, low alloy steel (LAS) [4].
Methods consistent with those described in Section 4.3 were used to calculate the T-RTNDT for the shell discontinuity for a hydrotest pressure of 1563 psig and the two transient cases.
The adjusted reference temperature values shown in Table 4-4 were added to the T-RTNDT to determine the temperature "T". The value of "7" was compared to that of the beltline G-2
GE-NE-0000-0003-5526-02a region for the same condition as described in Sections 4.3.2.2.1 for the hydrotest pressure case and 4.3.2.2.4 for the transient cases.
As shown below the stresses that result from the transition discontinuity are not significantly greater than those remote from the discontinuity (the difference in stress is less than 1 ksi for the pressure case and less than 2 ksi for the thermal cases). Therefore, the shell transition discontinuity stresses are also bounded by the beltline shell calculation.
The methods of ASME Code Section XI, Appendix G [5] are used to calculate the pressure test and thermal limits. The membrane and bending stress were determined from the finite element analysis and are shown below. The hoop stresses were more limiting than the axial stresses.
The stress intensity factors, KIm and KIb, are calculated using Code Case N-640 [6], and ASME Code Section XI Appendix A [7] and Appendix G [5]. Therefore, Km= Mm*CTm and Kib
= Mb*ab. The values of Mm and Mb were determined from the ASME Code Appendix G [5].
The stress intensity is based on a 1/4 T radial flaw with a six-to-one aspect ratio (length of 1.5T). The flaw is oriented normal to the maximum stress direction, in this case a vertically oriented flaw since the hoop stress was limiting.
The calculated value of Klm + KIb is multiplied by a safety factor (SF) (1.5 for pressure test and 2.0 for the transient cases), per ASME Appendix G [5] for comparison with KiR, the material fracture toughness expressed as K~c.
The relationship between KIc and temperature relative to reference temperature (T - RTNDT) is provided in ASME Code Section XI Appendix A [7] Paragraph A-4200, represented by the relationship (K1 units ksi-in 0.5):
Kjc = 33.2 + 20.734 exp[0.02 (T - RTNDT)]; therefore, T-RTNDT = ln[(K~c-33.2)/20.7 34]/0.02,
G-3 GE Nuclear Energy
GE-NE-0000-0003-5526-02a where Kic = SF * (K4m + Kib) for pressure test and Kqc = (SF
- Ip) + Kms for transient case.
This relationship is derived in the Welding Research Council (WRC) Bulletin 175 [8] as the lower bound of all dynamic fracture toughness data. This relationship provides values of pressure versus temperature (from KIR and (T - RTNDT), respectively).
The RTNDT is added to the (T-RTNDT) to determine the hydrotest, heat-up, and cool-down temperatures.
Analysis Information:
Thin Section Thickness tmin = 6.13 inch 41(t) = 2.47 inch0,5 Thick Section Thickness tin=
7.13 inch 4/(t) = 2.67 inch 5 G-4 GE Nuclear Energy
GE-NE-0000-0003-5526-02a Analysis and Results for the Hydrotest Pressure (Case 1):
Note that the axial stress is approximately 1/2 of the hoop stress.
Results and
Conclusions:
The maximum LaSalle Unit 1 plant-specific T-RTNDT for the thickness discontinuity is 68°F as shown in the table above. The limiting beltline weld material RTNDT at the region of the discontinuity is 88°F (see Table 4-4), so T = 156 0F. The limiting beltline plate RTNDT at the region of the discontinuity is 61°F (see Table 4-4), so T = 1290F.
At 1563 psig, Curve A is limited by the beltline curve. The T-RTNDT for the beltline region Curve A is 81°F at 1563 psig, and T = 1690F.
Because the beltline region pressure test temperature "T" of 169°F bounds the limiting plant specific thickness discontinuity for the case with the limiting ART value (T = 156°F for the weld G-5 GE Nuclear Energy
GE-NE-0000-0003-5526-02a material in the region of the discontinuity), the thickness discontinuity remains bounded by the beitline curve.
