ML20248L401
| ML20248L401 | |
| Person / Time | |
|---|---|
| Site: | Monticello, Prairie Island  |
| Issue date: | 03/06/1998 |
| From: | Hammer M NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20248L403 | List: |
| References | |
| TAC-M96238, NUDOCS 9803200169 | |
| Download: ML20248L401 (33) | |
Text
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Mr Northem States Power Company 1
Monticello Nuclear Generating Plant 2807 West Hwy 75 Monticello. Minnesota 55362-9637 March 6,1998 US Nuclear Regulatory Commission i
Attn: Document Control Desk l
Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 l
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Response to February 11,1998 Request for Additional information (RAI) on l
Monticello Power Rerate License Amendment (TAC No. M96238) l Ref.1 Letter from NRC to R.O. Anderson, NSP, "Monticello Nuclear Generating Plant -
Request for Aoditional Information on License Amendment Request Entitled
' Supporting the Monticello Nuclear Generating Plant (MNGP) Power Rerate Program'(TAC No. M96238)"
By letter dated February 11,1998 (Ref.1), staff provided a Request for Additionai information (RAl) to complete its review of NSP's license amendment request for the Monticello Nuclear Generating Plant (MNGP) Power Rerate Program.
NSP's partial response to the subject RAI is attached. As previously discussed with the staff, the response to question 6 is under development and will be submitted under separate cover. Attachments 1 and 2 contain affidavits, and Attachments 3 through 6 contain primary and supplementary information to the NSP responses herein. Certain responses contain references to calculations. These calculations are current as of the date of this submittal. Future revisions to these calculations, if any, will be available onsite for staff review.
Please contact Joel Beres, Monticello Licensing, at (612) 295-1436 if additional information is required.
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Michael F. Hammer j) l Plant Manager Monticello N0 clear Generating Plant UlllE. I.M. lu11111!!
9003200169 980306 PDR ADOCK 05000263 p
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Regional Administrator-lil, NRC NRR Project Manager, NRC
- Sr. Resident inspector, NRC State of Minnesota, Attn: Kris Sanda J. Silberg, Esq.
Attachments Affidavit to the US Nuclear Regulatory Commission Affidavits to GE Rerate Documents Submitted on December 5, 1997 NSP Response to Staff Power Rerate RAI dated February 11, 1998 Radiological Analyses of Design Bases Accidents Annual Report Review of Meteorological Data 1991 Monticello and Prairie Island Stations Revision to Proposed Changes to MNGP Technical Specification Page 39 f
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UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 Response to February 11,1998 Request for Additional Information (RAI) on Montice!!o Power Rerate License Amendment (TAC No. M96238)
Northem States Power Company, a Minnesota corporation, by letter dated March 6, 1998 provides its response for the Monticello Nuclear Generating Plant to a US Nuclear Regulatory Commission (NRC) letter dated February 11,1998, with the subject
Monticello Nuclear Generating Plant - Request for Additional information on License i
Amendment Request Entitled ' Supporting the Monticello Nuclear Generating Plant (MNGP) Power Rerate Program' (TAC No. M96238)." This letter contains no restricted or other defense information.
NORTHERN STATES POVER COMPANY By hb f1Ay AAA]) )
Michael F. lWmm'e'r~
Plant Manager Monticello Nuclear Generating Plant On this k day of
$Q(ch 19 9 before me a notary public in and for said County, personally appeared ' Michael F. Hammer, Plant Manager, Monticello
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Nuclear Generating Plant, and being first duly swom acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, and that to the best of his knowledge, information, and belief the statements made in it are true.
Colleen Hhnn'on' COLLEEN A.HANNON Notary Public-Minnesota sw errusus.esumena Sherbume County
,..at ess a a o.Jan.st,ases My Commission Expires January 31,2000 l
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Original Amdavits to GE Rorate Documents Submitted on December 5,1997 l
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l General Electric Company AFFIDAVIT I, George B. Stramback, being duly sworn, depose and state as follows:
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(1) I am Project Manager, R.egulatory Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in the GE proprietary report NEDC-32514P, Revision 1, Monticello SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis, Class III (GE Proprietary Information), dated October 1997. The proprietary information is delineated by bars marked in the margin adjacent to the l
specific material.
(3) In making this application for withholding of proprietary information of which it is l
the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which j
exemption from disclosure is here sought is all " confidential commercial information", and some portions also qualify under the narrower definition of" trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory l
Commission. 975F2d871 (DC Cir.1992), and Public Citizen Health Research Group l
- v. FDA,704F2dl280 (DC Cir.1983).
(4) Some examples of categories of information which fit into the definition of l
proprietary information are:
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a.
Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention ofits use by General Electric's competitors i
without license from General Electne constitutes a competitive economic l
advantage over other companies; l
b.
Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; l
l GBS-97-7-afSAFER4. doc Affidavit Page 1 J
Information which reveals cost or price information, production capacities, c.
budget levels, or commercial strategies of General Electric, its customers, or its suppliers; d.'
Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial t
value to General Electric; 1
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Information which discloses patentable subject matter for which it may be e.
L desirable to obtain patent protection.
The information sought to be withheld is consider $d to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.
l-(5). The information sought to be withheld is being submitted to NRC in confidence.
The information is of a sort customarily held in confidence by GE, and is in fact so I
L held. The information sought to be withheld has, to the best of my knowledge and j
l belief, consistently been held in confidence by GE, no public disclosure has been l.
made, and it is not available in public sources. All disclosures to third parties j
including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide ~for i
maintenance of the information in confidence';' Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.
(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination l.
of the accuracy of the proprietary designation. Disclosures outside GE are limited to L
regulatory bodies, customers, and potential customers, and their agents, suppliers, E
and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of analytical models, methods and processes, including computer codes, which GE has developed, obtained NRC approval of, and applied to perform evaluations of the loss-of-coolant accident for the BWR.
l GBS-97-7 af5AFER4. doc Affidavit Page 2
_-_-________-__________-____-___-___________--___-__-______-__-_A
The development and approval of the BWR loss-of-coolant accident analysis computer codes used in this analysis was achieved at a significant cost, on the order of several million dollars, to GE.
The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience
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database that constitutes a major GE asset.
(9) Public disclosure of the information sought' to be withheld is likely to 'cause l-substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its commercial.value extends beyond the original development cost. The value of the technology base goes beyond the l
extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.
The research, development, en' ineering, analytical and NRC review costs comprise g
a substantial investment of time and money by GE.
The precise value of the expertise to devise an evaluation process and apply the conect analytical methodology is difficult to quantify, but it clearly is substantial.
