ML043440227
| ML043440227 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/06/2004 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TVA-BFN-TS-428 | |
| Download: ML043440227 (30) | |
Text
Tennessee Valley Authority. Post Office Box 2000, Decatur, Alabama 35609-2000 December 6, 2004 TVA-BFN-TS-428 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN P1-35 Washington, D.C. 20555-0001 Gentlemen:
In the Matter of
)
Docket No. 50-259 Tennessee Valley Authority BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 -
TECHNICAL SPECIFICATIONS (TS)
CHANGE TS 428 -
UPDATE OF PRESSURE-TEMPERATURE (P-T)
CURVES Pursuant to 10 CFR 50.90, TVA is submitting a request for a TS change (TS-428) to license DPR-33 for BFN Unit 1.
The proposed change revises the reactor vessel P-T limit curves.
The requested P-T curves will allow for improved flexibility during reactor in-service and hydrostatic pressure testing as well as extending the Effective Full Power Year (EFPY) limits.
The proposed P-T curves were developed in accordance with 10 CFR 50, Appendix G, and the 1995 Edition with 1996 Addenda of ASME Section XI.
The P-T curve methodology included the incorporation of ASME Code Cases N-640 and N-588 and the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress.
Regulatory Guide 1.147 states that the ASME Code Cases N-588 and N-640 are acceptable for use in licensee Section XI inservice inspection programs.
In accordance with the guidance in Regulatory Issue Summary 2004-04, the use of NRC approved ASME Code Cases in conjunction with the 1995 Edition with 1996 Addenda of ASME Section XI endorsed in 10 CFR 50.55a for the development P-T limit curves does not require an exemption.
The proposed change revises the reactor vessel P-T limits depicted on current TS Figure 3.4.9-1 and adds a new TS Figure 3.4.9-2.
Curves specific to the reactor bottom head 4'?o PRled on rwcycind pawte
U. S. Nuclear Regulatory Commission Page 2 December 6, 2004 region are being added to the TS via these figures.
The revised P-T limits being requested were calculated using neutron fluence values of 5.3E17 n/cm2 at 12 effective full power years (EFPY) and 7.06E17 n/cm2 at 16 EFPY.
The 16 EFPY value represents the estimated service life which will have been reached by Unit 1 at the expiration of its current operating license.
The 12 EFPY value represents the midpoint between the current EFPY and 16 EFPY.
These fluence values conservatively assume operation over the entire analyzed period at an extended power uprate condition of 3952 MWt.
This power level is 120% of the BFN Unit 1 currently licensed power.
The fluence values were calculated in accordance with Regulatory Guide 1.190 and General Electric's NRC approved Licensing Topical Report NEDC-32983P (References 1).
TVA has determined there are no significant hazards considerations associated with the proposed change and the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).
Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Alabama State Department of Public Health. to this letter provides the description and evaluation of the proposed change.
This includes TVA's determination that the proposed change does not involve a significant hazards consideration and is exempt from environmental review. contains marked up pages of the appropriate TS for Unit 1. Enclosure 3 contains copies of the revised pages as they would appear following approval of this request.
Enclosures 4 and 5 contain copies of the proprietary and non-proprietary versions of the report from which the submitted Unit 1 P-T curves were developed.
Please note the report in Enclosure 4 contain information the General Electric Company considers proprietary and subsequently, pursuant to 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1),
requests such information be withheld from public disclosure.
The report contains an affidavit supporting this request. contains the redacted version of this reports, with the proprietary material removed, and is suitable for public disclosure.
NRC has previously approved a similar TS change on BFN Units 2 and 3 (References 2-5).
This change is necessary to ensure fidelity with the Units 2 and 3 TS.
The proposed TS change is necessary to support the restart of Unit 1. TVA requests the
U. S. Nuclear Regulatory Commission Page 3 December 6, 2004 amendment be approved by December 6, 2005, and the implementation of the revised TS be within 60 days of NRC approval.
TVA has determined there are no significant hazards considerations associated with the proposed TS change and the change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).
Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and attachments to the Alabama State Department of Public Health.
There are no regulatory commitments associated with this submittal.
If you have any questions about this amendment, please contact me at (256) 729-2636.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on December 6, 2004.
Sincerely,
U. S. Nuclear Regulatory Commission Page 4 December 6, 2004
References:
- 1. NRC letter to GE Nuclear Energy, dated September 14, 2001, "Safety Evaluation for NEDC-32983P, "General Electric Methodology for Reactor Pressure vessel Fast Neutron Flux Evaluation" (TAC No. MA9891)".
