ML20206S091
| ML20206S091 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/17/1999 |
| From: | Hammer M NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20206S096 | List: |
| References | |
| TAC-MA4532, NUDOCS 9905210047 | |
| Download: ML20206S091 (12) | |
Text
Northem states Power Company Monticello Nuclear Generating Plant 2007 West County Road 75 Monticello, MN 55362 May 17,1999 10 CFR 50 Section 50.90 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Response to Request For Additional Information Regarding Proposed Amendment to the Pressure-Temperature Limits and Surveillance Capsule Withdrawal Schedule (TAC No. MA4532)
Ref.1 Letter from NRC to R.O. Anderson, NSP, *Monticello Nuclear Generating Plant -
Request for Additional Information Related to Submittal Requesting Revision of Pressure-Temperature Limits and Surveillance Capsule Withdrawal Schedule (TAC No. MA4532)" March 24,1999.
Ref. 2 Letter from NSP to US Nuclear Regulatory Commission, " License Amendment !
Request Dated December 31,1998 Revision of Reactor Pressure Vessel l l
Pressure-Temperature Limit Curves and Removal of Standby Liquid Control 1 '
Relief Valve Setpoint" December 31,1998.
Ref. 3 Conference call between NRC Staff and NSP on February 18,1999. j l CP Ref. 4 Conference call between NRC Staff and NSP on March 10,1999.
In two separate conference calls (Ref. 3 and 4) the NRC Staff and NSP discussed NRC questions regarding NSP's License Amendment Request (LAR) of December 31,1998 (Ref. 2) for the Monticello Nuclear Generating Plant (MNGP) to revise the existing reactor pressure vessel pressure-temperature curves. Based on these discussions, the NRC issued a Request for Additional Information (RAI) on March 24,1999 (Ref.1), to 9905210047 990517 PDR ADOCK 05000263 P PDR ,
6/17199 SIS J tlCENSDTech Specst A R's)R P %R*VNDT RAI 5 4 doc i
l USNRC NORTHERN STATES POWER COMPANY j May 17,1999 4 Page 2 document the conference calls and to complete its review of NSP's license amendment i request (Ref. 2). The responses to these questions are provided in Attachment 2.
Reference 1 established a target response date of April 23,1999. Due to a forced I mainternance outage at MNGP, on April 16,1999 NSP requested an extension which the NRC Staff agreed to.
NSP makes the following new commitment:
- 1. The USAR willbe revised to summarize the results of the surveillance capsule data obtained byirradiating the capsule to beyond End of Life (EOL) Reactor ;
Pressure Vessel (RPV) exposure. It will also state that the next surveillance l capsule will be removed during the 2003 refueling outage unless the results of the Integrated Surveillance Program (ISP) Focus Group determines that removal is unnecessary.
Please contact Sam Shirey, Monticello Licensing, at (612) 295-1449 if additional information is required.
fr1 4 f b Michael F. Hammer Site General Manager Monticello Nuclear Generating Plant c: Regional Administrator-lli, NRC NRR Project Manager, NRC Sr. Resident inspector, NRC State of Minnesota, Attn: Steve Minn J. Silberg, Esq.
Attachments Attachment 1 Affidavit to the US Nuclear Regulatory Commission Attachment 2 Response to Request For Additional Information Regarding Proposed Amendment to the Pressure-Temperature Limits and Surveillance Capsule Withdrawal Schedule Attachment 3 Structural Integrity Calculation No. NSP-21Q-302 Attachment 4 Structural Integrity Calculation No. NSP-21Q-304
i-UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 Response to Request For Additional Information Regarding Proposed Amendment to the Pressure-Temperature Limits and Surveillance Capsule Withdrawal Schedule (TAC No. MA4532)
Northem States Power Company, a Minnesota corporation, by letter dated May 17,1999 provides its response for the Monticello Nuclear Generating Plant to a US Nuclear Regulatory Commission (NRC) letter dated March 24,1999, with the subject "Monticello Nuc! ear Generating Plant - Request for Additional Information Related to Submittal Requesting Revision of Pressure-Temperature Limits and Surveillance Capsule Withdrawal Schedule (TAC No. MA4532)." This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY By b dL4 M 4 4 j ~
Michael F. Hammer Site General Manager Monticello Nuclear Generating Plant On this I day of v I 999 before me a notary public in and for ;
said County, personally appeated Michael P. Hammer, Site General Manager, Monticello Nuclear Generating Plant, and being first duly swom acknowledged that ho is l authorized to execute this document on behalf of Northem States Power Company, and
)
that to the best of his knowledge, Information, and belief the statements made in it are !