GE's competitive advantage will be lost ifits compet' tors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
The value of this information to GE would be lost if the information were disclosed q
to the public. Making such information available to competitors without their
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having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.
I-GBS 97-7 afSAFER4. doc Affidavit Page 3
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STATE OF CALIFORNIA
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COUNTY OF SANTA CLARA
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George B. Stramback, being duly sworn, deposes and says:
That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.
day of d/M 1997.
N Executed at San Jose, California, this D bHVLb-Ge6rge B. Stramback General Electric Company Subscribed and sworn before me this //-day of Md7##E
- 1997, i
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Wgycg. Om Notary Public, State ofCalifornia y
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' GBS-97-7 atSAFER4. doc Affidavit Page 4 l
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,Y General Electric Company AFFIDAVIT I, George B. Stramback, being duly sworn, depose and state as follows:
(1) I am Project Manager, Regulatory Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) wh'ich is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in the GE proprietary report NEDC-32546P, Power Rerate Safety Analysis Report for Monticello Nuclear Generating Plant, Revision 1, Class III (GE Proprietary Infonnation), dated December 1997. This document, taken as a whole, constitutes a proprietary compilation of information, some of it also independently proprietary, prepared by the General Electric Company. The independently proprietary elements are deliner.ted by bars marked in the margin adjacent to the specific material.
(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulathns 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information", and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enerev Project v. Nuclear Regulatory Commission,975F2d871 (DC Cir.1992), and Public Citi7en Health Research Group
- v. FDA,704F2d1280 (DC Cir.1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
Information that discloses a process, method,'or apparatus, including supporting a.
data and analyses, where prevention ofits use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other compames; l
e GDS-97-8-afrate01. doc Affidavit Page 1
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b.
Information which, if used by a competitor, would reduce his expenditure of-resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; Information which reveals cost-or price information, production capacities, c.
- budget levels, or commercial strategies of General Electric, its customers, or its suppliers; d.
Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs,' of potential commercial -
value to General Electric; Information which discloses patentable subject matter for which it may be e.
desirable to obtain patent protection.
Both the compilation as a whole and the marked independently proprietary elements.
incorporated in that compilation are considered proprietary for the reason described in items (4)a. and (4)b., above.
- (5). The information sought to be withheld is being submitted to NRC in confidence.
. hat information (both the entire body ofinformation in the form compiled in this T
document, and the marked individual proprietary elements) is of a sort customarily ~
i held in confidence by GE, and has, to the best of my knowledge, consistently been held in confidence by GE, has not been publicly disclosed, and is not available in -
public sources. All disclosures to third parties including any required transmittal to
-NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in d
confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
'(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value' and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.
l (7) The procedu : for approval of external release of such a document typically requires
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review by the staff manager, project manager, principal scientist or other equivalent j
authority', by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, -
and licensees, and others with a legitimate need for the information, and then only in eccordance with appropriate regulatory provisions or proprietary agreementr.
y GBS-97-8-afrase01. doc Affidavit Page 2 l
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(8) De information identified by bars in the margin is classified as proprietary because l
it contains detailed results and conclusions from these; evaluations, utilizing analytical models and methods, including computer codes, which GE has developed, obtained NRC approval ot, and' applied to perform evaluations of. transient and accident events in the GE Boiling Water Reactor ("BWR"). The development and
. approval of these system, component, and thermal hydraulic models and computer codes was achieved at a significant cost to GE, on the' order of several million
. dollars.-
The remainder of the information identified in paragraph (2), above, is classified as proprietary because it constitutes a confidential compilation of information,
-including detailed results of analytical models, methods, and proceses, including computer codes, and conclusions from these applications, which represent, as. a whole, an integrated process or approach which GE has deveicped, obtained NRC approval of, and applied to perform evaluations of the safety-significant changes necessary to demonstrate the regulatory acceptability of a given increase in licensed )
power output for a GE BWR. The developinent and approval of this overall-approach was achieved at a significant additional cost to GE, in excess of a million dollars,.over and above the very large cost of developing the underlying individual
. proprietary analyses.
To effect a change to the licensing basis of a plant requires a thorough evaluation of the impact of the change on all postulated accident and transient events, and all other regulatory requirements and commitments included in the plant's FSAR. The analytical process to perform and document these evaluations for a proposed power rerate was developed at a substantial investment in GE resources and expertise. The results from these evaluations identify those BWR systems and components, and those postulated events, which are impacted by the changes required to accommodate operation at increased power levels, and, just as importantly, those which are not so impacted, and the technical justification for not considering the.
latter in changing the licensing basis. The scope thus determined fo ms the basis for.
GE's offerings to support utilities in both performing analyses and providing
- licensing consulting services. Clearly, the scope and magnitude of effort of any attempt by a competitor to effect a 'similar licensing change can: be narrowed considerably based upon these results. Having invested in the initial evaluations and' developed the solution strategy and process described in the subject document GE derives an important competitive advantage in selling and performing these ~ services.
However, the mere knowledge of the impact on each system and component reveals the process, and provides a guide to the solution strategy.
(9) Public disclosure of the information sought to be withheld is likely to cause L
substantial harm to GE's competitive position and foreclose or reduce the availability.
of profit-making opportunities. The information is part of GE's comprehensive BWR technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive GBS-97 8-afrate01. doc' Afridavit Page 3 l
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physical database and analytical methodology and includes development of the
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expertise to determine and apply the appropriate evaluation process. In addition, the L
technology' base includes the value derived from providing analyses done with NRC-approved methods, including justifications for not including cenain analyses in applications to change the licensing basis.
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GE's competitive advantage will be lost ifits competitors are able to use the results of the GE experience to avoid fruitless avenues, or to normalize or verify their own b
process, or to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. In particular, the specific areas addressed by any document and submittal to support a change in the safety or licensing bases.
of the plant will clearly reveal those areas where detailed ~ evaluations must be performed and specific analyses revised, and also, by omission, reveal those areas not so affected.
While some of the underlying analyses, and some of the gross structure of the process, may at various times have been publicly revealed, enough of both the analyses and the detailed structural framework of the process have been held in confidence that this information, in this compiled form, continues to have great competitive value to GE. This value would be lost if the information as a whole, in the context and level of detail provided in the subject GE document, were to be disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources, including l
that required to determine the areas that are not affected by a power rerate and are '
therefore blind alleys, would unfairly provide competitors with'a windfall, and 1
deprive GE of the' opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing its analytical process.
