ML17219A222

Forwards Response to NRC 860930 Request for Addl Info Re Util 860228 Application for Amend to License DPR-67, Extending License Expiration Date to 160301

ML17219A222
Person / Time
Site:	Saint Lucie
Issue date:	11/14/1986
From:	Woody C FLORIDA POWER & LIGHT CO.
To:	Thadani A Office of Nuclear Reactor Regulation
References
L-86-458, NUDOCS 8611180422
Download: ML17219A222 (26)

Text

REGULATORY -QRNATION DISTRIBUTION SY I'l, (R IDS>

ACCESSION NBR: 8611180422 DOC. DATE: 86/11/14 NOTARIZED: NO DOCKET FAC IL: 50-335 St. Lucie Plant> Unit 1, Florida Power 8c Light Co. 05000335 AUTH. NANE AUTHOR AFFILIATION WOODY> C. O. Florida Pommer 8< Light Co.

REC IP. NANE RECIPIENT AFFILIATION THADANI.A. G. PWR Prospect Directorate 8

SUBJECT:

Forulards response to NRC 860930 request for addi info re util 860228 application for amend to License DPR-67>

extending li cense expiration date to 160301.

DISTRIBUTION CODE: A001D COPIES RECEIVED: LTR ENCL SIZE:

TITLE: QR Submittal: General Distribution NOTES:

REC I P I ENT COPIES RECIPIENT COP IES ID CODE/NANE LTTR ENCL ID CODE/NAl'fE LTTR ENCL PWR-B EB 1 1 PWR-B PEICSB 2 2 PWR-B FQB 1 PWR-B PD8 LA 1 0 PWR-B PDS PD 01 5 5 TOURIQNY> E 1 1 PWR-B PEICSB 1 PWR-B RSB 1 INTERNAL: ADN/LFNB 1 0 ELD/HDS2 1 0 H TSCB 1 1 NRR/ORAS 1 0 REC FILE 04 1 1 RQN2 1 EXTERNAL: EC8cQ BRUSKE> S LPDR 03 1 1 NRC PDR 02 NSIC 05 1 1 TOTAL NUNBER OF COPIES REQUIRED: LTTR 23 ENCL 19

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FLORIDA POWER 5 LIGHT COMPANY NOVEIIIBER 14 1986 L-86-458 Office of Nuclear Reactor Regulation Attention: Mr. Ashok C. Thadani, Director PWR Project Directorate II8 Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Thadani:

Re: St. Lucie Unit I Docket No. 50-335 0 eratin License Ex iration Date By letter L-86-66, dated February 28, l 986, Florida Power & I ight Company (FPL) proposed to extend the date of expiration of the St. Lucie Unit I operating license so that the forty year term of the license begins with the issuance of the operating license rather than the date of issuance of the construction permit. The expiration date of the operating license of St. Lucie Unit I would be extended from July I, 20IO to March I, 20I6.

By letter dated September 30, l986 (E. G. Tourigny to C. O. Woody), the NRC identified additional information required to continue its review of the proposed license amendment. The attached responds to the staff's request for additional information.

Please contact us if you have n

any questions about this submittal.

Very truly yours,

. O. Woody roup Vice President Nuclear Energy COW/E JW/gp Attachment cc: Dr. J. Nelson Grace, USNRC, Region II Harold F. Reis, Esquire, Newman & Holtzinger 8bi1180422 8b1114 PDR ADQCK 05000335 p PDR E JW2/002/I PEOPLE... SERVING PEOPLE

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REQUEST FOR ADDITIONALINFORMATION L'ICENSE EXTENSION FOR ST. LUCIE UNIT I REQUEST I. ALARA MEASURES Discuss specific ALARA measures which will be utilized during the seven additional years of service.

RESPONSE I.