Units 2 and 3 -
Technical Specifications (TS)
Change TS-441 Revision 1 -
Update of Pressure-Temperature (P-T) Curves".
TVA Response to NRC Request for Additional Information (RAI) Regarding Units 2 and 3 -
Technical Specifications (TS)
Change No. 441R1 -
Update of Pressure -
Temperature (P-T) Curves".
TVA Revision to Implementation Plan Described in Units 2 and 3 -
Technical Specifications (TS)
Change No. 441 Revision 1 -
Pressure -
Temperature (P-T) Curve Update (MC0807 and MC0808)".
- 5. NRC letter to TVA, dated March 10, 2004, "Browns Ferry Nuclear Plant, Units 2 and 3 -
Issuance of Amendments Regarding Pressure-Temperature Limits Curves (TAC Nos. MC0807 and MC0808)".
Enclosures:
Evaluation of Proposed Change -
Marked Pages -
Revised Pages -
Proprietary Supporting Information -
Non-proprietary Supporting Information
U. S. Nuclear Regulatory Commission Page 5 December 6, 2004 State Health Officer Alabama Dept. of Public Health RSA Tower -
Administration Suite 1552 P.O. Box 303017 Montgomery, Alabama 36130-3017 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 Mr. Stephen J. Cahill, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970 Margaret Chernoff, Senior Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
ENCLOSURE 1 BROWNS FERRY NUCLEAR PLANT (BEN)
UNIT 1 TECHNICAL SPECIFICATIONS (TS) CHANGE TS 428 UPDATE OF PRESSURE-TEMPERATURE (P-T)
CURVES TVA EVALUATION OF PROPOSED CHANGE
1.0 DESCRIPTION
This letter is a request to amend Operating License DPR-33 for BFN Unit 1. The proposed amendment revises the Unit 1 reactor vessel P-T curves to reflect the results of an analysis which calculates the Unit 1 curves for 12 and 16 Effective Full Power Years (EFPY) of reactor operation.
The present BFN P-T curves are valid up to 12 EFPY for Unit 1.
TVA previously submitted and NRC approved a similar TS change on BFN Units 2 and 3 (References 1-4).
This change is necessary to ensure fidelity with the Units 2 and 3 TS.
The proposed TS change is necessary to support the restart of Unit 1. TVA requests the amendment be approved by December 6, 2005 and the implementation of the revised TS be within 60 days of NRC approval.
2.0 PROPOSED CHANGE
The specific changes to the TS P-T figures are described below:
- a. TS Figure 3.4.9-1 is deleted and replaced in its entirety.
The new TS Figure 3.4.9-1 contains an added curve which specifically details reactor vessel bottom head temperature versus pressure limitations for non-critical heatup/cooldown operational conditions;
- b. A new TS Figure 3.4.9-2 has been added which contains curves for the reactor vessel bottom head and the upper vessel/beltline regions for in-service leak and hydrostatic testing activities; The legend information on these TS figures has been revised to appropriately describe the additional curves and their usage.
In addition, references to these figures within the TS have been revised as necessary to appropriately reflect their use.
E1-1 contains marked up pages of the appropriate TS for Unit 1. Enclosure 3 contains copies of the revised pages as they would appear following approval of this request.
3.0 BACKGROUND
The present BFN P-T curves are valid up to 12 EFPY. As well as extending the EFPY limits, the requested P-T curves will allow for improved flexibility during reactor in-service and hydrostatic pressure testing.
The addition of the separate, specific bottom head curves provides this improved flexibility.
Inclusion of the new bottom head curves directly on the TS figures will more clearly delineate the reactor vessel bottom head temperature limits, which are distinct from the temperature limits for the beltline and upper vessel regions.
4.0 TECHNICAL ANALYSIS
4.1 P-T Curve Overview All components of the reactor coolant system are designed to withstand the 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.
Therefore, P-T limits are established to ensure the reactor coolant system is operated under conditions that preclude brittle failure of the reactor coolant pressure boundary.
10 CFR 50 Appendix G requires the establishment of these P-T limits for reactor coolant pressure boundary materials.
Appendix G also requires an adequate margin to brittle failure be maintained during normal operation, anticipated operational occurrences, and system hydrostatic tests.