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GBS-97 8-afrate01. doc Affidavit Page 4 -
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- STATE OF CALIFORNIA
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ss:
i COUNTY OF SANTA CLARA
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l George B. Stramback, being duly sworn, deposes and says:
That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.,,
Executed at San Jose, California, this _h4[ day of
/ m M 1997.
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'" Gedrge B. Sframback General Electric Company 1
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Subscribed and sworn before me this 8 day of [dd"d'#8e.e 1997, i
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I WA#A hdA 0 Notary Public, State of California ANNA HANUN COMM. #1030164 5
NOTARY PUBLKX:AUFORNIA 9 SAN FRANCISCO COLNTY 1
Wy comm. Expires June 19,1998 y w
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GBS-97-B-afratc01. doc Affidavit Page 5
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NSP Response to Staff Power Rorate RAI dated February 11,1993 l
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On pages 24 and 25 of the submittal dated September 5,1997, it is stated that "the required time to initiate manual depressurization of the reactor vessel was changed from 26 minutes to 23 minutes...the time required to initiate SBLC
[ standby liquid control system] changes from 21 minutes to about 13.
minutes...Although required times to accomplish manual operator actions are decreased as illustrated above, there is still adequate time to accomplish these actions, and an expectation [ emphasis added] that the actions would indeed be accomplished.' The licensee noted that the two sub}ect operator actions are examples of operator actions most sensitive to power uprate.
Please provide the bases for assurance (e.g., simulator observations and l
licensee assessments) that operators can perform these actions in the required response times.
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' NSP Response
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The expected operator response times for these particular actions are based on an assessment performed by a PRA engineer with shift supervisor experience at Monticello. The assessment was reviewed by a PRA engineer with experience as a simulator instructor at Monticello. The necessary operator actions are simple, and the operators are frequently trained on the scenarios. For both of-these cases, only switch manipulation from the control room is required - one switch manipulation for SBLC initiation and up to three switch manipulations for manual depressurization.
NSP recently conducted simulator observations using an actual operating crew to -
confirm these assessments. This observation determined that operator response times for the manual depressurization initiation and the SBLC initiation are well within the response times cited above.
2.
NSP's response to question 36 in the submittal dated September 5,1997, provides a statement which willbe added to the revisedlicense amendment. The statement should clearly and specifically indicate that:
Forpower rerate, GE setpoint methodology provided in NEDC-31336, General Electric Setpoint Methodology, is used in establishing setpoints.
F NSP Response The setpoint program at Monticello is based on the generic GE setpoint methodology, and this methodology was used in establishing power rerote setpoints. NSP does not object to this statement. MNGP Technical Specification bases page 39 and the associated markup has been revised to reflect this change. These pages are contained in Attachment 6.
NSP requests that the staff not consider the effect of this statement to preclude the use of a well-justified and plant unique setpoint method, such as actual plant i
data, which is equivalent to the GE setpoint methodology and consistent with the
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licensing basis, s
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l-3.
NSP's response to question 38 provided that the setpoint for Condenser Low Vacuum has been revised from 23.25" Hg to 22.25" Hg. However, Exhibit B, TS l
. Table 3.1.1, page 28, and Exhibit A, pages A-6 and A-31, indicate that the l
Condenser Low Vacuum setpoint has been revised from 23"Hg to 22" Hg.
Clarify this discrepancy.
NSP Response This discrepancy arises from the various definitions of terms associated with setpoint methodology. The definition of a limiting trip setting is different than the definition of a nominalsetpoint. NSP's response to question 38 was in regard to nominal setpoints.
The values in Table 3.1.1 of the Monticello Technical Specifications are defined as Limiting Trip Settings. In the case of the condenser low vacuum trip, NSP is requesting a change in the limiting trip setting from the current 23" Hg to 22" Hg.
The nominal trip setpoint is set at a more conservative value than the Limiting Trip Setting. This bias is added to assure that the plant actual settings do not exceed the associated Limiting Trip Settings. In the case of the condenser low vacuum, the previous nominal trip setpoint of 23.25" will be changed to 22.25" Hg.
4.
Exhibit A, page A-9, item 2b, states that TS Table 3.1.1, page 30, item d, willbe revised to state"... when the reactor thermalpower <45% (798.75 MWt)."
Howevct, Exhibit B and Exhibit C indicate this to be, ".... when the reactor thermal poweris s 45% (798.75 MM). Clarify this discrepancy.
i NSP Response This discrepancy is due to a typographical error which NSP regrets. Page A-9 should read s 45%. Exhibit B and C are correct. This error does not affect the marked up and revised technical specification pages.
5.
Exhit.it B and Exhibit C of the TS amendment request provide more changes to the TS than listed in Table 5.1 of Exhibit E. For example, the following changes were notidentifiedin Table 5.1:
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a)
Turbine condenserlow vacuum (TS Table 3.1.1)
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Low pressure core cooling pumps discharge pressure interlock (TS Table 3.2.2, item c.3).
c)
Reactorpressure interlock (TS Table 3.2.1, item 8.a) l Explain why these changes were not included in Table 5.1 of Exhibit E.
NSP Response The executive summary of Exhibit E includes a statement that the report (read Exhibit E) summarizes the results of evaluations needed to justify a license amendment to allow for rersted power operation. As stated in Section 5.1.3 of 3
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Exhibit E, Table 5-1 provides a summary of the analytical limits used in rerate i
setpoint analyses. The analytical limits listed are directly related to operation at l
power rerate conditions, and evaluation of these limits, typically through event l
analyses, is needed to justify rerate operation.
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The change to the condenser vacuum setpoint is directly related to changes in operation due to rerate conditions. This change could have been included in Tr.ble 5-1.
As part of the rerate evaluation, all affected Monticello plant systems were
' analyzed for operation at power rerate. During these analyses, NSP identified certain changes to setpoints associated with equipment performance. These changes, although not fundamentally required to implement power rerate, were prudent to implement due to increased accuracy or were more representative of actual plant conditions and were justifiable at rerate operating conditions. Where applicable these latter changes have been accounted for in rerate analyses.
Changes b) and c) identified above belong to this set of changes.. NSP recognizes that these distinctions may not have been readily apparent.
Exhibit A contains descriptions of all changes to the technical specifications associated with power rerate including the equipment setpoint changes described above.
6.
For each component / equipment type (or one representative / bounding example of a component / equipment type) where expected environmental conditions at the uprate powerlevel exceeds the environmental conditions tested to, provide the following:
a)
Description showing the relationship between environmental conditions (i.e., temperature) tested to, the expected environmental conditions at l
current powerlevels (if applicable /available), and the expected i
environmental conditions at power uprate level from time 0 (i.e., initiation of accident) to the time the component / equipment type is required to remain operable for post LOCA [ loss-of-coolant-accident] operation.
b)
Evaluation demonstrating qualification for each segment of the power level temperature response that is not enveloped by the environmental conditions (i.e., temperature) tested to.
