L-MT-09-042, Extended Power Uprate: Response to NRC Reactor Inspection Branch Request for Additional Information (RAI) Dated March 20, 2009

Extended Power Uprate: Response to NRC Reactor Inspection Branch Request for Additional Information (RAI) Dated March 20, 2009

ML091671787
Person / Time
Site:	Monticello
Issue date:	06/16/2009
From:	O'Connor T Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-09-042, TAC MD9990
Download: ML091671787 (14)

Text

June 16,2009 L-MT-09-042 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License License No. DPR-22 Monticello Extended Power Uprate: Response to NRC Reactor Inspection Branch Request for Additional Information (RAI) dated March 20, 2009 (TAC No. MD9990)

References:

I. NSPM letter to NRC, License Amendment Request: Extended Power Uprate (L-MT-08-052) dated November 5, 2008, (Accession No. ML083230111)

2. Email P. Tam (NRC) to G. Salamon, K. Pointer (NSPM) dated March 20, 2009, Monticello - Draft RAI from Reactor Inspection Branch re: proposed EPU amendment (TAC No. MD9990)

Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota corporation (NSPM), requested in Reference 1 an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications (TS) to increase the maximum authorized power level from 1775 megawatts thermal (MWt) to 2004 MWt.

On March 20, the U.S. Nuclear Regulatory Commission (NRC) Reactor lnspection Review Branch provided four requests for additional information (RAls) described in Reference 2. Enclosure Iprovides the NSPM response.

In accordance with 10 CFR 50.91, a copy of this letter is being provided to the designated Minnesota Official.

Monticello Nuclear Generating Plant 2807 West County Road 75 e Monticello MN 55362

U.S. Nuclear Regulatory Commission L-MT-09-042 Page 2 of 2 Summary of Commitments There are no new commitments contained in this letter and no existing commitments are revised by this letter.

rjury that the foregoing is true and correct.

Ilo Nuclear Generating Plant any - Minnesota Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce

ENCLOSURE NSPM RESPONSE TO REACTOR INSPECTION BRANCH RAls DATED MARCH 20,2009

L-MT-09-042 Page Iof 11 NRC RAI No. 1 The Safety Analysis Report [SARI for the Monticello Constant Power Uprate, dated October 2008, page 2-343, indicates that an analysis predicts only a 3 percent increase in dose rates at the high pressure (HP) turbine due to increased N-16 in the steam entering the turbine. This is somewhat lower than analysis provided for similar uprates at other BWRs in the U.S.

NRC RAI No. 1(a)

Provide a detailed description of this analysis and input parameters. Include the main steam transient time, from the reactor head to the HP-LP turbine crossover line under full power CLTP [current licensed thermal power] conditions; and the corresponding transient time under full power EPU conditions.

NSPM RESPONSE The prediction of a 3 percent increase in dose rates was unintentionally misleading as used in this paragraph. Three percent refers to the increase in dose rates due to reduced transit and decay times only, and does not include the effect of increased hydrogen injection and the resulting increase in N-16 production in proportion to the increase in Feedwater flow rate of 14.8 percent. If this was included, as was done for the shine evaluation on page 2-344 of the SAR, the result would be 18.2 percent (1.I48 x 1.03).

A Microsoft Excel spreadsheet was used to estimate the changes for various plant areas based on the effect of steam transit time and N-16. The estimates were made by scaling original license thermal power (OLTP) transit times to various components from existing Calculation CA 67-086, first to CLTP, and then to EPU conditions, and then computing the difference in N-I6 decay time to compute a change in radiation level. The results are shown below, and the percent change EPU column is relative to CLTP.

L-MT-09-042 Page 2 of 11 0 6.15Et01 6.78Et06 0.00 7.26Et06 6.1 1Et01 0.00 8.32Et06 6.00Et01 -1.70%

2.05321 2.05321 5.04Et01 6.44Et06 1.95 6.77Et06 5.05Et01 1.59 8.32Et06 5.14Et01 1.84%

0.49 2.05321 5.04Et01 6.44Et06 1.95 6.77Et06 5.05Et01 1.59 8.32E1.06 5.14E+01 1.84%

1.I2 2,68321 4.74Et01 6.44Et06 2.55 6.77E+06 4.76Et01 2,08 8.32Et06 4.91Et01 2.95%

1.14 2,70321 4.73Et.01 6.44E+06 2.57 6.77Et06 4.76Et01 2-09 8.32Et06 4.90E+01 2.99%

1.53 3,09321 4.55Et01 6.04Et06 2.96 6.32Et06 4.58Et01 2.41 7.74Et.06 4.75Et0-01 3.62%

3.44 5.00321 3,78E+01 6.04Et06 4.78 6.32Et06 3.84Et01 3.91 7.74Et06 4.11E~01 7.04%