The P-T limits are acceptance limits in themselves, since operation in accordance with these limitations precludes operation in an unanalyzed condition.
The P-T limits are not derived from Design Basis Accident analyses.
The proposed 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.
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The P-T limits are currently in TS Figure 3.4.9-1.
The existing figure contains three separate P-T curves, which define the P-T limitations for the entire reactor pressure vessel for the following reactor operating conditions:
Curve 1 includes P-T restrictions on reactor vessel head boltup.
Hydrostatic/leak testing of the reactor vessel is performed in accordance with these limitations prior to startup after a refueling outage to verify the vessel is leak tight.
The minimum allowable testing temperatures are established by the P-T curves; Curve 2, the heatup and cooldown curve, includes P-T restrictions for startup and shutdown operations when the core is not critical; and Curve 3 specifies the P-T limits during operations when the core is critical.
The primary system pressure and temperature are monitored and compared to the applicable curve to ensure operation is within the allowable region.
The new P-T curve set being requested includes two additional curves.
These are:
A fourth curve, which specifies the temperature limits for the reactor vessel bottom head region during in-service and hydrostatic testing of the reactor vessel.
The minimum temperature for the bottom head during this testing is established by this P-T curve; and A fifth curve, which specifies the temperature limits for the reactor vessel bottom head during startup and shutdown operations when the reactor is not critical.
These five separate curves are depicted on the revised TS Figure 3.4.9-1 and new Figure 3.4.9-2 to clearly show the P-T limitations for the reactor bottom head area and the vessel beltline and upper vessel areas for all operating conditions.
4.2 Methodology The P-T limits are primarily dependent upon the fracture toughness of the vessel ferritic materials.
The key parameters which characterize a material's fracture toughness are the reference temperature of nil-ductility transition (RTNDT) and the Upper Shelf Energy (USE).
These E1-3
parameters are defined in 10 CFR 50, Appendix G, and in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI.
These documents also contain the requirements used to establish the P-T operating limits to avoid brittle fracture.
Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," provides an acceptable method for calculating P-T limits that satisfies the requirements of 10 CFR 50 Appendix G. The P-T curves for have been recalculated based on methodologies consistent with this regulatory guide using plant-specific material and fluence information.
The BFN Unit 1 specific RTNT, weld material composition, and fluence information has previously been provided to NRC (see References 5-12).
The fluence values used in development of these curves were calculated in accordance with General Electric (GE)
Licensing Topical Report NEDC-32983P-A, which was approved by NRC (Reference 13).
Principal assumptions for this analysis include:
Hydrostatic pressure testing will be conducted at or below 1064 psig; Maximum values of 16 EFPY will be reached within the limits of the current operating licenses. Midpoint values between the current EFPY and 16 EFPY at 12 EFPY are also calculated; and A peak neutron flux of 1.4E09 n/cm2-sec at extended power uprate (EPU) conditions (3952 MWt).
Note this power is 120% of the current licensed power for Unit 1.
This flux is assumed over the entire calculated EFPY period, even though at this time Unit 1 has not operated at the EPU power level.
4.3 Results The following fluence values were calculated using the EPU flux of 1.4E09 n/cm2-sec.
VALUE RESULT Unit 1 12 EFPY fluence 5.3E17 n/cm2 Unit 1 12 EFPY 1/4T fluence for lower-3.67E17 n/cm2 intermediate shell plate and axial welds E1-4
VALUE RESULT Unit 1 12 EFPY 1/4T fluence for lower 2.97E17 n/cm2 shell plate and axial welds and lower to lower-intermediate girth weld Unit 1 16 EFPY fluence 7.06E17 n/cm2 Unit 1 16 EFPY 1/4T fluence for lower-4.89E17 n/cm2 intermediate shell plate and axial welds Unit 1 16 EFPY 1/4T fluence for lower 3.96E17 n/cm2 shell plate and axial welds and lower to lower-intermediate girth weld The limiting adjusted reference temperature (ART) value of 1191F for the 16 EFPY calculation remains well below the 200OF criterion of RG 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision 2.
The USE equivalent margin analyses values calculated for end of life (i.e., 16 EFPY for Unit 1) remain within the limits of RG 1.99, Revision 2 and 10 CFR 50 Appendix G.
A single set of P-T curves for the heatup and cooldown operating condition at a given EFPY that apply for both the 1/4T and 3/4T locations was developed.