. c)
Where (orif) margins derived through the use of the Anhenius t
1 methodology are utilized as part of the basis for concluding continued i
qualification, provide the Arrhenius calculation at ths current (if applicablefavailable) and uprate powerlevels. Define the margins 1
available for the current and uprate ponerlevels and describe andjustify the reduced margin for the uprate ponerlevel.
d)
Provide MNGP Calculation CA 97-176 which shows that the equivalent temperature exposure time for the EQ [ environmental qualification) 4
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temperature evaluation profilt nxceeds the equivalent temperature exposure time for the DBA [ design basis accident] temperature pro!?!e.
l' NSP Response I
NSP will provide the response to question 6 in a subsequent submittal.
7.
. The Monticello normal design configuration includes provisions for the automatic
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fast bus transfer of the offsite power source through the 2R to the offsite power source through the 1R transformer. : As a result ofpower uprate and other design changes, loading on the 1R transformer has increased. To accommodate this increased loading end to assure acceptable voltages forsafety system loads, the licensee has derated the 1R transformer and implemented design provisions (when automatic transfer occurs) to trip both recirc MGs [motorgenerators], to trip both circulating waterpumps, and to re-energize only one feed pump.
a)
Describe design, operational testing, technical specification requirements, and/or otherprovisions that assum a trip of both Recirc MGs, a trip of both circulating waterpumps, and the re-energization of only one feed pump.
NSP Response -
Design The fast bus transfer scheme from the 2R to 1R plant source is initiated by plant or substation signals which result from a loss of the 2R source.
These signals include the following.
- 1) A lockout of 2R transformer
- 2) Breaker 3N4 open
- 3) Alockout of 2RS transformer
- 4) A lockout of 345 KV Bus #1
- 5) Breaker failure of 345 KV GCB 8N4
- 6) Breaker failure of 345 KV GCB 8N11
=l g) A trip signal to 345 KV GCB 8N5 when 345 KV GCB 8N11 is l
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- 11) A lockout of 1ARS transformer l
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The above signals energize three transfer relays. The transfer relays cause the following to occur.
- 1) Trip of 4 KV breakers connecting the 2R source to plant buses 11,12,-
13 and 14.
set drive motors.
- 3) ' Trip'of 4KV breakers connecting Buses 13 and 14 to Circ Water Pump
~ motors. This trip is included in the design to protect the synchronous motors feeding the circ water pumps from a potential out of phase transfer.
- 4). Closure of the 4KV breakers connecting 1R to Buses 13 and 14 once the 4KV breakers connecting 2R transformer to Buses 13 and 14 are sensed as open by breaker auxiliary switches.
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- 15) Closure of the 4KV breaker connecting 1R transformer to Bus 11 once the following interlocks all have been met.
a) The 4KV breaker interconnecting 2R transformer and Bus 11 is sensed as open by breaker auxiliary switch position.
b) The 4KV breaker connecting Bus 11 to 11 Recirc MG drive motor breaker is sensed as open by breaker auxiliary switch position.
c) Voltage on 4KV Bus 11_has decayed to less than or equal to 25%.
- 6) Closure of the 4KV breaker connecting 1R transformer to Bus 12 once the following interlocks are all met.
a) The 4KV breaker interconnecting 2R transformer and Bus 12 is sensed as open by breaker auxiliary switch position, b) The 4KV breaker connecting Bus 12 to 12 Recirc MG drive motor breaker is sensed as open by breaker auxiliary switch position.
c) Voltage on 4KV Bus 12 has decayed to less than or equal to 25%.
d) Either one of the following two conditions apply; i) a lockout has occurred on 4KV Bus 11, or li) the breaker connecting Bus 11 to 11 RFP is open.
Operational Testing The initiating signals which energize the transfer relays are tested via maintenance procedures each operating cycle.
The trip signals, close signals, and interlocks associated with the 2R to 1R transfer are tested in conjunction with the relay calibration and test tripping maintenance procedure associated with each of the affected breaker cubicles. The Division I cubicles are tested during a Division I
' electrical outage and the Division ll cubicles are tested during a Division ll electrical outage. Each operating cycle involves a Division i or a Division ll outage. Thus the control scheme for the individual breakers within the 6
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2R to 1R transfer scheme are tested via maintenance every other
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operating cycle.
Technical Specifications MNGP Technical Specification 3.9 provides the limiting conditions for l
operation (LCOs) for the offsite sources. Any degradation of an offsite L
source transfer scheme would necessitate an operability determination in accordance with staff guidance provided by Generic Letter 91-18.. The LCO actions of Specification 3.9 would be applied if the offsite source was l
found to be inoperable because of the degradation.
b).
Forfailure of one of the two operable offsite circuits, the'Monticello Updated Safety Analysis Repor1 (USAR) indicates that design provisions are provided for automatic fast bus transfer from (1) the normal offsite power supply (transformer 2R) to the standby offsite supply (transformer 1R), (2) the normal offsite power supply (transformer 2R) to the standby offsite supply (transformer 1AR), or (3) the standby offsite supply (transformer 1R) to the standby offsite supply (transformer 1AR).
Describe design, operational testing, technical specification surveillance and limiting conditions for operation, reliability data, and/or other provisions that assure that the correct (or abnormal) operation (or failure) of this automatic power supply transfer will not cause the loss of both offsite circuits following a LOCA. With respect to reliability for having at least one offsite circuit immediately available to redundant safety systems following a LOCA, describe the effect (with and without the automatic bus transfer being operable) due to power uprate.
NSP Response l
.{
Design i
l.
. The description in Section 8.3 of the MNGP USAR of the transfer to the 1 AR source upon failure of either the 2R or 1R transformers contains -
some descriptive errors.. NSP had previously prepared a 50.59 evaluation to address and correct these errors. This evaluation was processed after the last scheduled USAR update. The original description in the FSAR
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and the description contained in a December 30,1983 letter from NSP to the NRC, "Re-analysis of Adequacy of Station Electric Distribution System l
Voltages," more correctly describe this transfer, it appears that the errors
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L resulted from editorial mistakes made during preparation of Rev. O of the USAR. The accurate description of the transfer scheme is provided below.
