ML20206P127
| ML20206P127 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/31/1998 |
| From: | Hammer M NORTHERN STATES POWER CO. |
| To: | NRC |
| Shared Package | |
| ML20206P125 | List: |
| References | |
| NUDOCS 9901080036 | |
| Download: ML20206P127 (9) | |
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, UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 REQUEST FOR AMENDMENT TO OPERATING LICENSE DPR-22 i LICENSE AMENDMENT REQUEST DATED DECEMBER 31,1998 1 Northern States Power Company, a Minnesota corporation, requests authorization for changes ;
to Appendix A of the Monticello Operating License as shown on the attachments labeled Exhibits l A, B, and C. Exhibit A describes the background and proposed changes, describes the reasons l for the changes, and contains a Safety Evaluation, a Determination of No Significant Hazards 1' Consideration and an Environmental Assessment. Exhibit B contains current Technical Specification pages marked up with the proposed changes. Exhibit C is a copy of the Monticello Technical Specification pages incorporating the proposed changes.
l This letter contains no restricted or other defense information.
1 NORTHERN STATES POWER COMPANY i
By h 4LatM1}A .
Michael P. Hammer " I Plant Manager l Monticello Nuclear Generating Plant l On this 31 day of heceder \qqB before me a notary public in and for said County, personally appeared Michael F. Hammer, Plant Manager, Monticello Nuclear '
Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true i and that it is not interposed for delay.
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SAMUEL. l. sHIREY tamuel I. Shirey d J/
Notary Public- Minnesota Sherburne County Q)heuc r -- - -
WWWA uf comm.Exp.Jan.31M My Commission Expires January 31,2000
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9901000036 981231 PDR ADOCK 05000263 P PDR _
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l EXHIBIT A l MONTICELLO NUCLEAR GENERATING PLANT l
License Amendment Request Dated December 31,1998 i Revision of Reactor Pressure Vessel Pressure-Temperature Limit Curves and i Removal of Standby Liquid Control Relief Valve Setpoint i
Evaluation of Proposed Changes, to the Technical Specifications for Operating License DPR-22 1
Pursuant to 10 CFR Part 50, section 50.90, Northern States Power Company hereby proposes changes to Technical Specification Figures 3.6.1 (Core Beltline Operating l Limits Curve Adjustment vs. Fluence),3.6.2 (Minimum Temperature vs. Pressure for l Pressure Tests),3.6.3 (Minimum Temperature vs. Pressure Mechanical Heatup or '
Cooldown Without the Core Critical), and 3.6.4 (Minimum Temperature vs. Pressure for Critical Core Operation). Related to this, completed one time RPV sample surveillance i requirements are being relocated from the SR section to the Bases, and the requirement to withdraw a specimen at the next refueling outage ("three fourths service life")is being removed.
In addition, it is proposed to remove a redundant requirement for the Standby Liquid Control System relief valve setpoint.
Background:
An important input parameter used in estimation of allowable limits for Reactor Pressure Vessel (RPV) temperature and pressure is fracture toughness of the RPV beltline shell material. Fracture toughness of ferritic steels, which form the pressure boundary in reactor pressure vessels, is a strong function of temperature. In the American Society of Mechanical Engineers (ASME) Section Ill, Boiler and Pressure Vessel Code, Appendices A and G, fracture toughness is correlated with reference temperature, RTsor which is a characteristic temperature that defines transition from ductile to brittle behavior. For a given service temperature, fracture toughness decreases as RTsor increases. In the limiting vessel beltline region, accumulated fast i neutron irradiation leads to an increase in RTsor, thereby leading to a decrease in
' toughness. The increase in RTsor due to neutron irradiation is characterized by a parameter called adjusted reference temperature (ART). ART is defined by guidelines of Regulatory Guide 1.99, Rev. 2 and consists of an unirradiated initial value, plus an increase due to irradiation, and margin terms which account for scatter in the metallurgical data.