FPL COMMITMENT TO ALARA Florida Power and Light Company is committed to ensure that radiation exposure to personnel is kept as low as reasonably achievable (ALARA). ALARA means that anytime personnel exposure can be effectively reduced without excessive cost, it should be done. This includes total person-rem exposure for the facility as well as individual exposure. This requires a general awareness by all personnel that they should make every reasonable effort to avoid any exposure that is not necessary to accomplish their work assignment. Examples of ALARA techniques that are evaluated include adequate training of personnel working in the radiation controlled area (RCA), preplanning of job activities, installation of shielding, and design review of new facilities or plant modifications. The present emphasis on ALARA is the result of changes in regulations and issuance of regulatory guides that are intended to promote a more formal (documented)approach to the ALARA concept.

Radiation protection is the responsibility of all personnel working in a nuclear plant; therefore, each individual is responsible to ensure that his exposure, as well as all other personnel's exposure, is maintained ALARA.

ALARA EXECUTIVE ORGANIZATION The Group Vice President of Nuclear Energy has the overall responsibilty for administering the Company ALARA program.

The Vice President of 'Nuclear Operations has the responsibility to ensure implementation of the Company ALARA program and to evaluate periodically the effectiveness of the program.

The Plant Manager has the responsibility to implement the overall plant ALARA program ensuring the effectiveness of this program and that sufficient resources are available within the plant for support of program implementation.

The Corporate Health Physicist provides basic guidelines for the implementation of ALARA concepts, implementation of the corporate policy and keeping management abreast of the effectiveness of the ALARA program, goals and objectives.

The Plant Health Physics Supervisor has the responsibility to develop, maintain, and supervise the health physics ALARA program at the plant. He is also responsible for periodically evaluating the effectiveness of the program and keeping the Plant Manager abreast of its effectiveness, goals and objectives.

The Health Physics ALARA Coordinator has, as his primary responsibility, the coordination, implementation and monitoring of the health physics program in accordance with established health physics procedures.

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The Health Physics Training Supervisor is responsible for ensuring that an effective training program is maintained stressing the importance of ALARA in personnel's daily activities.

Each department supervisor and contract project leader has the overall responsibility to ensure that every individual under his supervision adheres to the established ALARA program. He verifies the effectiveness of the implementation of this program in his department by routine review of person-rem accumulated, saved and trending and the preplanning of job activities. He is also responsible to incorporate ALARA concepts in design modifications and new facilities.

OPERATIONAL GUIDELINES The following guidelines should be followed when assessing the need for preplanning job/task activities to maintain exposure as low as reasonably achievable:

I. For any task where the collective dose is estimated to be less than I person-rem, no special formal ALARA documentation other than a Radiation Work Permit (RWP) (See Radiation Work Permits below for conditions requiring an RWP) is required.

2. For any task where the collective dose is estimated to exceed I person-rem, additional ALARA review should be performed, documented and, if necessary, implemented prior to issuance of an RWP.

3. For any task where the collective dose is estimated to exceed 50 person-rern, in addition to an ALARA review as stated in item 2 above, an historical review of similar tasks performed and the effectiveness of the ALARA techniques used is performed prior to issuance of an RWP.

4,, For any task where the collective dose is estimated to exceed 50 person-rem, in addition to item 3 above, a formal, documented, post-operational review of the task documenting the degree of success (or failure) of ALARA techniques used is performed.

Documentation of ALARA reviews is maintained for the specific tasks stated above, as appropriate.

A plant ALARA Review Board periodically evaluates the effectiveness of the plant's ALARA program. The appropriate board personnel normally meet on a quarterly basis and/or after each major outage.

RADIATIONWORK PERMITS The primary purpose of a Radiation Work Permit (RWP) is to provide Health Physics with a vehicle whereby it can evaluate and plan jobs in order to maintain radiation exposure ALARA. The Florida Power & Light Company RWP philosophy is based on the fact that control of radiation and contamination is accomplished primarily by training, Health Physics job surveillance, pre-job planning, post-job evaluation, and special instructions. An RWP normally describes the radiological conditions of a job, the protective clothing, monitoring to be performed, and any other special instructions.

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RWP REQUIREMENTS An RWP shall be required for the following conditions.