The transfer from the 2R or 1R source to the 1 AR source is not a fast transfer and is independent of the 2R-1R transfer. A five second delay is inherent in the design of the transfer of the safety buses to the 1 AR source. Upon a loss of 2R transformer when 1R is not available or a loss of 1R transformer when 1 R is feeding the safety buses, the de-7
p b
l energization of the source transformer will result in an initiation of the essential bus loss of voltage transfer circuit. When a loss of voltage condition has been present on the safety buses for five seconds, the loss of voltage relays initiate trips of the tie breakers connecting the safety buses to the balance of plant buses. Once the tie breakers are sensed L
open via breaker auxiliary switch contacts, the transfer of the safety buses.
to the 1AR source is permitted. This interlock assures that the 1 AR l
source is not inadvertently connected to the lost transformer and i
eliminates the risk that the 1AR source is lost due to whatever event caused the loss of the transformer initially feeding the plant.
Operational Testing The trip of the tie breakers connecting the plant's non-safety an'd safety buses is verified by surveillance testing each operating cycle.
The breaker position interlocks between the tie breakers and the 1AR.
source breakers are tested in conjunction with the relay testing procedures associated with the 1 AR source breaker cubicles.
Consequently, the Division 11 AR source breaker interlocks are checked during Division I electrical outages, and the Division il 1 AR source -
breaker interlocks are checked during Division ll electrical outages. This results in a test of the specific breaker auxiliary switch contacts on an every other outage basis, it should be noted, however, that the proper re-positioning of the breaker auxiliary switch for the tie breakers and source breakers is verified each operating cycle by surveillance testing. The 1 AR transfer scheme is verified by surveillance testing.
Technical Specification Requirements and Reliability For a discussion of the 2R to 1R transfer with regard to technical specification requirements, please see the NSP response to question 7.a above. Please see the responses to questions 7.c and 7.d below for a discussion of offsite source reliability at power rerate conditions.
c)
For weak grid conditions (i.e., when substation autotransformer No.10 is out of service), the offsite system design (afterpower uprate) willhave sufficient capacity and capability to permit operation of safety systems following a LOCA. It is the staff's understanding that the substation autotransfnrmer No.10 is considered (oris representative of) the worst case transmission system contingency. If a transmission network failure wore to occur with transformer 10 out of serv!ce, verify that the offsite system would still have sufficient capacity and capabil4y to permit operation of safety systems following LOCA. If this is not the case, verify that the offsite system would be considered inoperable because licensing / design basis requirements are not being met when autotransformer No.10 is out of service.
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NSP Response
. The offsite source configuration at MNGP is robust and not susceptible to loss of all sources from single grid failures under normal operating conditions.' Consider that the Monticello plant is fed from three separate and diverse offsite power sources, and the MNGP main generator is not unit-connected to the grid. A pre-existing outage of the No.10 autotransformer is a worst case condition for the 1R source because of less voltage regulation provided to the 115 KV system. The primary source of offsite power to the plant, the 2R transformer, is unaffected by l
this condition. A failure of No.10 autotransformer, of itself, does not prohibit operation of the 2R transformer, and the 1R source would not supply any LOCA loads if 2R was iwailable. For the weak grid condition to occur with the 1R source supplying LOCA loads, both the No.10 and 2R transformers must be inoperable. Thus the weak grid condition is an allowable yet very unlikely operating condition. NSP has determined, however, that the 1R source is fully capable of supplying the plant coincident with a LOCA and plant trip at power rerate operating conditions with the No.10 transformer and the 2R offsite source out of service.
NSP has evaluated the grid response during a trip of the Monticello plant with six limiting expected transmission system lineups which includes the weak grid condition with the No.10 autotransformer initially out of service.
See Appendix ! of NSP letter dated December 4,1997. This evaluation, CA 97-219, determined that sufficient offsite voltage would be available l.
following a plant trip at rerate conditions. That is, the resultant grid voltage would be within the operating voltage bands that determine offsite source operability as derived from the associated load studies.
The scenario described in question 7.c involves four separate failures - a -
pre-existing outage of the 2R source, a pre-existing outage of the No.10 transformer, coupled with a 115 kV network failure and a simultaneous LOCA. With the 2R source and No.10 transformer both out of service, NSP would align the 1 AR source feed from the 345 kV system which is separate and independent from the 115 kV system. Even given a fourth failure from the 115 kV network which disables the 1R source, the 1 AR source would be available with the EDGs in backup to supply all safety related loads during a LOCA at power rerate conditions. Please see the response to 7.d below for the network failure licensing basis.
l d)
Based on a review oflicensing! design basis commitments documented in the Monticello USAR, it is the staff's understanding that operability of an i
immediately available offsite circuit at Monticello (more conservatively) requires that the transmission network have sufficient capacity and capability (as demonstrated by stability analysis) so that acceptable voltage (from the offsite system via at least one of two offsite circuits) will remain available following simultaneous LOCA and any single failure on i
the offsite system or transmission network. Confirm this understanding.
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i NSP Response l
The original design basis for the reliability of the offsite power system is based on compliance with Principle D'esign Criterion 1.6 and with NSP's stated position on the proposed AEC GDC Criterion 39. See MNGP USAR Section 1.2.6 and Appendix E of the USAR page 34. These design criteria do not involve design basis accidents coincident with network failures. This is evident in NSP's response to Questions 6.1 and 6.4 of l
Amendment 16 to the MNGP FSAR. In interpreting Question 6.4 and its l
response, it is important not to equate steady state offsite source availability with reliability of non-safety related offsite sources during design basis accidents. Stability analyses are done to qualitatively assess steady state availability. The licensing basis does not require stability analyses for grid failures that occur simultaneously with a low l
probability LOCA which are not caused by the LOCA event. In the.
l extremely unlikely event that an unrelated grid failure occurred simultaneously with a LOCA that reduced the capacity of all three
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separate offsite power sources, the EDGs would be available to supply all required loads necessary to mitigate the accident.
Degraded bus voltage protection was not part of the original auxiliary l
power or offsite system design at MNGP. The degraded bus voltage design at MNGP is the result of a subsequent generic NRC action. Five l
specific design criteria for offsite power reliability have been identified for Monticello. See letter from D.B. Vassallo, NRC, to L.O. Mayer, NSP,
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"Monticello Nuclear Generating Plant," dated May 28,1982 (Ref. 7d-1).