When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (assumed inside surface flaw) and the 3/4T location (assumed 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 since 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 GE reports for Unit 1 provided in Enclosures 4 and 5 demonstrate the technical methods and contain the data for producing the composite P-T curves which are to be placed in the TS.
Table B-2 in the Unit 1 report contains the data for producing the composite 12 EFPY curves.
In the same manner, Table B-4 of the Unit 1 report contains the data for producing the composite 16 EFPY curves.
EI-5
4.4 Conclusion The proposed P-T curves have been developed utilizing the methodology of RG 1.190 and ASME Section XI.
The regulatory guidance provides an allowance for margin to be included in the bounding values of the ART.
Use of this methodology ensures adequate safety margins are maintained.
In addition, the analysis conforms to the requirements of 10 CFR 50, Appendix G, which ensures the most limiting material is considered in the development of the P-T curves.
The vessel is in compliance with the regulatory requirements, adequate safety margins are maintained, and, therefore, Unit 1 operation to 12 or 16 EFPY will not have an adverse effect on reactor vessel fracture toughness.
5.0 REGULATORY SAFETY ANALYSIS Pursuant to 10 CFR 50.90, TVA is submitting a request for a Technical Specifications (TS) change to license DPR-33 for BFN Unit 1. The proposed change revises the reactor vessel pressure-temperature (P-T) limits depicted on current TS Figure 3.4.9-1 and adds a new TS Figure 3.4.9-2.
Curves specific to the reactor bottom head region are being added to the TS via these figures.
5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:
- 1.
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
No.
The proposed changes deal exclusively with the reactor vessel P-T curves, which define the permissible regions for operation and testing.
Failure of the reactor vessel is not considered as a design basis accident.
Through the design conservatisms used to calculate the P-T curves, reactor vessel failure has a low probability of occurrence and is not considered in the safety analyses.
The proposed changes adjust the E1-6
reference temperature for the limiting material to account for irradiation effects and provide the same level of protection as previously evaluated and approved.
The adjusted reference temperature calculations were performed in accordance with the requirements of 10 CFR 50 Appendix G using the guidance contained in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," to reflect use of the operating limits to no more than 16 Effective Full Power Years (EFPY).
These changes do not alter or prevent the operation of equipment required to mitigate any accident analyzed in the BFN Final Safety Analysis Report.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2.
Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response
No.
The proposed changes to the reactor vessel P-T curves do not involve a modification to plant equipment. No new failure modes are introduced.
There is no effect on the function of any plant system, and no new system interactions are introduced by this change.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3.
Does the proposed change involve a significant reduction in a margin of safety?
Response
No.
The proposed curves conform to the guidance contained in Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," and maintain the safety margins specified in 10 CFR 50 Appendix G. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
E1-7
Based on the above, TVA concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of ano significant hazards consideration" is justified.
5.2 Applicable Regulatory Requirements/Criteria The regulatory requirements for fluence calculations are in General Design Criteria (GDC) 30 and 31.
NRC issued RG 1.190 in March 2001, which provided state-of-the-art calculations and measurement procedures that are acceptable to the NRC staff for determining pressure vessel fluence.
NRC has approved vessel fluence calculation methodologies which satisfy the requirements of GDC 30 and 31 and are done with approved methodologies or with methods which are shown to adhere to the guidance in RG 1.190.
The analyses supporting this submittal were performed in accordance with RG 1.190 guidance.
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
6.0 ENVIRONMENTAL CONSIDERATION
A review has determined 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.
E1-8
7.0 REFERENCES
- 1.
TVA letter to NRC, dated September 18, 2003, "Browns Ferry Nuclear Plant (BFN) -
Units 2 and 3 -
Technical Specifications (TS) Change TS-441 Revision 1 -
Update of Pressure-Temperature (P-T) Curves".
- 2.
TVA letter to NRC, dated December 8, 2003, "Browns Ferry Nuclear Plant (BFN) -
TVA Response to NRC Request for Additional Information (RAI) Regarding Units 2 and 3 -
Technical Specifications (TS) Change No. 441R1 -
Update of Pressure -
Temperature (P-T) Curves".
- 3.
TVA letter to NRC, dated February 24, 2004, "Browns Ferry Nuclear Plant (BFN) -
TVA Revision to Implementation Plan Described in Units 2 and 3 -
Technical Specifications (TS)
Change No. 441 Revision 1 -
Pressure -
Temperature (P-T)
Curve Update (MC0807 and MC0808)".