true. -
l
,A '
Samuel l. Shirey' ;--------;-;-_
Notary Public - Minnesota l SAMUEL 1. SHIREY Sherburne County ,
) NOTARY PUBUC MIIIIIt90th x-gy.=~
Response to Request for Additional Information Regarding Proposed Amendment to the Pressure - Temperature Limits and Surveillance Capsule Withdrawal Schedule Monticello Nuclear Generating Plant Docket No. 50-263
REFERENCES:
- 1. Battelle Report, BCL-585-84-2, Rev.1, " Examination, Testing and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Monticello Nuclear Generating Plant", November 5,1984.
- 2. Letter from Michael F. Hammer, NSP to U S Nuclear Regulatory Commission Document Control Desk, " Submittal of Report on Reactor Pressure Vessel Specimen Tests," December 21,1998.
- 3. Structural Integrity Calculation No. NSP-21Q-302, Revision 2, "Charpy V-Notch (CVN) Test Data Files for the Monticello RPV Plates, Curve Fits of the CVN Data, and Calculation of Plant Specific Chemistry Factors," May 11,1998. (Attachment 3)
- 4. Structural Integrity Calculation No. NSP-21Q-304, Revision 1, " Pressure Test, Non-Critical Core Operation, and Critical Core Operation P-T Curve Development," April 6,1998. (Attachment 4)
- 5. Letter from NRC to R.O. Anderson, NSP," Closeout of Response to Generic Letter 92-01, Revision 1, Supplement 1 for Monticello Nuclear Generating Plant (TAC No.
M92699)" December 9,1996.
- 6. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice inspection of Nuclear Power Plant Components,[Nonmandatory] Appendix G, " Fracture Toughness Requirements," 1989 Edition.
- 7. Welding Research Council Bulletin 175,"PVRC Recommendations on Toughness Requirements for Ferritic Materials," PVRC Ad Hoc Group on Toughness Requirements, August 1972.
- 8. NRC Generic Letter 92-01, Revision 1, " Reactor Vessel Structural integrity,10 CFR 50.54(f), March 6,1992 NRC QUESTIONS AND NSP RESPONSES:
- 1. Provide a complete descdption of the methodology used at Monticello Nuclear Generating Plant (MNGP) for the determination of reactorpressure vessel (RPV) fluences. This shouldinclude information on the neutron transport code models (their spatial mesh refinement, angular quadrature, energy group divisions, l
NSP Resp:nse to 3/24/99 NRC Request far Additi:n Iinformation 1 differencing schemes, cross-sectionallibrart.s, etc.) and on the analysis and application of dosimeter wire measuremerus for establishing best-estimate fluences. In particular, this should addre'is if and/or how the dosimeter wire measurements are used to modify the msults of the neutron transport code ,
analysis. The staff notes that pmliminrry staff guidance on determining RPV i fluences was published as draft Regulatory Guide (RG)-1053.
Response: The methodology used at Monticello Nuclear Generating Plant :
(MNGP) for the determination of reactor pressure vessel (RPV) i fluences is based on the industry standard 1/4T fluence. A one- <
time calculation to provide a proportionality factor to convert from flux wire dosimetry fluences to vessel 1/4T fluences was performed 1 as part of Ref.1. The following description of the calculation characteristics for determining this proportionality factor is an excerpt from Ref.1.
"The energy and spatial distribution of neutron flux in !
the reactor were calculated using the DOT 3.5 computer i program. DOT solves the Boltzrnan transport equation in two-dimensional geometry using the method of discrete ordinates. Balance equations are solved for the density of particles moving along discrete directions in .
each cell of a two-dimensional spatial mesh. I Anisotropic scattering is treated using a Legendre ,
expansion of arbitrary order. 1 The two-dimensional geometry that was used to model the Monticello reactor is shown in Figure 8 [of Ref.1]. l There are 17 circumferential meshes and 51 radial meshes. Capsule [G-)1 [ removed in November of 1981) includes circumferential meshes 7 and 8 and radial meshes 41,42, and 43. Third order scatter was used (P3 ) and 48 angular directions of neutron travel (24 positive and 24 negative) were used (Ss quadrature).