I. Entry into high radiation areas, airborne radioactivity areas, areas contaminated to levels in excess of l0,000 dpm/l00 cm,or into any area posted as "RWP REQUIRED FOR ENTRY"

2. Entry into the reactor containment at any time during and subsequent to initial reactor startup.

3. Maintenance or inspection of equipment contaminated in excess of I0,000 dpm/l00 cm .

4. Work assignments involving changes (withdrawing, uncovering, opening, valving, disassembling, moving) that have the following potential as the work progresses:

a. Exposure of a major portion of the body to a radiation dose in excess of I 00 milli-rem in any one hour.

b. Inc~easing surface contamination levels to exceed I 0,000 dpm/ l00 cm ~

c. Increasing airborne radioactivity concentration exceeding 25% of those listed or referred to in IOCFR20 Appendix B, Table I, Column I.

MAINTAININGALARA IN THE FUTURE FPL will remain committed to ALARA in the future. Potential enhancements currently under review and/or implementation are ALARA awareness, Robotics and Data Base Management.

ALARA Awareness Methods and data sheets are being developed to allow every department, through the ALARA Coordinator, to establish an awareness of exposure control goals, track the progress towards those goals and select tasks on which the data indicates the need for special emphasis on radiation exposure reduction.

Robotics Robotics, as they are developed and approved for use, have the potential for large scale exposure reduction. Robotics have been used for steam generator work and will continue to be evaluated and applied, as appropriate.

Data Base Mana ement Future efforts in Data Base Management may result in exposure reduction by identifying which jobs in a major evolution result in the most exposure. This serves to target additional exposure control efforts. Subsequent exposure is then tracked to determine the effectiveness of any changes made.

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REQUEST 2. DOSE ASSESSMENT a) Provide a table showing St. Lucie, Unit I personnel exposure experiences for 19 8 1988 1 81*

regardless of how these exposures were obtained (e.g., during normal operations, maintenance, repair or refueling activities) and by whom (e.g.,

by plant operations personnel, plant maintenance personnel, contractor/vendor personnel, etc.)

RESPONSE 2.

a) The yearly exposure data and outage reports for the years l977 to l985 were reviewed to determine the man-rem exposures attributable to routine operations and refueling outages. The normal operations exposure is determined as the difference between the site annual total and outage totals.

Each outage is divided into l5 general categories such as pressurizer work, steam generator primary side, etc. The categories are further sub-divided into the applicable RWP numbers and the total man-hours and man-rem data are listed. Table I provides the Refueling Outage Summary, and Table 2 provides the Outage Exposure Summary, Category 6 Year.

REQUEST 2.

b) Provide a similar table for the years 20IO to 20I6. Include doses from expected decontamination, decommissioning, additional maintenance and related doses. Provide discussion which confirms that your ALARA program will maintain the state-of-the-art for reducing personnel exposures to a minimum.

RESPONSE 2.

b) An estimate of dose from decommissioning and extensive decontamination was not attempted. Utilizing the data from Tables I and 3, predictions of dose for the additional years can be made. The predicted value is l60 rnan-rem for a non-outage year and 760 man-rem for a refueling outage year. The net Man-Rem is based on typical refueling outage conditions. The exposures attributable to special projects, e.g. thermal shield removal, coolant pump modifications, etc., are not included in the data base for the projections.

This prediction is being treated as an upper value. Dose allowance for crud build-up will be offset by dose savings from a continually improving ALARA program. It is expected that state-of-the-art technologies will be in use including robotics, enhanced chemistry control and modern decontamination processes.

Noting the projected refueling outage schedule for Unit I (Table 4) and assuming eighteen month fuel cycles, for the years 20l I to 20I6, there will be four additional refueling cycle years and two non-refueling cycle years. For these additional years, the predicted total dose is 3360 man-rem.