These criteria form the design basis for the offsite power sources (1R,2R, and 1 AR) and consequently form the design basis of the load study cases. The allowable grid voltage operating bands are derived from load study cases combined with inputs on steady state grid operating voltages from the system operator. The adequacy of the offsite source reliability at MNGP and the associated load study cases were reviewed and approved i
by the staff by letter dated March 20,1985. None of these cases l
postulate a simultaneous unrelated grid failure coincident with the LOCA as a requirement to establish offsite source reliability. The degraded bus voltage analysis is used to demonstrate the reliability of the offsite sources to the MNGP auxiliary power system. In establishing this reliability, a limiting case or weak grid initial condition is assumed. The limiting case is consistent with Staff Position 1 of Ref. 7d-1 above. An unrelated grid fault coincident with a LOCA is not indicated by the original design criteria or by the staff review positions for offsite reliability described in Ref. 7d-1.
The evaluation of offsite power reliability under rerate conditions was conducted using the established assumptions for grid conditions described above in some cases, more accuracy has been introduced into the assumptions. These cases are identified in the section entitled, Electrical System Changes and Load Study Enhancements, of Appendix l to NSP's power rerate license amendment request dated Decernber 4, 10 L
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1997. By Calculation CA 97-219, NSP determined the grid voltages
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available following a plant trip coincident with a LOCA from all expected and bounding power rerate operating conditions including degraded grid conditions. In all cases, the resultant voltage was within the assumed limits. In addition, by Calculation CA 97-144, NSP analyzed non-accident grid stability at rerate conditions. The results of this calculation are summarized in Appendix 1 (pg. l-9). This analysis demonstrated the stability of the grid during rerate conditions. These studies are consistent with the design bases for offsite source reliability as described above.
l 8.
Section 3, Design Basis Accidents, of Revision 1 to the license amendment request dated July 26,1996 supporting the Monticello Nuclear Generating Plant Power Rerate Program, states that the radiological consequences of the limiting design basis accidents were re-evaluated. It further states that the evaluation was performed using inputs and evaluation techniques consistent with the current regulatory guidance and they are different from those used in the current licensing basis evaluation presented in the Monticello USAR and the safety evaluation performed by NRC (AEC).
Provide majorparameters and assumptions usedin the re-evaluation of the radiological consequences complete with dose calculations performed for the site boundaries (Exclusion Area Boundary and Low Population zone) and for the control room operators resulting from (1) the LOCA, (2) the fuel handling accident (refueling accident), and (3) the Control Rod Dmp Accident. Include a description of and the bases for the applicability of the inputs, assumptions, models, and resultant calculations related to the relative concentration values (X/Qs) used in the dose assessments. The description should also address the basis for selection of the period of meteorological data used, includingjustification oflong-term (e.g. 30 years) and site (e.g., free from local obstructions such as trees or plant structures) representativeness and measures to assure high data quality.
NSP Response NSP is including the rerate accident analysis report, Accident Radiological Analyses of Design Basis Accidents, as Attachment 4 to this letter. This report provides the information requested above with the exception of the atmospheric dispersion factors. It should be noted that credit is assumed for the isolation of
'I the mechanical vacuum pump and for holdup in the condenser in the analysis of the Control Rod Drop Accident.
NSP submitted atmospheric dispersion data as Enclosures 1 and 2 to a letter from W.J. Hill to the US NRC Document Control Desk entitled, " Supplementary information to Revision One to License Amendment Request Dated July 26,1996 Reactor Coolant Equivalent Radioiodine Concentration and Control Room Habitability, May 5,1997." The supporting dispersion data is also applicable to rerate accident analyses.
Attached is an independent audit of the 1991 meteorological data which was I
used for rerate radiological analyses. See Attachment 5. The validity and 11
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variability of the relative concentration (X/Q) data was verified and found acceptable.
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9.
In response to Question 23 in the submittal dated September 5,1997, it is stated that the response to this question as it applies to motor-operated valve (MOV) performance following the power uprate is in progress and will be submitted at a later date. As of the date of this RAI, we have not yet received your response.
l Provide a list of safety-related valves affected by the power uprate, their functions
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l and operating conditions (including, pressure, temperature, differentialpressure, l
flow rate, ambient pressure and stroke times) at the current (100 percent power
' design basis) and the power uprate conditions. Identify mechanical components l,
for which operability at the uprated powerlevel could not be confirmed. Also, provide a discussion of how the Monticello MOV and air-operated valve programs have been updated to reflect the extended power uprate condition.
NSP Response l
Summary The MOVs with safety related functions were evaluated to determine if they would be capable of performing their intended functions at the rerate (1,775 MWt) power levels. Operational scenarios for each MOV were reviewed and bounding conditions (Operating Situations, Maximum Expected Differential Pressure, Actuator Ambient Temperature) were established.
Based on these operational scenarios, MOVs were evaluated to determine if they were capable of performing their design basis functions at rerate conditions. This evaluation was conducted in accordance with the recommendations of Generic Letter 89-10. By letter dated May 11,1995 entitled, "Close-Out inspection of Generic Letter 89-10 (NRC Inspection Report No. 50-263/95003(DRS))," staff found Monticello's MOV program acceptable.
4 NSP determined the environmental conditions for the GL 89-10 MOVs under power rerate conditions. This evaluation included valves contained in the following systems.
i Reactor Building Closed Cooling Water Core Spray Residual Heat Removal Reactor Recirculation High Pressure Core injection Reactor Core Isolation Cooling Main Steam Reactor Water Cleanup Combustible Gas Control t
All MOVs were found to be capable of performing their design basis functions at I
rerate power levels without modification with the singular exception of MO-2030, RHR Shutdown Cooling Outboard Isolation which is discussed below. This 1
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conclusion is based on the results of NSP Calculation 97-012, Effects of Rerate and Revised Analyses on MOV Operability. The calculation is available onsite for i
staff review.
l' Operational Scenarios o
As discussed below, bounding MOV service conditions for plant operations at current and rerate power levels were utilized for the MOV operational scenarios.
Operational Situations.
Basic system configurations, operation, and event response is not changed by power rerate conditions. The design basis functions for these valves were not changed.
Operating Pressure Safety / Relief Valve Setpoint The as-found set pressure testing criterion for the safety / relief valves will be changed from 2% to 3% at power rerate conditions.
Thus, the reactor pressure to be used for applicable containment isolation MOVs corresponds to the current setpoint of 1,109 psig plus 3% (1142 psig). In anticipation of increasing this tolerance, MOV operational scenarios were modified to incorporate the increased reactor pressure of 1142 psig, and affected MOVs were found capable of performing their design basis functions under this increased operating pressure.
Shutdown Cooling isolation a
The shutdown cooling supply isolation trip settireg is proposed to be changed to 75 psig at the reactor vessel steam dome. This setpoint change was incorporated into the operational scenario, and it was determined that the torque switch setting for MO-2030 is set such that the valve may not close under this service condition. The torque switch setting will be reset prior to changing the shutdown cooling isolation interlock setting to 75 psig at the.
i reactor steam dome.