4.04 5,60321 3.57Et01 5.40Et06 5.34 5.67E+06 3.63Et01 4.33 7.00Et06 3.94Et01 8.45%

4.07 5,63321 3,56E+01 5.40Et06 5.37 5,67E+06 3.62Et01 4,35 7.00Et06 3.93EtOl 8.50%

4.07 5.63321 3.56EtOI 6.43Et06 5.36 6.76Et06 3.63Et01 4.35 8.34Et06 3,93E+01 8.48%

4.11 5.67321 3.54Et.01 6.43Et06 5.40 6.76Et06 3.61Et01 4.38 8.34Et06 3,92E+01 8.55%

2.18 3.74321 4.28Et01 3.63Et05 3.36 4.05Et05 4.41Et01 2.55 5.33Et05 4.68Et01 6.30%

6.81 8.37321 2.73Et.01 3.63Et.05 7.51 4.05Et05 2.94EtOl 5.71 5.33Et05 3.45Et01 17.10%

4-88 6.44321 3.29Et01 7.89Et05 6.17 8.25Et05 3.35Et01 5.19 9.79E+05 3.62Et01 8.07%

8.48 10.04321 2.32Et01 7.89Et05 9.61 8,25Et05 2.40Et01 8.09 9.79Et05 2.73Et01 13.94%

0 1.56321 5.29Et01 7.00Et03 1.33 8.20Et03 5.36Et01 0.90 1.22EtO4 5.50Et01 2.58%

9.1 10.66321 2,18E+01 7.00Et.03 9.10 8.20Et03 2.52Et01 6.12 1.22Et04 3.31Et01 31.40%

4.11 5,67321 2.29Et05 8.00Et02 5.67 8.00Et02 2.45Et05 5.67 8.00Et02 3.28E+05 33.76% 0.8 factor described in USAR 12.3.2.2.2 121 122,56321 2.66Et00 8.00Et02 122.56 8.00Et02 2.45Et05 122.56 8.00Et02 3.28Et05 33.76%

4.1 1 5.67321 7,12E+00 6.43Et06 5.40 6.76E+06 7.23Et00 4.38 8.34Et06 7.85E+00 8.55% 0.2 factor described in USAR 12.3.2.2.2

L-MT-09-042 Enclosure I Page 3 of 11 NRC RAl No, l(b)

It is the staffs understanding that the steam crossover line from the HP to LP turbines is the major source of N-I6 gamma radiation shine from BWR turbine buildings. Verify that this is the case for Monticello or provide the transient time information in 1.a. above from the reactor head to the turbine building component determined to be the major gamma source.

NSPM RESPONSE In general, the dose changes due to N-16 in the equipment above grade will be the most significant factor in skyshine although radiation scatter from other sources may be present. The equipment above grade at MNGP includes steam piping, turbines, feedwater heaters, the upper portions of moisture separators, and the transition between the turbines and condenser. The largest increase due to reduced transit and decay time (17.1 percent) and increased N-16 production (14.8 percent) is at the outlet of the 15 Feedwater Heaters and is 34.4 percent (1.171 x 1.148).

NRC RAI No. 2 The Safety Analysis Report for the Monticello Constant Power Uprate, dated October 2008, page 2-344, indicates that "EPU may result in a maximum skyshine source dose rate increase of up to 34.4 percent" and that this results in a maximum increase in offsite dose due to sky shine at EPU conditions of less than 6 mremlyr.

NRC RAI No. 2(a)

Resolve the apparent discrepancy between the 3 percent increase stated on page 2-343 and the 34.4 percent increase stated on page 2-344.

NSPM RESPONSE See discussion under the response to RAI No. 1(a).

L-MT-09-042 Enclosure I Page 4 of 11 NRC RAI No. 2(b)

Describe how Monticello currently demonstrates that the annual dose to the maximum exposed member of the public meets the 25 mremlyr requirement of 40 CFR 190.

NSPM RESPONSE The 2006 Annual Radiological Operating Report for MNGP reported the results of radiation monitoring for the plant. The report stated:

Ambient radiation was measured in the general area of the site boundary, at an outer ring 4 - 5 mi [miles] distant from the plant, at special interest areas and at four control locations. The means were similar for both inner and outer rings (16.5 and 15.6 mReml91 days, respectively). The mean for the control locations was 15.7 mRem191 days. Dose rates measured at the inner and outer ring locations were similar to those observed from 1991 through 2005.

No plant effect on ambient gamma radiation is indicated.

The data is provided in Table Ion the following page. The conclusion in the report is that there is no plant effect on ambient gamma radiation. This would support an estimate that skyshine changes due to EPU will not have any impact on measured dose rates offsite.