- 4. NRC letter to TVA, dated March 10, 2004, "Browns Ferry Nuclear Plant, Units 2 and 3 -
Issuance of Amendments Regarding Pressure-Temperature Limits Curves (TAC Nos. MC0807 and MC0808)".
- 5.
TVA letter to NRC, dated July 7, 1992, "Browns Ferry Nuclear Plant (BFN), Sequoyah Nuclear Plant (SQN), and Watts Bar Nuclear Plant (WBN) -
Response to Generic Letter 92-01 (Reactor Vessel Structural Integrity)".
- 6.
TVA letter to NRC, dated December 1, 1992, "Browns Ferry Nuclear Plant (BFN) -
Completion of Commitment Made in Response to Generic Letter 92-01, "Reactor Vessel Structural Integrity"".
- 7.
TVA letter to NRC, dated August 2, 1993, "Browns Ferry Nuclear Plant (BFN) -
Response to Request for Additional Information, Generic Letter 92-01, Revision 1".
- 8.
TVA letter to NRC, dated May 23, 1994, "Browns Ferry Nuclear Plant (BFN) -
TVA's Response to NRC's Letter Dated April 19, 1994, Generic Letter (GL) 92-01, Revision 1, Reactor Vessel Structural Integrity"".
- 9.
TVA letter to NRC, dated July 28, 1994, "Browns Ferry Nuclear Plant (BFN) -
Supplemental Response to TVA letter Dated May 23, 1994, Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity".
Units 1, 2 and 3 -
Generic Letter (GL) 92-01, Reactor Vessel Structural Integrity -
Update to the Initial Reference Nil-Ductility Temperature (RTNDT), Chemical Composition and Fluence Values".
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- 11. TVA letter to NRC, dated November 7, 1995, "Response to NRC Generic Letter (GL) 92-01, Revision 1, Supplement 1; Reactor Vessel Structural Integrity - Browns Ferry (BFN), Watts Bar (WBN), and Sequoyah (SQN) Nuclear Plants (TAC Nos. M92649, M92650, M92651, M92730, M92731, M83525, AND M83526)."
Units 1, 2, and 3 -
Generic Letter (GL) 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity - Response to NRC Request for Additional Information (TAC Nos. MA1179, MA1180, and MA1181)".
- 13. NRC letter to GE Nuclear Energy, dated September 14, 2001, "Safety Evaluation for NEDC-32983P, "General Electric Methodology for Reactor Pressure vessel Fast Neutron Flux Evaluation" (TAC No. MA9891)".
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ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 1 TECHNICAL SPECIFICATIONS (TS)
CHANGE TS 428 UPDATE OF PRESSURE-TEMPERATURE (P-T)
CURVES MARKED UP TECHNICAL SPECIFICATION PAGES
RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 NOTES
- 1. Only required to be performed during RCS heatup and cooldown operations or RCS inservice leak and hydrostatic testing when the vessel pressure is > 312 psig.
- 2. The limits of Figure 3.4.9-24, CurGe
-N.1,l may be applied during nonnuclear heatup and ambient loss cooldown associated with inservice leak and hydrostatic testing provided that the heatup and cooldown rates are < 15F/hour.
- 3. The limits of Figures 3.4.9-1 and 3.4.9-2 do not apply when the tension from the reactor head flange bolting studs is removed.
Verify:
30 minutes
- a. RCS pressure and RCS temperature are within the limits specified by Curves No. 1 and No. 2 brof Figures 3.4.9-1 and 3.4.9 (Cu~r.
No. 1 and Cur.'o N. 2);
and
- b. RCS heatup and cooldown rates are
< 1 00F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.
SR 3.4.9.2 Verify RCS pressure and RCS temperature Once within 15 are within the criticality limits specified in minutes prior to Figure 3.4.9-1, Curve No. 3.
control rod withdrawal for the purpose of achieving criticality (continued)
B FN-U NIT I 3.4-26 Amendment No. 234
RCS PIT Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.9.5
- 1. Only required to be performed when tensioning the reactor vessel head bolting studs.
- 2. The reactor vessel head bolts may be partially tensioned (four sequences of the seating pass) provided the studs and flange materials are > 700F.
Verify reactor vessel flange and head flange temperatures are > 8300F.
30 minutes SR 3.4.9.6
NOTE
Not required to be performed until 30 minutes after RCS temperature < 850F in MODE 4.