Environmental Temperature New High Energy Line Break (HELB) and heatup analyses were performed in support of the rerate project. The appropriate motor ambient temperature was selected and incorporated into the affected MOV operational scenarios. The effects of these temperatures on MOV I
performance capability were considered, and affected MOVs were found capable of performing their design basis functions under these new conditions.
1 13
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I The temperature results from a preliminary reactor building heatup calculation were used in this evaluation. NSP believes that these results represent bounding conditions that are appropriate for MOV analyses. As stated in Rev.1 to the rerate License Amendment Request, a Reactor Building post LOCA heatup analyses from a detailed reactor building model is in progress. See Appendix H to NSP's letter of December 5, 1997. This calculation will be completed under a 10CFR 50 Appendix B QA Program and will provide results which will be incorporated into the I
affected MOV functional requirements. Any changes to the conclusions l
above will be' reported to the staff prior to implementation of power rerate.
i Stroke Time Changes l
Because the basic system configurations, operation and event response is not changed by the power rerate project, existing stroke time requirements are not changed under rerate operating conditions.
System Flow Changes j
Because the basic system configurations, operation and event response is not changed by the power rerate project, existing flow requirements 1
are not changed under rerate operating conditions.
Monticello MOV Program The Monticello MOV program elements including program scope, actuator capability evaluation methodology, and periodic verification methodology were not changed as a result of the proposed power rerate. A' discussed above, a limited number of s
parameters affecting the evaluation of actuator capability changed as a result of the power rerate analyses. Monticello uses an MOV program data base to analyze MOV performance capability and documents this analysis using the Monticello calculation control process. The MOV program data base was updated to reflect the revised actuator ambient temperatures, maximum expected differential pressures and maximum expected line pressures, as applicable to the MOV, and as determined by the power rerate MOV analysis. The MOV performance calculations were performed using the revised set of input parameters, and these calculations confirmed that the affected MOVs
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are capable of performing the required design basis functions at power rerate operating conditions.
Monticello AOV Program j
Because the basic system configurations, operation and event response are not changed by the power rerate project, existing AOV system requirements will not be affected. A qualitative review of systems that contain safety related AOVs has been performed, and there are no changes to system performance parameters that will affect safety related AOV performance.
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NSP is closely following the industry developments regarding the application of AOVs in the nuclear power industry and s participating in the Joint Owner's Group on AOV issues. A generic AOV Program has not yet been established at Monticello, and consequently NSP did not apply a generic methodology to evaluate the effects of power rerste on AOV performance.
I 10.
In your response to Question 25 in the ' ubmittal dated September 5,1997, s
1 regarding the maximum calculated stresses for the critical BOP [ balance-of-plant]
piping systems, NSP stated that the maximum piping stress increases shown in p
Table 3-5 of the power uprate license amendment request. Table 3-5,
- Piping Stress Comparisons,"provides the maximum percent increases in piping stresses forlimiting BOP systems. Examination of the data in Table 3-5 indicates
. that the percent stress increases are substantial at some locations. We request that NSP provide a comparison of maximum stresses against the code allowable stress lim'ts to demonstrate that the piping systems and their supports are within the allowable limits at the uprated powerlevel.
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NSP Response Maximum piping stress ratios for the limiting load combinations are shown in the table below. These limiting load combinations are selected to represent pipe stresses that are affected by rerate as identified in Table 3-5 of the revised rerste license amendment request. The worst case stress locations for pipe supports and anchors are found to be those associated with torus attached piping. A few torus attached piping supports are at 99% of the code allowables at rerate conditions. Other pipe support stresses are well within code allowables.
The acceptance criteria for piping and pipe supports is the applicable code stress limits as defined in USAR Section 12. See NSP responses to Questions 24,25, 26 and 27 of the September 5,1997 submittal. All piping and pipe support stresses are within the applicable USAR code limits. The power rerate piping and pipe support calculations are available for review onsite.
System Load Case Max. Stress / Code Allowable Maximum Stress Stress Ratio Main Steam P + DW + TSVC (1) 17539 psi /18000 psi 97%
Feedwater P + DW + TH 20565 psi /37500 psi 55 %
HPCI P + DW + TH 25313 psi /37500 psi 68 %
P + DW + TSVC See main steam See main steam RCIC P + DW + TH 18452 psi /37500 psi 49%
P + DW + SSE +TSVC 22707 pair 26332 pal 86 %
Torus Attached TH (2) 22220 pair 22500 psi 99%
Piping FW Heat P + DW + TH 31360 psi /37500 psi 84 %
Exchanger Turbine Steam P + DW + TH 32395 psi /37500 poi 86 %
Notes:
P = Pressure DW = Deadweight TH = Thermal SSE = Safe Shutdown Earthquake TSVC = Turbine Stop Valve Closure (1) Transient Load. Normal stress ratio is 80%.
(2) Accident (DBA LOCA) Load 15
1 11.
Please describe how NSP has verified that the safety limits and operating limits.
2 provided by the fuel vendor for the Monticello-specific core are calculated in accordance with NRC-approved codesimethodologies with applicable limitations (if any) containedin the statisafety evaluation. List allrestrictions and conditions specified in the referenced topical reports and their associated. safety evaluations that are appropriate for Monticello's specific core.
NSP Response j
NSP's current practice is 'to review the vendor's calculations of the provided safety limits by surveillance audits and third party technical reviews to ensure that -
L the calculations are based upon an approved methodology and are consistent with all constraints.
The safety evaluations performed in support of power rerate that used GE-NE's (GE Nuclear Energy) standard analysis codes and application methodologies are referenced in the GESTAR ll documentation. The GE analysis codes and methodologies have been developed and approved for application to a wide range of GE BWR plant types and operating conditions. The power rerate operating and accident conditions analyzed for Monticello are within the range of analysis experience for GE BWRs and are within the allowed range of code and methodology application. The restrictions and conditions applicable to GE-NE's core and fuel design are documented in GESTAR 11, NEDE-24011-P-A-11, Revision 11, General Electric Standard Application for Reactor Fuel, November 1995 and GESTAR 11, NEDE-24011-P-A-11-US, Revision 11, General Electric Standard Application for Reactor Fuel (Supplement for United States), November 1995.
The restrictions and conditions applicable to the SAFER /GESTR-LOCA methodology are described in Section 3.2 of Exhibit G (NEDC-32514P Rev.1).