Neutron energies were divided into 22 groups with energies from 14.9 MeV to 0.01 eV. The 22 group structure is that of the RSIC Data Library DLCICask, and neutron absorption, scattering, and fission cross sections used are those supplied by this library."
Subsequent determination of the reactor vessel peak fluence at time t is made using the following equation:
Fi = (P/Er) x Ei (Eq 1)
Where:
Page 2 of 9
NSP Response to 3/24/99 NRC Requ;ct f:r Additisn:l Inf:rmiti::n 1
Ft = Peak vessel 1/4T fluence at time t P = Peak vessel 1/4T fluence at the time the flux !
wires were removed from tlie vessel, determined by multiply':ig the calculated peak vessel surface fluence (attenuated to the 1/4T fluence) by the ratio of it.: Woal flux wire fluence to the calculated flux wire fluence, in n/cm2, Er = Integrated core exposure when the flux wires were removed from the vessel, in Effective Full Power Years.
Et = Integrated core exposure at time t, in Effective Full Power Years.
For cycles 19 and beyond, an adjustment to Et was made to account for the 6.3% increase in rated thermal power that was l implemented in October,1998. This is done by multiplying the '
effective full power years accumulated after October,1998 by !
1.063. l
- 2. Summarize the results of the last end-of-cycle fluence analysis that was performed at MNGP to demonstrate how the methodology describedin (1) is applied for the purpose of establishing RPV /eak test pressure-temperature (P-T) ;
limits. '
Response: The last end-of-cycle fluence analysis (EOC-19) was calculated as follows: .
P = 7.20 x 10'7 Neutrons /cm2 Er = 7.08 Effective Full Power Years Et = 21.08 Effective Full Power Years F = (P/Er) x Et (Eq 1)
F = (7.20 x 10'7/7.08) x 21.0L i
2 Ft = 2.14 x 10'8 neutrons /cm l Page 3 of 9 l
7 NSP Respon 3 to 3I24/99 NRC Request far AdditienslInfcrm tinn Pressure and temperature limits are established using a calculated fluence as determined above, in conjunction with Monticello Technical Specification Figures 3.6.1, " Core Beltline Operating Limits Curve Adjustment vs. Fluence" and Figure 3.6.2, " Minimum Temperature vs. Pressure for Pressure Tests."
- a. How many surveillance capsules remain to be withdrawn from the MNGP RPV and when are these capsules scheduled to be removed?
Response: Currently, two capsules remain in the Monticello RPV. One capsule is due for removal during the year 2000 refueling outage which corresponds to three fourths service life as specified in Technical Specification Section 4.6.B.2. Our December 31,1998 License Amendment Request (LAR) requests deletion of this requirement.
No plans or TS requirements currently exist for removing the final specimen capsule,
- b. What basis or cdteria will be used to determine when the remaining surveillance capsules willbe withdrawn? Licensees have been permitted to remove surveillance capsule requirements from their technical specifications (TS), but in those cases a definitive program has been addressed in the facility's Updated Safety Analysis Report to ensure that the facility continues to comply with the provisions of Title 10 of the Code of Federal Regulations Part 50, Appendix H.
Response: 10 CFR 50, Appendix H states "The design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased." As stated in TS section !
4.6.B.2, for Monticello this is the 1966 edition. This edition recommends that sets of specimens be withdrawn at three or more separate times.
The proposed LAR change to delete the three-fourths service life surveillance capsule removal requirement is {
based on the following:
Reference 2 discussed removal, testing, re-irradiation iri the Prairie Island (PI) RPV, and further testing of the Page 4 of 9 i
1 l
NSP R::pon23 to 3/24/99 NRC Requ:ct far Additi n:lInformation surveillance capsule removed at "one fourth ... service life."
as required by TS section 4.6.B.2. While in the Pl RPV, the capsule received exposure equivalent to 40 years operation in the Monticello RPV. This information was used to produce the figures submitted in Ref. 2. Since this encompasses the remaining Monticello RPV exposure, removal of a capsule at the "three-fourths service life" is
, unnecessary.
Monticello is currently participating with the Integrated Surveillance Program (ISP) Focus Group which has been i formed to improve the surveillance program currently being implemented by the BWR industry. This program is expected to review the existing material surveillance ;
programs within the industry and verify the programs are l adequately monitoring embrittlement in BWR vessels.