The data shows a maximum difference between the inner and outer ring mean of all locations of 1.Imrem for a quarter. If this is taken as a measure of skyshine, it represents a maximum of 4.4 mrem per year at current conditions. Scaling this result by 34.4 percent is less than 6 mremlyr. This is considered a conservative upper bound for offsite dose to skyshine at EPU conditions.

From Table 2 it can be seen that the average exposure due to gaseous emissions and liquid effluents to an individual are less than a total of Imrem per year. Adding this to the skyshine estimate of 6 mremlyr is a total of 7 mrem. As a result, it is concluded that the maximum potential dose to any member of the public will remain well within the 40 CFR 190 limit of 25 mremlyr.

L-MT-09-042 Enclosure I Page 5 of 11 Table 1: Ambient Gamma Radiation as Measured by Thermoluminescent Dosimetry, Average Quarterly Dose Rates, lnner vs. Outer Ring Locations Inner Ring ) Outer Ring Year Dose rate (mRemIqtr) 1991 15.2 15.8 1992 15.I 15.1 1993 15.6 15.9 1994 14.6 14 1995 14.4 13.6 1996 14 13.5 1997 13.3 12.8 1998 15 14.4 1999 15.1 14.3 2000 15.1 14.5 2001 14.3 13.7 2002 15.9 14.8 2003 15.6 15 2004 16 15.4 2005 15.6 15.2 2006 16.5 15.6 Average 15.5125 14.8125

L-MT-09-042 Enclosure I Page 6 of 11 Table 1A below compares the mean for all locations in both the inner and outer rings and the mean of the peak location in each ring for the last 11 years. The maximum difference between the inner and outer ring peak locations is I. 7 .

mremlqtr. If this is taken as skyshine, as done above, it represents a maximum of 6.8 mremlyr at current conditions. Scaling this by 34.4 percent results in a maximum projected upper bound for offsite dose due to skyshine of 9.1 mremlyr.

Adding this to the average exposure from Table 2 of 1 mremlyr results in a total of approximately 10 mremlyr maximum potential dose to any member of the public. This is well within the 40 CFR 190 limit of 25 mremlyr.

Table 1A Off Site Ambient Gamma Radiation as Measured by TLD at the Peak lnner and Outer Ring Locations Compared to the Mean of all Locations in Each Ring lnner Ring Mean lnner Ring Peak Outer Ring Mean Outer Ring Peak Year All Locations Location Mean All Locations Location Mean (mrlqtr) (mrlqtr) (mrlqtr) (mrlqtr) 1997 13.3 14.1 12.8 14.8 1998 15.0 16.4 14.4 15.9 1999 15.1 17.0 14.3 15.9 2000 15.1 16.9 14.5 16.2 2001 14.3 16.0 13.7 15.0 2002 15.9 17.4 14.8 16.2 2003 15.6 17.6 15.0 16.2 2004 16.0 18.4 15.4 16.7 2005 15.6 17.4 15.2 16.5 2006 16.5 18.6 15.6 17.0 2007 16.1 18.1 15.1 16.5 Average Mean 15.3 17.1 14.6 16.1

L-MT-09-042 Enclosure 1 Page 7 of 11 Table 2: Offsite Radiation Dose Assessments from 2001 through 2006 10 CFR 50 Appendix I Limits 10 CFR 20 10 / 20 1 15 1 5 1 15 / 15 3 1 10 100 1 Source: Annual Gaseous Releases Liquid Releases Gaseous Releases Radioactive Effluent Release Max Site Boundary Maximum Dose to Most Likely Exposed Max Dose to Individuals due to Reports for Max Offsite Dose Gamma Member of General Public (1) Activities Inside Site Boundary (1)

MNGP Organ Max Whole Whole Whole Gamma Beta Skin Thyroid Organ Thyroid Organ Body Body Body (Skin) mradlyr mradlyr mremlyr mremlyr mremlyr mremlyr mrem mrem mrem mrem mrem 2001 3.00E-03 4.00E-03 I.IOE-02 6.00E-03 7.00E-03 I.IOE-02 I.61E-05 I.72E-04 1.20E-02 1.40E-02 1.50E-02 2002 1.00E-03 2.00E-03 1.40E-02 6.00E-03 8.00E-03 I.40E-02 0.00E+00 0.00E+00 1.40E-02 1.80E-02 1.60E-02 2003 2.20E-02 1.70E-02 4.70E-02 3.90E-02 7.30E-02 4.70E-02 2.45E-07 5.55E-07 2.00E-02 3.00E-02 3.00E-02 2004 1.30E-02 1.00E-02 3.70E-02 2.20E-02 3.70E-02 3.70E-02 1.94E-I 0 1.94E-10 9.00E-03 1.IOE-02 9.00E-03 2005 3.00E-03 3.00E-03 2.50E-02 1.60E-02 2.50E-02 2.50E-02 0.00E+00 0.00E+00 1.50E-02 1.60E-02 1.90E-02 2006 1.00E-03 1.00E-03 1.40E-02 8.00E-03 6.00E-03 9.00E-03 0.00E+00 0.00E+00 8.00E-03 8.00E-03 1.00E-02 Averages 7.1 7E-03 6.17E-03 2.47E-02 1.62E-02 2.60E-02 2.38E-02 2.72E-06 2.88E-05 1.30E-02 1.62E-02 1.65E-02 Note 1: Maximum doses are calculated using the GASPAR code to provide data from the airborne pathways combined with the maximum site boundary doses.