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TABLE I REFUELING OUTAGE

SUMMARY

Note: This Table does not include refueling dose associatied with 1983 - 1984 outage, which was compounded with Thermal Shield Removal and Core Support Barrel repair Gross Adjustments Net Year Man-Rem Man-rem Descri tion Man-Rem 1978 144.25 (6.7) Neutron Shielding 131.85 (5.7) F lammastic Application (12.4) 1979 263.024 (14.2) S/G Rim Cut 248.82 1980 446.880 (12.2) RX Head Stud Modifications 400.56 (I 6.4) Complete Containment Painting (15.12) TMI Modifications (2.60) PAR lnc. Equip Modifications (46.32) 1981 719.829 (159.31) RCP Modifications 504.42 (32.34) Containment Modifications (6.82) ICI Flange Modification (9.54) TMI Modifications (4.80) UGS Modifications (2.60) Hot Tool Room Construction (215.41) 1985 1025.835 (254.93) S/G Mods for Nozzle Darns 651.8 (44.80) S/G Tube Pull (42.50) S/G Tube Sheet Work (21.56) S/G Secondary Special Work (l0.26) Plant Modifications (374.05)

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TABLE 2 OUTAGE EXPOSURE

SUMMARY

, CATEGORY & YEAR Man Man Man Activit Cate or Year Hours(MH) REM(MR)

Decon & Clean Up 1978 1,282.4 11.375 1979 972.4 7.530 1980 215.9 2.825 1981 ',351.5 28.290 1982 116.8 1.205

'83-'84 35,423.3 172.880 1985 8 719.0 85.610 (TOT) 52,081.3 (TOT) 309.715 Reactor Head Work 1978 3,318.7 60.070 1979 3,617.6 80.822 1980 5,528.1 161.925 1981 4,269.1 106.815 1982 Steam Generator Inspection Outage

'83-'84 5,344.0 117.880 1985 4 356.5 82.340 (TOT) 26,434.0 (TOT) 609.852 Pressurizer Work 1978 170.5 1.520 1979 506.6 8.470 1980 1,094.3 16.718 1981 800.1 -10.960 1982 296.4 10.355

'83-'84 301.9 6.465 1985 356.7 9.050 (TOT) 3,526.5 (TOT) 63.538 Keyway 1978 304.2 0.945 .003 1979 349.2 1.950 1980 5,686.7 16.400 1981 134.0 1.195 1982 18.0 0.340

'83-'84 9,209.2 59.83 1985 79.9 3.63 (TOT) 15,981.2 (TOT) 84.29 E JW2/002/7

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OUTAGE EXPOSURE

SUMMARY

, CATEGORY & YEAR (cont'd)

Man Man Man Activit Cate or Year Hours (MH) Rem (MR)

Steam Generator 1978 381.8 6.,1 75 Primary Job 1979 I,'156.0 16.755 1980 1,165.0 13.505 1981 . 2,874.9 130.160 1982 1,048.0 24.360

'83-'84 6,988.1 396.185 1985 7 I I3.2 458.305 (TOT) 20,727.0 (TOT) 1,045.895 Steam Generator 1978 55.5 4.320 Secondary Side 1979 306.3 5.940 1980 81.9 3.420 1981 13.7 4.200 1982 48.4 0.775

'83-'84 231.5 7.860 1985 975.5 25.300 (TOT) 1,712.8 (TOT) 51.915 Reactor Coolant Pumps 1978 802.2 10.470 1979 2,145.3 23.009 1980 3933203 35.857 1981 5,719.0 64.052 1982 386.2 9.255

'83-'84 1,788.7 40.170 1985 2 708.9 47.105 (TOT) 16,882.6 (TOT) 229.918 Refueling Work 1978 477.3 3.130 1979 1,384.4 8.200 1980 862.5 7.255 1981 1,920.2 15.760 1982 Steam Generator Inspection Outage

'83-'84 5,212.7 36.740 1985 I 654.2 I4.865 (TOT) 11,511.3 (TOT) 85.95 E JW2/002/8

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TABLE 2 (Cont'd)

OUTAGE EXPOSURE

SUMMARY

, CATEGORY & YEAR (cont'd)

Man Man Man Activit Cate or Year Hours (MH) Rem (MR)