Implementation of the ISP is not expected until the end of 2000. Monticello's implementation of the integrated BWR surveillance program will not only ensure compliance with the intent of 10 CFR 50, Appendix H and ASTM-E-185-66, but will also provide data that better represent the BWR materials and operating conditions.
NSP will use the ISP report as a basis for determining if and when the remaining surveillance capsules will be withdrawn consistent with the following commitment .
The USAR will be revised to summarize the results of the surveillance capsule data obtained by irradiating the capsule to beyond End of Life (EOL) Reactor Pressure Vessel (RPV) exposure. It will also state that the next surveillance capsule will be removed during the 2003 refueling outage unless the results of the Integrated Surveillance Program (ISP) Focus Group determines that removal is unnecessary.
- c. Explain what actions MNGP is taking to address the lack of unirradiated baseline data foryour RPV surveillance weld.
Response: To address the lack of unirradiated baseline data for the MNGP RPV surveillance weld data, NSP is participating with the ISP. This program will examine all BWR data and potentially provide a more accurate representation of the unirradiated RTuor value for RPV surveillance welds. If a more accurate value of RTuor is obtained, the data will be utilized as appropriate. It should be noted, however, that Page 5 of 9 1
NSP R:cponna to 3/24/99 NRC R:quoct fsr Additirn:l Infarmntion since plate material is more limiting than weld material, lack of weld material data makes the impact less significant.
Prior to the advent of the ISP, the NRC had addressed the issue of lack of RPV material data with Generic Letter 92-01, Revision 1, " Reactor Vessel Structural Integrity,10 CFR 50.54(f)" (Ref. 8). In closecut of Generic Letter 92-01, Revision 1 (Ref. 5), the NRC staff acknowledged the lack of data on RPV materials and stated:
... Your response indicates that NSP has performed some additional reviews of the pertinent Owners Group databases and has not found any additional data regarding the best-estimate chemistries for the RPV materials and surveillance capsules at the Monticello Nuclear Generating Plant. Since NSP has submitted the requested information and has indicated that the previously submitted evaluations remain valid, the staff considers the RPV integrity data for Monticello to be complete at this time. The staff therefore concludes that no additional information regarding the structural integrity of the RPV at Monticello is available at this time, and that your efforts regarding GL 92-01, Supp.1 are complete."
NSP agrees with this acknowledgement, but will continue to work with the ISP.
- 4. Regarding the revised Figures 3.6.1, 3.6.2, 3.6.3, and 3.6.4 in your submittal package:
- a. Confirm that the margin term (proposed in your submittal as 34 'F) used to calculate the limiting adjusted reference temperature after a period of operation is directly incorporated into Figure 3.6.1 for determining 'RTuor Shift."
Response: The 34 F margin term used to calculate the limiting adjusted reference temperature after a period of operation is directly incorporated into Figure 3.6.1. Reference 3 provides an explanation for the calculation of the ARTNor versus Fluence curve. Figure 6-4 of this calculation shows the 34 F margin term being added to the ARTworfor developing the proposed Technical Specification Figure 3.6.1. The 34*F margin applies over the entire fluence range of Figure 3.6.1.
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NSP Rreponse to 3/24I99 NRC R:qu:st far Additi:nal information
- b. Figures 3.6.2, 3.6.3, and 3.6.4 indicate that the temperature scale on the vertical axis is relative to the " minimum vessel metal temperature."
- 1. Explain how the instrumentation used to monitor the conditions in the RPVis used to determine the "vesselmetal temperature."
Response: External reactor vessel metal temperature is determined by monitoring thermocouples at ten diverse locations. This monitoring provides a cross section of vessel temperatures over the entire surface of the vessel. The bottom head and vessel shell adjacent to flange thermocouples provide the most diverse temperatures, and therefore are the primary remote-from-core-beltline-region temperature locations monitored for brittle fracture limits.
Additionally, reactor recirculation loop temperatures are used as the primary temperature indicator for the core beltline region.
- 2. Indicate how instrument uncertainties are included when relating plant conditions to the P-Tlimits.
Response: To account for instrument loop uncertainties a 5'F margin is added to the required minimum temperature. The 5*F margin is based on vendor specified accuracy of the thermocouples and display system,
- c. For Figures 3.6.3 and 3.6.4, explain:
- 1. What heatup or cooldown rates these curves are valid for (e.g., less than or equal to 10047hr) and indicate where the limits on how rapidly the MGNP RPV can heatup or cooldown are located.