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'The IVionticello RPV originally contained several sets of reactor vessel material specimens. Each set of specimens contained materials representative of the Monticello RPV beltline region. Two plate surveillance specimens, designated C2220-1 and C2220-2 from heat C2220, are representative of the Monticello RPV beltline region lower intermediate shell course. This plate material has the most limiting material l properties. l l
Two specimen sets were rernoved from the Monticello RPV in 1981. Those specimen l sets had a low lead factor of 0.3 (ratio of specimen neutron fluence to highest neutron j fluence experienced by the RPV wall). One of the two specimen sets was tested for '
RTer at that time. The second set of specimens was later installed in the Prairie island l RPV for continued irradiation at an accelerated fluence (lead factor > 10). It was removed from the Prairie Island RPV and tested in 1996. I l
Results of the second surveillance material tests were evaluated utilizing guidance of Regulatory Guide 1.99, Rev. 2, Position 2.1. Based on this testing and evaluation, a more accurate chemistry factor, which is a function of copper and nickel content, was determined and new RPV P-T limit curves were developed. In addition, curves of Figures 3.6.2,3.6.3, and 3.6.4 labeled "RPV CORE BELTLINE (ZERO FULL POWER YEARS)" have been revised to include the results of Oak Ridge National Laboratories (ORNL) testing of archived non-irradiated plate material (Reference 1).
To reflect these changes, Monticello TS Figures 3.6.1 through 3.6.4, and associated Limiting Conditions for Operation, Surveillance Requirements and Bases sections are being revised.
The current prescribed schedule for test specimen removal at "one fourth and three fourths service life" was determined prior to issuance of Monticello's full term operating license in 1981. Since that time, two sets of surveillance specimens have been analyzed. The first sample was tested in 1981. The second set of specimens was irradiated in the Prairie Island RPV to a fluence level equivalent to 40 years operation of the Monticello RPV. Therefore, available surveillance data bounds end of life (EOL')
exposure. NSP therefore proposes to revise Technical Specification Section 4.6.B.2 to allow the flexibility permitted by ASTM E 185-66 in determining when remaining samples are withdrawn and tested. Removal of a surveillance specimen set at the currently prescribed time would adversely impact the availability of future RPV
- surveillance data.
l Currently, the Standby Liquid Control (SBLC) relief valve setpoint is specified in TS Section 4.4.A.2.c. Listing tHs pdfic setpoint is not consistent with other TS sections
' EOL based on past eporating i e cry ,wer uprate conditions (1775MWt authorized 9/16/98) and 80%
capacity factor.
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,and does not allow revising the setpoint without NRC approval. The setpoint of the SBLC system relief valves is governed by provisions of the ASME Code Section XI as required in TS section 3.15. Tnis amendment request proposes to delete this setpoint, making it consistent with NUREG-1433, Standard Technical Specifications.
l Proposed Changes and Reasons for Changes:
LIST OF FIGURES (page v)
Revise Figure 3.6.4 title from " Minimum Temperature vs. Pressure for Core Operation,"
to " Minimum Temperature vs. Pressure for Critical Core Operation." This is a more descriptive title and is consistent with changes being made to the figure itself.
l Section 4.4.A.2.c (page 94)
L Delete the requirement to test the Standby Liquid Control relief valves between 1350 l l and 1450 psig. Specific test requirements and associated setpoint values need not be stated in Technical Specifications. This proposed change will allow flexibility of revising the relief valve setpoint.
l Section 4.6.B (page 122) l Part of section 4.6.B.2 and all of section 4.6.B.3 are being deleted. This inforrnation is historical since these surveillance requirements have been completed. For continuity
, and historical background, this information is being moved to Bases Section 3.6/4.6. I
! Deletion of the prescribed specimen withdrawal at three fourths service life provides l
! flexibility to remove the remaining RPV surveillance specimen sets at optimum l intervals.
l Figures 3.6.1 through 3.6.4 (pages 133,134,135,136) l Update Figures 3.6.1 (Core Beltline Operating Limits Curve Adjustment vs. Fluence),
3.6.2 (Minimum Temperature vs. Pressure for Pressure Tests),3.6.3 (Minimum Temperature vs. Pressure Mechanical Heatup or Cooldown Without the Core Critical),
l and 3.6.4 (Minimum Temperature vs. Pressure for Critical Core Operation). This will l provide current reactor pressure vessel property curves based on the latest empirical l testing of Monticello RPV sample material and NRC approved methodology. In the title l of Figure 3.6.4, the word " critical" has been added to clarify conditions under which this curve applies.
l Bases 3.6/4.6 (page 146)
Information removed from section 4.6 has been updated and added to this Bases section as stated above. Additionally, this section has been updated to reflect the latest RPV specimen testing.
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4 Safety Evaluation:
Proposed changes shift the curves in a bounding, slightly more conservative direction, thus restoring previously predicted safety factors. By revising the P-T curves to account for a slightly larger shift in RTsor, assurance of protection against non ductile RPV failure is assured. l l
K,, is the lower bound critical stress intensity factor for crack-arrest toughness of ferritic steels and represents a level below which rapidly running cracks can be arrested. it is considered to be a material property and has a unique value for a given material at a given temperature and fluence.