L-MT-09-042 Enclosure I Page 8 of 11 NRC RAI No. 2(cl What is the nominal annual dose (allowing for variations from year to year) to the maximum exposed member of the public from Monticello operations under CLTP conditions? What are the contributions to this dose from N-16 shine, Nobel Gas, and other plant effluents?

NSPM RESPONSE See the NSPM response to RAI No. 2(b).

NRC RAI No. 3 The Safety Analysis Report for the Monticello Constant Power Uprate, dated October 2008, page 2-343, Table 2.10-2, indicates a possible increase in localized dose rates in the Balance-of-Plant (BOP) of up to 1130% under EPU conditions. Verify that these increases do not change the radiation zoning of the BOP spaces.

NSPM RESPONSE Post shutdown dose rates are primarily driven by the deposition of activation, corrosion, and fission products in Balance Of Plant (BOP) equipment and piping. The change of deposition sources at EPU is driven by increased moisture carryover, increased activation due to core neutron flux, and increased generation of erosionlcorrosion products due to flow increases. Carryover of radioactivity increases as a function of moisture carryover. For this evaluation it is assumed that moisture carryover will increase from 0.05 percent to 0.5 percent at EPU, an increase by a factor of 10. It is also assumed that the generation of erosionlcorrosion products in coolant increases in proportion to power (13 percent). A worst case net change is estimated as the product of these two increases (a factor of 1.13 (13 percent power uprate) times 10 (moisture carryover increase) or an 1130 percent increase).

If this worst case increase in shutdown dose rates were to occur, there are four zones in the Turbine Building that could go from a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> occupancy (dose less than Imrlhr) to as little as a 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> occupancy (dose less than 12 mrlhr). Three zones are locations within the reactor feedwater and lube oil reservoir corridor and the fourth is the feedwater pipe and cable penetration area, which are not normally occupied areas. The remaining areas affected by this potential increase are in steam piping locations in the condenser hot side area which is inaccessible during operation.

L-MT-09-042 Page 9 of 1I These areas are all located on the east end of the 911 foot Elevation, the attached dose map shows the general area dose as 1 mrlhr. Actual operating surveys, taken at full CLTP with normal hydrogen injection flowrates, show the general area dose is 0.2 mrlhr maximum. The increase of 1130 percent would result in a general area dose rate of 2.2 mrlhr, which is still considered acceptable since this is not a continuously occupied area, and it would not affect access to the other normally accessible areas of the Turbine Building.

This increase is also considered acceptable because it is a theoretical worst case estimation. Post-shutdown doses are normally very low. In most areas they are significantly less than detectable with radiation survey equipment and even this large increase will not prevent access for normal operation or maintenance. In addition, as stated on PUSAR page 2-343, this build up would occur over time and plant surveys should provide prompt detection of these conditions. Periodic and pre-maintenance surveys and monitoring are used to detect these changing conditions. Work planning and training enable workers respond to these conditions and maintain radiation exposures ALARA.

L-MT-09-042 Page 10 of II TURBINE BUILDING - ELEVATION 91%'-U*

HIGHER n-IQIN SURROUNDING G E N E W AREA DOSE RATES Wb LINE RATESARE ~4 BAREYW~?~

L-MT-09-042 Page 11 of 11 NRC RBI No. 4 The Safety Analysis Report for the Monticello Constant Power Uprate, dated October 2008, on page 2-340, within Table 2.9-1, indicates the dose consequences in the Control Room and the Technical Support Center, from a design-basis loss-of-coolant accident under EPU conditions, as 3.80 rem and 0.83 rem, respectively. Verify that these results include direct radiation exposure from plant systems containing the accident source term, consistent with the assumptions in NUREG-0737, item 11.6.2. If not, demonstrate that the direct radiation dose rates for these two vital areas meet the GDC-I9 dose criteria, as specified in NUREG-0737, item 11.6.2.

NSPMRESPONSE The Control Room and Technical Support Center (TSC) total calculated doses include a component due to direct shine dose from plant systems and the reactor building as required by NUREG-0737 Item 11.6.2. The shine contribution for the Control Room is 0.771 Rem of the total 3.8 Rem and the TSC is 0.0939 Rem of the total 0.83 Rem.