Verify reactor vessel flange and head flange 30 minutes temperatures are > 83§0 F.
--- NOTE----
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature < 1000F in MODE 4.
Verify reactor vessel flange and head flange 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> temperatures are > 8300F.
I I
BFN-UNIT 1 3.4-28 Amendment No. 234
RCS P/T Limits 3.4.9 s I
--X REPLACE WITH I FIGURE AND ADD FIGURE 3.4 1bUU 1500 1400 1n 1300 0.
o 1200 M
a 1100 i
1000 w
UI W'
900 C
800 M
700 Z
600 M
i 500 t
400 en 300 a-
.BROWNS FERRY 1_
UNIT I 1
2 3
II IEW
.9-2 Curve No 1 Minimum te erature for pressurets such as requirey Section Xl.
rve No. 2 Minimum temperature for mechanical heatup or cooldown following nuclear shutdown.
Curve No. 3 Minimum temperature for core operation (criticality).
Notes These curves include sufficient margin to provide protection against edwater nozzle radation. The curves allo Xor shifts in RTNDT of the Re
~tor vessel beltline material accordance with Reg.
- de 1.99 Rev. 2 to com nsate for radiation embri ent for 12 EFPY.
i0 1/vI/ I/
312 PSIG
'p.
I If A4 200 100 0 j CURVES 1, 2, & 3 ARE VALID FOR i/
12 EFPY OF OPERATION 4
-l.
-C--.-.-......
I...
I,
, I,. I.,,
50 100 150 200 250 300 350 40 MINMUM REACTORVESSEL METAL TEMPERATURE (¶)
A
,W Figure 3.4.9-1 Pressure/Temperature Limits BFN-UNIT 1 3.4-29 Amendment No. 234
Reactor Steam Dome Pressure 3.4.9 1400 Curve No.
1 11 2
13 Minimum temperature for I
bottom head during 1300 mechanical heatup or BR WNS FE RY NI1 1
l cooldown following CUF ES l, 2
ND 3 nuclear shutdown.
0E ERA 1 Curve No. 2 1100-1 Minimum temperature for i
upper RPV and beltline during mechanical 1000 hcatup or cooldown 0100
- following nuclear shutdown.
Curve No. 3 Minimum temperature for o
800 core operation (criticality).
LU 7Notes 60 These curves include o
600 D ufficicnt margin to providc protection against feedwater IX 500 nozzle degradation.
Z The curves allow for I
shifts in RT of the 400 Reactor vess" beltline X
materials, in accordance with Reg.
300 Guide 1.99 Rev.
2 to 1U compensate for I
/radiation embrittlement 200 for 16 EFPY.
Borrom H
.AD FLANGE The acceptable area for 8F LREGION operation is to the 100 73-F
-right of the applicable O JJl 2
L
.curves.
0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE F Figure 3.4.9-1 Pressure/Temperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations BFN-UNIT 1 3.4-29 Amendment No. 234
Reactor Steam Dome Pressure 3.4.9 1400 Curve No. 1 3-1 12 Minimum temperature for bottom head during BRONS FERY UIT 1
.in-service leak or B
CRO S FE RY IT 1 hydrostatic testing.
1200 CURVES 1 ND 2 ARE VALID FOR 12.
EFPY OF 0 ERATON :
C 0
I Curve No. 2 1100 Minimum temperature for Iupper RPV and beltline during in-service leak 1000 or hydrosLaLic LesLing.
X 900
- 1.
NoLes These curves include c
- sufficient margin to 800
/-
-provide protection X
,.against feedwater nozzle degradation.
The curves allow for j
700 shifts in RTN of the X
.Reactor vessel beltline materials, in cc 600
-accordance with Reg.
o Guide 1.99 Rev. 2 to CI:
compensate for 500 radiation embrittlement W
.for 16 EFPY.
2 400 The acceptable area for operation is to the t
0Dright of the applicable C
300 curves.
U) lU
____FLANGE
- 0. 200
-BOTTOM REGION HEAD 83'F 08'F 100 0
25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE F Figure 3.4.9-2 PressurelTemperature Limits for Reactor In-Service Leak and Hydrostatic Testing BFN-UNIT 1 3.4-29a Amendment No. 234
Reactor Steam Dome Pressure 3.4.9 1400 1300 1200 1100 1000 DO NOT USE THIS FIGURE Curve No. 1 Minimum temperature for bottom head during mechanical heatup or cooldown following nuclear shutdown.