Briefly summarized, the Licensing Basis PCT must be below 2200 F, the Upper Bound PCT must be less than the Licensing Basis PCT, and the Upper Bound -
i PCT must be less than 1600 F. Compliance with these restrictions and I
conditions for the Monticello SAFER analysis is shown in Section 5.2 and summarized in Table 6-1 of Exhibit G.
The NSP Nuclear Analysis & Design (NSPNAD) group performs reactor reload and transient analyses to develop core operating limits in compliance with the topical reports and safety evaluations listed below. Please note that the current l
revisions of the topical reports differ from the revision reviewed by the NRC staff j
only by the actual inclusion of the staffs safety evaluation in the topical report.
l
- 1. NSPNAD-8609-A, Revision 3, " Qualification of Reactor Physics Methods for Application to Monticello, Revision 3," October 1995 Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to 4
Revision 2 of Topical Report NSPNAD-8609 " Qualification of Reactor Physics Methods for Application to Monticello" for Northem States Power Company l.
16 s
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t Monticello Nuclear Generating Plant Docket No. 50-263, dated September 11,1995.
- 2. NSPNAD-8608-A, Revision 4,"Monticello Reload Safety Evaluation Methods," October 1995.
Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to L -
Revision 3 of Topical Report NSPNAD-8608 Reload Safety Evaluation Methods for Application to Monticello for Northern States Power Company Monticello Nuclear Generating Plant Docket No. 50-263, dated September i
29,1995.
Topical report NSPNAD-8609-A specifies the use of the CASMO-3/ SIMULATE-3 methodology package for steady-state BWR core physics reload design and.
3 licensing applications, including fuel bundle and loading pattern analysis for the generation of core physics control rod worth and startup predictions, reactivity coefficients for transient and safety analysis input, and for the potential support of i
the process computer core monitoring system. The safety evaluation limits the i
application of the methodology to the range of fuel configuration and core design parameters verified and referenced by the topical report; introduction of significantly different fuel designs may require further validation by the licensee.
Topical report NSPNAD-8608-A specifies the use of the SIMULATE-3 based.
DYNODE-B transient application to the Monticello BWR. The safety evaluation provided the following restrictions to the application of the Monticello DYNODE-B methodology.
A. Transient evaluations are designed to use time varying axial power shapes for all Monticello 1-D transient simulations.
B. Transients events are analyzed using the appropriate kinetics (i.e.1-D or 0-D).
C. The 1-D collapsing adjustment factors should be shown to be conservative for each reload application or included in the ACPR determination.
D. The operating initial critical power ratio ICPR is calculated as stated in the I
safety evaluation with the imposed 0.05 ACPR/ICPR uncertainty.
l E. If uncertainties or sensitivities used in the DYNODE-B uncertainty analysis increase, the effect of these changes should be included in the ACPR uncertainty allowance.
F. The application of DYNODE-B is restricted to core inlet flow oscillations with
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frequencies below 5 Hz.
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G. DYNODE-B can be used to calculate transient ACPR and vessel pressure j
j but does not include the calculation of the core decay ratio.
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,,l 12.
Please provide the upper bound PCT [ peak cladding temperature] for the Monticello plant at the limiting large and small break sizes (at the DBA-LOCA and the 0.06 sq It. size).
NSP Response The Upper Bound PCT for the DBA-LOCA is less than 1550*F for GE11 fuel and less than 1460*F for GE10 fuel. The Upper Bound PCT for the 0.06 sq. ft. break is less than 1480*F for GE11 fuel and less than 1420*F for GE10 fuel.
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13.
In Section 4.1, Exhibit E of the December 4,1997, submittal, it references NRC's i
safety evaluation dated July 25,1997, which reviewed and approved the NSP's license amendment request dated June 19,1997. This submittalincluded confirmatory calculations with the SHEX code and the HXSIZ code conducted at 1880 MM to bound the calculated core shutdown power that would result from the use of ANS 5.1-1979 decay heat model with a 2-sigma uncertainty adder at.
the currently licensed powerlevel at 1670 MWt. Please provide similar analyses at 1775 MM in support of the proposed power terate.
NSP Response Sufficient conservatism exists in the 1880 MWt decay heat profile to conclude that it bounds the rerate decay heat profile at 1775 MWt by 2o. The GE generic 1880 MWt decay heat profile used for the power rerate analyses, without adjustment, bounds the MNGP-specific decay heat profile at 1775_MWt with a 2a adder for the first 30 days following shutdown. This is significant because the bounding values of design parameters used for equipment performance typically l
occur within this 30 day period. _ For instance, the peak suppression pool temperature occurs within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
For periods greater than 30 days, the 2o adder can be demonstrated in the decay l
heat profile by making a reasonable correction to a conservative analytical assumption. GE's generic application of the ANS 5.1-1979 decay heat standard accounts for the effect of neutron capture in fission products by directly applying the specified G(t) factor. The G(t) factor is based on neutron flux conditions and l
average cross section data appropriate for pressurized water reactors which introduces a significant conservatism for BWR applications. The GE generic 1880 MWt decay heat profile is about 1% lower than the MNGP-specific 1775 MWt profile with a 2o adder for times greater than 30 days. Use of a BWR-specific G(t) factor would reduce the MNGP decay heat profile by more than 3%
l over this time penod. Therefore the BWR specific 1880 MWt decay heat profile I
also bounds the MNGP 1775 MWt decay heat profile with a 2a adder for periods greater than 30 days.
l 4'
As stated in the staffs safety evaluation dated July 25,1997, the 2a adder is l
intended to provide a 95% confidence level on the decay heat used in the L
analysis. This 95% confidence criterion can be met over the entire profile in the I
form of a statistical percentile irrespective of the G(t) factor argument. For 18 j
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licensing analyses, the upper bound of the normal distribution is an appropriate parameter since lower values of decay heat result in less limiting analytical results. Thus it is reasonable to construct the confidence interval from the one-sided upper tail or percentile of the normal distribution. For this distribution, an uncertainty of 1.645a corresponds to a 95% confidence interval. In terms of the i
decay heat profile, it can be concluded that the actual decay heat will be less than l
the bounding calculated value with 95% confidence if the sampling difference is 1.645o.
The GE generic 1880 MWt decay heat profile used for the MNGP power rerate analyses bounds the MNGP-specific decay heat profile at 1775 MWt by an amount greater than 1.6450 on the upper bound. Given the above, it is reasonable to conclude that the 1880 MWt decay heat profile used in the power rerste containmr.nt analyses bounds the MNGP-specific 1775 MWt decay heat profile at the 9f.% confidence level.
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Radiological Analyses of Degg n Bases Accidents g
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