Routine Maintenance 1978 1,773.9 23.460 1979 2,062.6 33.210 1980 4,038.1 56.635 1981 5,575.6,,'0.236 1982 227.5 2.915

'83-'84 3,288.7 23.915 1985 I 346.2 9.845 (TOT) 18,312.6 (TOT) 210.216 Special Maintenance 1978 719.6 10.385 1979 587.9 4.555 1980 213.9 2.870 1981 1,098.8 16.360 1982 195.0 3.395 83 '84 16,090.3 148.340 1985 12 593.4 109.175 (TOT) 31,498.9 (TOT) 295.080 Plant Modifications 1978 2,144.9 6.010 1979 424.4 14.710 1980 5,098.3 29.790 1981 18,424.1 182.175 1982 No Plant Mods during S/G Outage

'83-'84 28,764.8 191.40 1985 3 230.I 10.565 (TOT) 58,086.6 (TOT) 434.39 Calibration 1978 Cal Data Not Segregated before 1980 1979 1980 4.0 0.135 1981 679.0 7.165 1982 101.1 2.245

'83-'84 899.6 11.765 1985 913.9 8.745 (TOT) 2,597.6 (TOT) 30.055 E JW2/002/9

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OUTAGE EXPOSURE

SUMMARY

, CATEGORY & YEAR (cont'd)

Man Man Man Activit Cate or Year Hours (MH) Rem (MR)

Ventilation, 1978 Data not segregated/Work not done Filters, Coolers, etc. 1979 37.6 0.435 1980 46.5 1.975 1981 33.1 0.350 1982 10.7 0.140

'83-'84 964.8 3.215 1985 ~

412.4 2.160 (TOT) 1,505.1 (TOT) 8.275 General Entries & 1978 551.0 4.995 Inspections 1979 7,715.8 48.198 1980 10,523.4 95.885 1981 8557.1 73.871 1982 1,379.0 18.710

'83-'84 38,052.1 244.505 1985 21 240.9 147.850 (TOT) 88,019.3 (TOT) 634.014 Miscellaneous 1978 110.4 1.490 1979 1,413.6 9.240 1980 166.7 1.670 1981 1,942.2 17.790 1982 51.8 1.035

'83-'84 Misc Dose Data included in TSR/CSBR (below) 1985 2 889.4 I I. I 90 (TOT) 6,574.1 (TOT) 42.415 Thermal Shield '83-'84 59,105.8 504.920 Removal/Core Support Barrel Repair (TSR/CSBR)

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TABLE 3 DERIVATION OF "NORMAL OPERATIONS" EXPOSURE Total Outage Net Oprng Year Man-Rem Man-Rem Man-Rem 1977 142.2 No outages 142.2 1978 317.54 144.25 173.29 1979 420.25 263.02 157.23 1980 522.44 446.88 75.56 1981 889.34 719.83 169.51 1982 254.25 74.73 179.52 1983 +1,141.33 659.94 %481.41 240.7 1984 +1,196.89 917.68 279.21 139.6 1985 <<1,274.07 1,025.835 %248.24 124.1 Total 1,401.71 Average 156

+ The total Man-Rem now includes two fully operating reactors.

The net operating dose is divided equally between the units.

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TABLE 4 ST. LUCIE UNIT REMAINING AND PROJECTED REFUELING OUTAGES Year Month ~Cele 1985 October 1986 -None-1987 April 1988 October 1989 -None-1990 April 10 1991 October II 1992 -None-1993 April 12 1994 October 13 1995 -None-1996 April 14 1997 October 15 1998 -None-1999 April 16 2000 October 17 2001 -None-2002 April 18 2003 October 19 2004 -None-2005 April 20 2006 October 21 2007 -None-2008 April 22 2009 October 23 2010 -None- End of Current License Man-Rem 2011 April 24 760 2012 October 25 760 2013 -None- 160 2014 April 25 760 2015 October 27 760 2016 -None- 160 End of License 3360 E JW2/002/12

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