Curve No. 2 Minimum temperature for upper RPV and beltline during mechanical heatup or cooldown following nuclear ture for
- h is dude in to ion er ion.
w for Rof the Reactor vesse~l beltline This curve applies to operations >12 EFPY.
For current operation, use previous curve, whic valid up to 12 EFPY.
LU U,
Cn LU W-400 300 200 materials, in accordance with Reg.
Guide 1.99 Rev. 2 to compensate for radiation embrittlement for 16 EFPY.
The acceptable area for operation is to the right of the applicable curves.
100 0
0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE eF Figure 3.4.9-1 PressurelTemperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations BFN-UNIT 1 3.4-29b Amendment No. 234
Reactor Steam Dome Pressure 3.4.9 1400 1300 1200 1100 1 000 I__
.12 BROW S FE RY L IT 1 CERV id 2,Al D?_3_
ARE 'ALID FOR 6
./
EFPY OF O ERATION r
_ I Curve No. 1 Minimum temperature for bottom head during in-service leak or hydrostatic testing.
Curve No. 2 Minimum temperature for upper RPV and beltline during in-service leak or hydrostatic testing.
/I DO NOT USE THIS FIGURE This curve applies to operations >12 EFPY.
For current operation, use previous curve, which is valid up to 12 EFPY.
ude to n.
for the ltline eg.
Lo lement 6i U)
U, 0.
400 300 200 BOTTOM REGNON HEAD F
68'F I--
8'
-Ii
_ 'lie acceptaole a]ea for operation is to the right of the applicable curves.
100 0
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE OF Figure 3.4.9-2 Pressure/Temperature Limits for Reactor In-Service Leak and Hydrostatic Testing BFN-UNIT 1 3.4-29c Amendment No. 234
ENCLOSURE 3 BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 TECHNICAL SPECIFICATIONS (TS) CHANGE TS 428 UPDATE OF PRESSURE-TEMPERATURE (P-T)
CURVES REVISED TECHNICAL SPECIFICATIONS PAGES
RCS PIT Limits 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1
-a
NOTES----
- 1. Only required to be performed during RCS heatup and cooldown operations or RCS inservice leak and hydrostatic testing when the vessel pressure is > 312 psig.
- 2. The limits of Figure 3.4.9-2 may be applied during nonnuclear heatup and ambient loss cooldown associated with inservice leak and hydrostatic testing provided that the heatup and cooldown rates are < 15F/hour.
- 3. The limits of Figures 3.4.9-1 and 3.4.9-2 do not apply when the tension from the reactor head flange bolting studs is removed.
Verify:
30 minutes
- a. RCS pressure and RCS temperature are within the limits specified by Curves No. 1 and No. 2 of Figures 3.4.9-1 and 3.4.9-2; and
- b. RCS heatup and cooldown rates are
< 100F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.
SR 3.4.9.2 Verify RCS pressure and RCS temperature Once within 15 are within the criticality limits specified in minutes prior to Figure 3.4.9-1, Curve No. 3.
control rod withdrawal for the purpose of achieving criticality (continued)
BFN-UNIT 1 3.4-26 Amendment No. 234
RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.9.5
- 1. Only required to be performed when tensioning the reactor vessel head bolting studs.
- 2. The reactor vessel head bolts may be partially tensioned (four sequences of the seating pass) provided the studs and flange materials are > 700F.
Verify reactor vessel flange and head flange temperatures are > 830F.
30 minutes SR 3.4.9.6
NOTE------
Not required to be performed until 30 minutes after RCS temperature < 850F in MODE 4.
Verify reactor vessel flange and head flange 30 minutes temperatures are > 830F.
SR 3.4.9.7 A-----------------NOTE Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature < 1 000F in MODE 4.
Verify reactor vessel flange and head flange 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> temperatures are > 830F.
BFN-UNIT I 3.4-28 Amendment No. 234
Reactor Steam Dome Pressure 3.4.9 1400 1300 1200 1100 C 1000 a
900 a.
o 800 II-j n
700 or o
600 I-a 500 z
16-400 Cn 300 Cn LU 9L200 I ~
12 ~3 BR NSFEFRY NI 1_
CURVES 2, END 3 ARE VA ID FOR 12.