G-6 GE Nuclear Energy
r U -
r r
I r
r F
r rr V
r
[
r r
GE Nuclear Energy GE-NE-0000-0003-5526-02a Analysis and Results for Cool-down (CD - Case 2) and Heat-up (HU - Case 3):
Hoop Stress G-7 Primary Primary Secondary Secondary Location membrane bending membrane bending and Pressure Pm Pb Sm Sb Mm Mb =
Kp Kis T'RTNDT Case (psig)
(psi)
(psi)
(psi)
(psi) 2/3 Mm (psi Ini"2) (psi In"'2)
("F)
Maximum Hoop Stress from Discontinuity in thick section adjacent to discontinuity 1000 20920 475 ID-CD 1050 21966 499 135 6040 2.47 1.65 55136 10290 71.9 OD - HU 1050 21966
-499
-161 7253 2.47 1.65 53491 11558 70.7 Thin section remote from the discontinuity (t = 6.125")
1000 20740 534 ID-CD 1050 21777 561 0
3671 2.29 1.53 50785 5611 63.6 OD - HU 1050 21777
-561 1
5506 2.29 1.53 49070 8418 63.2 Thick section remote from the discontinuity (t = 7.125")
1000 17820 460 ID-CD 1050 18711 483
-1 6122 2.47 1.65 47061 10089 61.6 OD - HU 1050 18711
-483 1
7351 2.47 1.65 45470 12120 60.7
GE Nuclear Energy GE-NOE-0000-0003-5526-02a Results and
Conclusions:
The maximum LaSalle Unit 1 plant-specific T-RTNDT for the thickness discontinuity is 72°F. The limiting beltline material RTNDT in the region of the discontinuity is 88°F (see Table 4-4), so T = 160 0F. The limiting beltline plate RTNDT in the region of the discontinuity is 61OF (see Table 4-4), so T = 133°F.
At 1050 psig, Curve B is limited by the beltline curve. The T-RTNDT for the beltline region is 82°F at 1050 psig, and T = 1700F.
Because the beltline region pressure test temperature 'T' of 170°F bounds the limiting plant-specific thickness discontinuity for the case with the limiting ART value (T = 160°F for the weld material in the region of the discontinuity), the thickness discontinuity remains bounded by the beltline curve.
G-8
GE-NE-0000-0003-5526-02a Appendix G
References:
- 1. RPV Drawings a) CE Drawing # 232-788, Rev. 3, "General Arrangement Elevation for 251" I.D. BWR," (GE VPF #2029-117, Rev. 4).
b) CE Drawing # 232-790, Rev. 8, "Lower Vessel Shell Assembly Machining
& Welding for 251" I.D. BWR," (GE VPF #2029-036, Rev. 8).
c) CE Drawing # 232-791, Rev. 15, "Upper Vessel Shell Assembly Machining & Welding for 251" I.D. BWR," (GE VPF #2029-037, Rev. 14).
- 2. GE Drawing Number 731E776, "Reactor Vessel Thermal Cycles", GE-NED, San
- b.
Jose, CA, Revision 3 (GE Proprietary).
- 3. "Reactor Vessel Purchase Specification, Reactor Pressure Vessel",
(21A9242AF, Revision 9), December 1975.
- 4. T.A. Caine, "LaSalle Unit I RPV Surveillance Materials Testing and Analysis",
(GE-NE-523-A166-1294, Revision 1), June 1995.
- 5. "Fracture Toughness Criteria for Protection Against Failure", Appendix G to Section III or XI of the ASME Boiler and Pressure Vessel Code, 1995 Edition with Addenda through 1996.
- 6. "Alternative Reference Fracture Toughness for Development of P-T Limit CurvesSection XI, Division 1, "Code Case N-640 of the ASME Boiler and Pressure Vessel Code, Approval Date February 26, 1999.
- 7. "Analysis of Flaws", Appendix A to Section XI of the ASME Boiler and Pressure Vessel Code, 1995 Edition with Addenda through 1996.
- 8. "PVRC Recommendations on Toughness Requirements for Ferritic Materials",
Welding Research Council Bulletin 175, August 1972.
G-9 GE Nuclear Energy