ML20203H000

Forwards Request for Addl Info Re 960726 Amend Request to Increase Monticello Nuclear Generating Plant Operating License Max Power Level to 1775 Megawatts Thermal & Revise Supporting Plant TSs

ML20203H000
Person / Time
Site:	Monticello
Issue date:	02/11/1998
From:	Kim T NRC (Affiliation Not Assigned)
To:	Richard Anderson NORTHERN STATES POWER CO.
References
TAC-M96238, NUDOCS 9803030131
Download: ML20203H000 (6)

Text

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Mr. Roger O. Anderson, Director February 11, 1998 Licensing and M:nagemInt is:ues Northern States Power Company 414 Nicci,et Mall Minr,capolis, Minneapolis 55401

SUBJECT:

MONTICELLO NUCLEAR GENERA flNG PLANT - REQ ~UEST FOR ADDITIONAL INFORMATION ON l.lCENSE AMENDMENT REQUEST ENTITLED " SUPPORTING THE MONTICELLO NUCLEAR GENERATING PLANT (MNGP) POWER RERATE PROGRAM" (TAC NO. M96238)

Dear Mr. Anderson:

- By letter dated July 26,1996, Northern States Power Company (HSP) submitted a license amendment request to increase the MNGP operating license maximum power level to 1773 megawatts thermal. and revise supporting MNGP Technical Specifications. This change  ;

reflects an increase of 6.3 percent above the currently licensed power level of 1670 megawatts L thermal.

On April 14,1997, the staff issued its request for additional information (RAl) based on a preliminary review of the July 26,1996, submittal. NSP responded to the staffs RAI in a letter dated September 5,1997. Subsequently, by a letter dated December 4,1997, NSP submitted Revision 1 to the original submittal dated July 26,1996.

Based on a review of the submittals dated September 5 and December 4,1997, the staff has determined that additionalinformation is necessary to complete i*s review. The enclosed RAI provides details of the required material. Please advise NRC of NSP's schedule for responding to the enclosed RAl.

S;ncerely, ORIGINAL SIGNED BY Tae Kim, Senior Project Manager {

Project Directorate ill-1 I Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation Docket No: 50-263

Enclosure:

As stated cc w/ encl: - See next page h3 ,

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• ,. - Mr. Roger O. Anderson, Director. Monticello Nuclear Generating Fhnt Northem States Power Company cc:

J. E. Silberg, Esquire. Kris Sanda, Commissioner Shaw, Pittman, Potts and Trowbridge Department of Public Service

- 2300 N Strett, N. W. 121 Seventh Place East Washington DC 20037 . Suite 200 St. Paul, f.iinnesota 55101 2145

~ U.S. Nuclear Regulatory Commission Resident inspector's Office Adonis A. Nebiett

- 2807 W. County Road 75 Assistant Attomey General Monticello, Minnesota 55362 Office of the Attomey General 445 Minnesota Street Plant Manag t Sute 900 Monticello Nuclear Generating Plant St. Paul, Minnesota 55101-2127 ATTN: Site Licensing Northern States Power Company 2807 West County Road 75 Monticello, Minnesota 55362w' 37

]

Robert Nelson, Pruident -

Minnesota Environmental Control l

Citizens Association (MECCA) 1051 South McKnight Road St. Paul, Minnesota 55119 Commissioner Minnesota Pollution Control Agency 520 Lafayette Road St. Paul, Minnesota 55119 -

Regional Administrator, Region 111 U.S. Nuclear Regulatory Commission A-801 Warrenville Road

Lisle, I'linois 60532-4351 .

Commissioner of Health Minnesota Department of Health 717 Delaware Street, S. E.-

Minneapolis, Mir'.nesota 55440 Darla Groshens, Auditor / Treasurer Wright County Govemment Center 10 NW Second Street Buffalo, Minnesota 55313 January 1995

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REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED POWER UPRATE FOR THE M')NTICEll O NUCL EAR GENERATINO PLANT Docket Number 50-283

1. 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 mlnutes 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 exception [ emphasis added] that the actions would indeed be accomplished." The licensw noted that the two subject operator actions are examples of operator actionu most sensitive to power rarate, i Please provide the bases for assurance (i.e., simulator observations and licensee assessments) that operators can perform these actions in the required response times.

l 2. NSP's response to question 36 in the submittal dated September 5,1997, provides a statement that was to be added to the revised license amendment. The statement-should clearly and specifically indicate that:

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• For p:)wer rerate, GE setpoint methodology provided in NEDC-31336, General Elect,-ic Setpoint Methodology, is used in establishing setpoints.

3. NSP's response to question 38 provided that the setpoint for Condenser Low Vacuum has N 'n revised from 23.25" Hg to 22.25" Hg. However, Exhibit B, TS Tab'e 3.1.1,

. pr , and Exhibit A, pages A-8 and A-31, indicate that the Condenser Lew Vacuum p setpoint has been revised from 23" Hg to 22" Hg,l Clarify this discrepancy.

4. , Exhibit A, page A-9, item 2b, states '. hat TS Table 3.1.1, page 30, item d, will be revised -

to state, "when the reactor thermal power is <45% (798.75MWt)." However, Exhibit B and Exhibit C indicate this to be, "when the reactor thermal power is 545%

. (798.75MWt). Clarify this discrepancy.

5. ' Exhibit 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 not -

identified in Table 5.1:

a) Turbine condenser low vacuum (TS Table 3.1.1).

b) Low pressure core cooling pumps discharge pressure interlock (TS Table 3.2.2, item c.3).

- c) Reactor pressure interlock (TS Table 3.2.1, item 6.a).