EFPY 0' OIERA Io JI 2
,..1 IL-BOTTOM G
HEAD FLANGE 8
REGION___
Curve No.
1 Minimum temperature for bottom head during mechanical heatup or cooldown following nuclear shutdown.
Curve No.
2 Minimum temperature for upper RPV and beltline during mechanical hcatup or cooldown following nuclear shutdown.
Curve No.
3 Minimum temperature for core operation (criticality).
Notes These curves include sufficient margin to provide protection against feedwater nozzle degradation.
The curves allow for shifts in RT" 0 of the Reactor vessel beltline materials, in accordance with Reg.
Guide 1.99 Rev.
2 to compensate for radiation embrittlement for 16 EFPY.
The acceptable area for operation is to the right of the applicable curves.
100 0
0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE °F Figure 3.4.9-1 Pressure/Temperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations BFN-UNIT 1 3.4-29 Amendment No. 234
Reactor Steam Dome Pressure 3.4.9 1400 Curve No.
1
,1 12 Minimum temperature for 1300 bottom head during BROW S FE RY Iin-service leak or hydrostatic testing.
120CURV S 1 2MD 2 ARE rALID FOR 2.
EFPY OF 0 ERATON
.Curve No.
2 1100
/-
Minimum temperature for upper RPV and beltline during in-service leak 1000
.or hydrosLaLic LesLing.
U 900 NoLes These curves include
.'sufficient margin to provide protection t
800 against feedwater E
nozzle degradation.
R
- The curves allow for U
700-shifts in RTN of the e
.*Reactor vessel beltline materials, in 600 accordance with Reg.
o
.Guide 1.99 Rev. 2 to U
.compensate for j
500 radiation embrittlement for 16 EFPY.
400 The acceptable area for operation is to the U
right of the applicable I
300 i
curves.
LII I]
.FLANGE IL 200
-BOTTOM-;
REGION HEAD 3
100 I-KB 1
0 25 50 75 100 125 150 175 200 MINIMUM REACTORVESSEL METALTEMPERATURE F Figure 3.4.9-2 Pressure/Temperature Limits for Reactor In-Service Leak and Hydrostatic Testing BFN-UNIT 1 3.4-29a Amendment No. 234
Reactor Steam Dome Pressure 3.4.9 1400 1300 1200 1100 1000 BRLMS FE RY I
1 1
3 CU S I 2112 D 3 l
l ARE -ID FOR 16 EFEY OIERA IO.
I-7-
-1 I
I Curve No. I Minimum temperature for bottom head during mechanical heatup or cooldown following nuclear shutdown.
Curve No. 2 Minimum temperature for upper RPV and beltline during mechanical heatup or cooldown following nuclear IlI DO NOT USE THIS FIGURE This curve applies to operations >12 EFPY.
For current operation, use previous curve, whic valid up to 12 EFPY.
ture for h is clude in to ion er ion.
w for of the Reactor vessel beltline materials, in accordance with Reg.
Guide 1.99 Rev. 2 to compensate for radiation embrittlement for 16 EFPY.
The acceptable area for operation is to the right of the applicable curves.
u)
UL U,W 400 300 200 100 0
0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE eF Figure 3.4.9-1 Pressure/Temperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations BFN-UNIT 1 3.4-29b Amendment No. 234
Reactor Steam Dome Pressure 3.4.9 1400 1300 1200 1100 1000 1
/21 BROW4S FE RY U IT 1 i ZMRK-
_s
$-I 2
I ARE VALID FOR6 EFPY OF O ERAT6ON.
1' _ I Curve No. 1 Minimum temperature fox bottom head during in-service leak or hydrostatic testing.
Curve No. 2 Minimum temperature for upper RPV and beltline during in-service leak or hydrostatic testing.
_I DO NOT USE THIS FIGURE This curve applies to operations >12 EFPY.
For current operation, use previous curve, which is valid upto 12 EFPY.
ude I to n.
m.
for
- the Itline eg.
Lo lement i,
C.
400 300 200 I
I_
I i
.i I
-C--1 BOTTOFLANGE BOTTOMREGION HEAD 3
68'F 8
IL I 71J17 II Tlme acceptaune area for operation is to the right of the applicable curves.
100 0
0 25 5a 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE *F Figure 3.4.9-2 Pressure/Temperature Limits for Reactor In-Service Leak and Hydrostatic Testing BFN-UNIT 1 3.4-29c Amendment No. 234