References 2 and 3 provide the basis for determining the hydrostatic test temperature such j
.. that applied stress intensity does not exceed K,/1.5. i Thus, compared to a straight )
L fracture mechanics evaluation, the code recommends a safety factor of 1.5. This safety
! factor is exclusive of other built-in conservatisms such as use of the curve " lower bound" i for ferritic steel, or margin included in'the adjusted reference temperature to account for error in Charpy data.
The following table lists safety factors determined from various combinations of existing 4 l and proposed curves, test temperatures, and exposures (cycles). To demonstrate that the new curves provide recommended safety factors, four test cases were calculated. '
Cases one and two compare end of cycle (EOC) 18 conditions using existing and proposed l Technical Specification curves. The plant is currently operating in cycle 19. The safety
! factor of 1.34 for case two indicates a higher temperature would be required to achieve the i recommended safety factor of 1.5. While measurements of surveillance specimens showed l the existing P-T curves to be less conservative than originally thought, application of a 1.5 safety factor in developing the curves was sufficiently conservative that the net effect resulted in an actual safety factor of greater than or equal to 1.34 in previous cycles. The L ' proposed P-T curves restore the safety factor to the recommended value of 1.5.
Cases three and four were performed to demonstrate continued compliance with code !
recommended safety factors as fluence increases, Minimum, Vessel Metal: ; Safety;
'cC l lCasef :: Data Sourcer 7
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' ycle; 44 RTemperature? ' t FactorV 1 Existing Curves 203 F EOC 18 1.53 2 Proposed Curves 203 F EOC 18 1.34 2 l 3 Proposed Curves 229 F EOC 19 1.51 e 4 Proposed Curves 246 F EOL 1.51
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- Actual EOC 18 tes1 temperature was 231*F as determined by proposed curves resulting in a safety 1
factor of > 1.5.
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'New data has resulted in an increase in our knowledge of RPV material, and the resulting proposed curves ensure continued compliance with the recommended safety factor of 1.5.
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RPV Surveillance Requirements The proposed change deletes from the Surveillance Requirements section of TS the requirement for one time tests that have been completed. A description of these tests i is added to the TS Bases for historical purposes. Since all requirements being deleted !
have been completed, there is no effect on plant safety.
! Removing the prescribed RPV sample withdrawal schedule allows flexibility and will l permit RPV sample removal at times determined to be most advantageous based on the limited number of surveillance specimen sets remaining in the Monticello RPV. j Since specimens already tested have been irradiated to EOL fluence levels, not removing a RPV surveillance specimen set at the next prescribed date has no effect on plant safety. The TS requirement of a the RPV surveillance program to conform to l ASTM E185-66 will remain in effect.
f SBLC Relief Valve Setpoint Testing The testing requirement of TS section 4.4.A.2.c for SBLC System relief valve setting is proposed to be deleted. The testing required by TS section 4.4.A.2.c is enveloped by the current testing performed by Monticello's IST Program. The IST program l implements an edition of ASME Code Section XI that has been approved in 10 CFR l Part 50.55a.Section XI provides testing requirements for all ASME Code Class 1,2, and 3 valves, including relief valves. The IST program requires all relief valves to be
- tested to their nameplate data setpoints. Any modification to a relief valve's nameplate l data is controlled by the plant's configuration control process which would ensure the requirements of ASME Section XI are invoked as required by TS section 3.15. The IST program required by TS 4.15 ensures the SBLC relief valves will be properly tested for operability.
l Determination of No Significant Hazards Consideration:
! Proposed change to the Operating License has been evaluated to determine whether it constitutes a significant hazards consideration as required by 10CFR Part 50, Section 50.91 using standards provided in Section 50.92. This analysis is provided below:
The proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.
RPV P-T Curve Changes it is proposed that P-T curves be revised to accommodate the shift in RT, determined using actual surveillance program data rather than generic data provided in Regulatory i
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. H Guide,1.99 Revision 2 (Radiation Embrittlement of Reactor Vessel Materials). The new P-T curves will increase the margins provided in the P-T limit curves against non-ductile l failure of the RPV. Regulatory Guide 1.99 Revision 2 encourages use of plant specific surveillance data as data becomes available.
Eliminating prescriptive requirements to remove a RPV test specimen sample at three l
fourths service life will result in an overall improvement in the RPV surveillance program l since the limited number of remaining surveillance samples will be removed at optimum l intervals. Therefore, proposed changes will neither significantly increase the probability ;
I or the consequences of an accident previously evaluated.