Explain why these changes were not included in Table 5.1 of Exhibit E.

ENCLOSURE

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- 6. For each component / equipment type (or one repret Vative/ bounding example of a coinponentlequ pment type) where expected environtr.antal conditions at ine uprate power level 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 current power levels (if applicable /available), and tne expected 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 - i of-coolant accioent) operation, b) Evaluation demonstrating qualification for each segment of the uprata power level temperature response that is not enveloped by the environmental conditions (i.e., temperawre) tested to.

c) Where (or if) margins derived through the use of the Arrhenius methodology are utilized M part of the basis for concluding continued qualification, provide the Arrhenius calculation at the current (if applicable /available) ed uprate power I

levelw. : Define the margins available for the currant and upra.o power levels and describe and justW the reduced margin for the uprate power level.-

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d) Provide MNGP Calculation CA 97-176 which shows that the equivalent -

temperature exposure time for the iM (environmental qualification) temperature .

evaluation profile exceeds the equivalent temperature exposure time for the DBA [ design basis eccident) temperature profile.

7. The Monticello normal design configuration includes provisions for the automatic fast bus transfer of the offsite power source through the 2R to the offsite power source through the 1R transformer. As a result of power uprate and other design changes, loading on the 1R transformer has increased. To accommodate this increased loading nnd to assure acceptable voltages for safety system loads, the licensee has derated the 1R transformer and implemented design provis,ons (when automatic transfer occurs) to trip both recirc MGs (motor generators], to trip both circulating water pumps, and to re-energize only one feed pump.

a) ' Describe design, operational, testing, technical specification reqairements, and/or other provisions that assure a trip of both recirc MGs, a trip of both circulating water pumps, and the re-energization of only one feed pump.

b) For failure of one of the two onerable offsite circuits, the Monticello Updated Safety Analysis Report (USAR) indicates that design provisions are provided for automatic fast bus transfer from -(1) the normal o*fsite 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 1 AR). Describe design, operational, testing, technical specification surveillance and limiting conditions for operation, reliability data,

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. 3 and/or other provisions that assure that the correct (or abnormal) operation (or failura) 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.

c) For " weak grid

• conditions (i.e., when substation autotransformer No.10 is out of service), tne of' site system design (after power uprate) will have sufficient capacity and capability to permit operation of safety systems following a LOCA.

It is the staffs understanding that the substation autotransformer No.10 is considered (or is representative of) the worst-case transmission system y

contingency. If a transmission network failure were to occur with transformer 10 i' out of service, verify that the offsite system would still have sufficient capacity and capwollity to permit operation of safety systems following a 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.

d) Based on a review of 1: censing / design basis commitments documented in the Monticello JSAR, it is the staffs unde standing that operability of= an immediately available offsite circuit at Monticello (more conservatively) requires that the transmission network have sufficient capacity and captbility (as demonstrated oy 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 the offsite system or transmission -

network. Confirm this understanding.

8. Section 3, Design Basis Accidents, of Revision 1 to the license amendment request dated July 2S,1996, supporting the Monticello Nuclear Generating Plant Fower Rerate Program, states that the radiological consequences of the limiting design-basis accidonts 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 evniustion performed by NRC (AEC).

Provide major parameters and assumptions used in 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 drop accident. Include a description of and the bases for the applicabibf 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 J

' data used, including justification of long-term (e.g., 30 years) and site area (e.g., free from local obstructions such as trees or plant structures) representativeness and measures to assuro high data quality.

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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 por, er uprate is in progress and will be submitted at a later date. As of the date of this RAI, we have not received your response. Provide a list of safety-related valves affected by the power uprate, their functions and operating conditions (including pressure, temperature, differential pressure, flow rate, ambient pressure, and stroke times) at the current (100 percent power design basis) and the power uprate conditions, identify mechanical components for which operability at the uprated power level could not be confinned. Also, provide a discussion of how the Monticello MOV and air-operated valve programs have been updated to reflect the extended power

- uprate condition.

10. In response to Quewon 25 in the submittal dated September 5,1997, regarding the maximum calculatcd stresses for the critical BOP [ balance-of-plant) piping systems,-

NSP stated that the maximum piping stress increases are shown in Table 3-5 of the power uprate license amendment request. Table 3-5, ' Piping Stress Comparisons,"

provides the maximum percent increases in piping stresses for limiting BOP systems.

Exam mtion of the data in Table 3-5 indicates that the percent stre ss increases are substantial at some locations. We request that NSP provide a comparison of maximum stresses against the code allowable stress limits to demonstrate that the piping systems and their supports are within the allowable limits at the uprated power level.

11. Please describe how NSP has verified that the safety limits and operating limits provided by the fuel vendor for the Monticello-specific core are calculated in accordance with NRC-approved codes / methodologies with applicable limitations (if any) contained in the staff safety evaluation. List all restrictions and conditions  ;

specified in the referenced topical reports and their associated safety evaluations that are appropriate for Monticello's specific core.

- 12. Provide the uppar 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 ft size),

p

13. In Section 4.1, Exhibit E of the December 4 -1997, submittal, it references NRC's safety evaluation dated July 25,1997, which renwed and approved the NSP's license amendment request dated June 19,1997. This submittalincluded confumatory calcolat'ons with the SHEX code and the HXSIZ code conducted at 1880 MWt to bound the calculated core shutdown power that would result from the use of ANS 5.1-1979 dacay heat model with a 2-sigma uncertainty adder at the currently licensed power -

- level of 1670 MWt. Pleate provide similar analyses at 1775 MWt in support of the proposed power rerate.

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