Response: TS Figures 3.6.3 and 3.6.4 were developed for a cooldown rate of 100 F/hr, as discussed in Section 4.0 of Reference 4. For lower cooldown rates, the thermal stress intensity factor (Kn) decreases, which leads to a higher allowable pressure. The curves are also applicable for heatup rates that are equal to or less than 100*F/hr, as discussed in the response to Question 4.c.4 below. Therefore, the curves are valid for all heatup and cooldown rates less than or equal to 100*F/hr.
Page 7 of 9
NSP R=ponse to 3/24/99 NRC R:qurt far Additi:ncl infermition The limits for controlling heatup and cooldown rates for the reactor vessel are located in Section 3.6.A.1 of the Monticello Technical Specifications.
- 2. Explain how the through-wall thermal gradients in the RPV are determined when defining the heatup and cooldown P-Tlimits.
Response: Through-wall thermal gradients in the RPV were determined using classical heat transfer theory for one-dimensional heat transfer in a cylinder, as described in Section 4.0 of Reference 4. This methodology readily provides a temperature solution as a function of time for any point through the wall thickness. The heat transfer analysis was conducted to establish the following two items: (1) the temperature drop between the reactor fluid and the crack tip (i.e., at the 1/4T location), and (2) the through-wall temperature drop for computing Kir in accordance with Figure G-2214-2 of the ASME Code,Section XI, Appendix G (Ref. 6). The temperature range used for this cooldown analysis was from 550*F to 50*F. The associated thermal stresses are at their maximum when the temperature is decreasing at its maximum rate of 100*F/hr.
- 3. Explain how the contribution of the thermal stresses developed during heatup and cooldown was calculated and provide the Kn-values used for determining the heatup and cooldown curve.
Response: From the temperature solution described in 4.c.2 above, Kir was computed using Figure G-2214-2 of the ASME Code,Section XI, Appendix G (Ref. 6) for the beltline and bottom head regions. For the feedwater nozzle, Figure 4-5 of WRC Bulletin 175 (Ref. 7) was utilized. The values obtained for Kir were 7.2 ksiV(in),
10.6 ksiV(in), and 6.4 ksiV(in) for the beltline, bottom head and feedwater nozzle regions, respectively. For a 4 complete explanation, refer to Section 4.0 of Reference 4, " Thermal Stress Intensity Factor Analysis." l
- 4. The staff has noted that the methodology for establishing heatup and cooldown P-Tlimits in your TS is based on an evaluation of the i Page 8 of 9 '
NSP Rrponse to 3/24/99 NRC Request far Additi:nr1 information 1/4Tflawlocation. During heatup, since the tensile stresses occur on the outside of the vessel, the 3f4Tflewlocation may be limiting depending upon the heatup rate, the through-wall temperature
. gradient, the water temperature, and the RTuor of the 1/4T and 3/4Tlocation.
Is the 3/4T flaw limiting at any point during the most severe heatup transient permitted byyour TS? If so, explain how the methodology proposed in your TS results in the generation of P-Tlimits which are bounding for the most severe heatup transient permitted by your TS.
l Response: The 3/4T flaw is not limiting for any heatup or I cooldown transient permitted by the Monticello Technical Specifications (less than or equal to 100'F/hr). Outside surface thermal stresses for a 100*F/hr heatup are approximately the same as inside surface thermal stresses during a 100*F/hr cooldown. The same is true for pressure stresses in a thin-walled cylinder. However, fluence effects are significantly different at the 1/4T and 3/4T locations due to attenuation through the vessel wall. As a result, the adjusted reference temperature (ART) is significantly greater at the 1/4T location, which results in a lower allowable material toughness, Kic, at this location. The lower value of Kic leads to a lower allowable pressure. Therefore, evaluation of a 100*F/hr cooldown for the 1/4T location bounds a '
100*F/hr heatup at the 3/4T location. This conclusion is further supported by the fact that the value of Kic is more limiting a lower temperatures, such as those associated with cooldown transients. Based on this, ,
the MNGP P-T curves were developed for a limiting I 100*F/hr cooldown transient, and they may be applied i for both heatup and cooldown rates equal to or less than 100*F/hr. I 1 '
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