RPV Surveillance Requirements Deleting completed, one time surveillance requirements of SR section 4.6.B and incorporating a discussion of the results in the Bases is an administrative change and !
has no effect on probability or consequences of accidents.
SBLC Relief Valve Setpoint Testing The testing requirements of TS section 4.4.A.2.c are enveloped by the current testing performed by Monticello's IST Program, which implements ASME Code Section XI, approved by 10 CFR 50.55a. The IST program requires all relief valves to be tested to their nameplate data setpoints. Any modification to a relief valve's nameplate data is controlled by the plant's configuration control process which would ensure the requirements of ASME Section XI are invoked as required by TS section 3.15. The IST program required by TS 4.15 ensures the SBLC relief valves will be properly tested for operability. Therefore, revising section 4.4.A.2.c to remove specific setpoints does not increase the probability or consequences of an accident.
The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously analyzed.
RPV P-T Curve Change Updated RPV P-T limit curves will not create the possibility of a new or different kind of accident nor alter operational standards. New limits continue a system of operating bounds which are in place to prevent damage to reactor vessels during normal operating conditions including hydrostatic pressure and leakage testing, and anticipated transients. The updated P-T curves incorporate the results of RPV surveillance specimen testing utilizing criteria defined in RG 1.99, Revision 2. No change is being made to the way the P-T limits provide plant protection. No new modes of operation l are involved. The changes do not necessitate physical alteration of the plant.
l RPV Surveillance Requirements Deleting completed, one time surveillance requirements of section 4.6.B and incorporating a discussion of the results in the Bases is an administrative change and therefore has no effect on previously analyzed accidents.
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'SBLC Relief Valve Setpoint Testing The testing requirements of TS section 4.4.A.2.c are enveloped by the current testing performed by Monticello's IST Program, which implements ASME Code Section XI, approved by 10 CFR 50.55a. The IST program requires all relief valves to be tested to
. their nameplate data setpoints. Any modification to a relief valve's nameplate data is I
controlled by the plant's configuration control process which would ensure the requirements of ASME Section XI are invoked as required by TS section 3.15. The IST program required by TS 4.15 ensures the SBLC relief valves will be properly tested for operability. Therefore, revising section 4.4.A.2.c tc Semove specific setpoints does not create the possibility of a new or different kind of accident, from any accident previously analyzed.
The proposed amendment will not involve a significant reduction in the margin of safety.
RPV P-T Curve Change The proposed RPV P-T curve changes are designed to maintain the recommended safety factors specified in the ASME Boiler and Pressure Vessel Code, Section Ill, l Appendix G, and 10 CFR Part 50, Appendix G. The revised curves are based on current NRC guidelines utilizing actual RPV surveillance program test results. The proposed changes shift the curves in a slightly more conservative direction thus maintaining or increasing the previous margins of safety.
RPV Surveillance Requirements Deleting completed, one time surveillance requirements from Section 4.6.B and incorporating a discussion of the results in the Bases is an administrative change and has no effect on any margin of safety.
SBLC Relief Valve Setpoint Testing The testing requirements of TS section 4.4.A.2.c are enveloped by the current testing performed by Monticello's IST Program, which implements ASME Code Section XI, approved by 10 CFR 50.55a. The IST program requires all relief valves to be tested to their nameplate data setpoints. Any modification to a relief valve's nameplate data is controlled by the plant's configuration control process which would ensure the requirements of ASME Section XI are invoked as required by TS section 3.15. The IST program required by TS 4.15 ensures the SBLC relief valves will be properly tested for operability. Therefore, revising section 4.4.A.2.c to remove specific setpoints will not reduce the margin of safety.
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~ Environmental Assessment
- Northern States Power has evaluated the proposed changes and determined that:
- 1. The changes do not involve a significant hazards consideration.
- , 2. The changes do not involve a significant change in the type or significant increase in
-the amounts of any effluent that may be released offsite, or l
3; The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51, Section 51.22(b), and an environmental assessment of the proposed changes is not requirei References
! 1). " Review of the Test Results of Two Surveillance Capsules, and
- ' Recommendations for the Materials Properties and Pressure-Temperature l Curves to be Used for the Monticello Reactor Pressure Vessel," Report No. SIR-97-003, Structural Integrity Associates, Inc., May,1998. Submitted for NRC Staff review by NSP letter dated December 21,1998.
- 3) WRC Bulletin 175, "PVRC Recommendations on Toughness Requirements Sr Ferritic Materiais," PVRC Ad Hoc Group on Toughness Requirements, Welding Research Council, August 1972.
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