ML18153B956

Application for Amends to Licenses DPR-32 & DPR-37,adding License Condition Stating That Current Assessment of Control Room Dose Calculations/Habitability Documented in 890601 Submittal & Limiting Predicted Control Room Doses Revised

ML18153B956
Person / Time
Site:	Surry
Issue date:	10/26/1989
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
89-381A, NUDOCS 8911060322
Download: ML18153B956 (14)

Text

. .

_.. :--:r e

..- VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 10CFR50.90 October 26, 1989 United States Nuclear Regulatory Commission Serial No. 89-381A Attention: Document Control Desk NO/CGL:vlh R2 Washingt<;m, D.C. 20555 Docket No. 50-280 50-281 License No. DPR-32 DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 CONTROL ROOM DOSE CALCULATIONS/HABITABILITY ASSESSMENT PROPOSED OPERATING LICENSE AMENDMENT By letter Serial No.89-381, dated June 1, 1989, Virginia Electric and Power Company submitted an assessment of control room habitability, including a reevaluation of control room dose consequences. The limiting control room doses identified in this reevaluation exceeded those previously reported and, therefore, this item was determined to be an unreviewed safety question, as defined by 10CFR50.59. In each case evaluated, the predicted control room doses remain within the limits of 10CFR50, Appendix A, General Design Criteria 19. Attachment 1 presents a summary of the thirty day control room dose calculation results documented in letter Serial No.89-381.

Pursuant to 10CFR50.59(c), the Virginia Electric and Power Company requests a license amendment, per the requirements of 10CFR50.90, for Surry Units 1 and 2.

The amendment would add a license condition stating that the current assessment of the control room dose calculations/habitability is documented in letter Serial No.89-381 and that the limiting predicted control room doses are revised in accordance with that submittal. Attachment 2 provides proposed license amendments addressing this item for Units 1 and 2.

Attachment 3 provides a summary of the bases for the atmospheric dispersion factors used in the Surry control room habitability calculations. This supplemental information was requested by the NRC in a June 22, 1989 telephone conversation in order to allow the NRC to perform an independent evaluation of the control room dose calculations. In addition to identifying the reference for the methodology used, Attachment 3 identifies the following information for each accident discussed in letter Serial No. 89-381:

a. Control room atmospheric dispersion factors (X/0 values)

b. Control room occupancy factors

c. Distance between release point and control room ventilation intake 8911.060322 891026 PDR ADOCK 05000280  !

P PNU J

"d. Containment projected area (if used in determination of building wake effects)

e. Wind speed used

f. Assumed atmospheric stability class

g. Building wake effects As stated above, the reevaluated limiting control room doses exceeded those previously reported and, therefore, this item was identified as an unreviewed safety question. However, the reevaluation of the Surry control room doses does not involve a significant hazards consideration. The basis for no significant hazards determination was originally submitted via letter Serial No.89-381 and is resubmitted for completeness as Attachment 4.

The Surry Power Station UFSAR will be revised accordingly to reflect the information documented in this letter and letter Serial No.89-381.

If you have questions regarding this submittal, please contact us.

Very truly yours,

• I

\ l --

! ~

~ L~tewa~

• Senior Vice President - Power Attachments

1. Summary of Control Room Dose Calculations

2. Proposed Units 1 and 2 Operating License Amendments

3. Requested Additional Information on Surry Control Room Dose Calculations

4. Basis for No Significant Hazards Determination cc: U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, N.W.

Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Commissioner Department of Health Room 400 109 Governor Street Richmond, Virginia 23219

COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by W. L. Stewart who is Senior Vice President - Power, of Virginia Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this ~ d a y of ([}~l_ ' 1989.

My Commission Expires: ~/Jiiu411 d5, 19~

___da~

Notary Public (SEAL)

ATTACHMENT 1 Summary of Control Room Dose Calculations

e

SUMMARY

OF CONTROL ROOM DOSE CALCULATIONS 30 Day Doses in the Control Room

(

Reference:

Letter Serial No.89-381, dated 6/1/89)

Thyroid Gamma Beta Accident Dose (Rem) Dose (Rem) Dose (Rem)

LOCA-*uprated 1000 cfm intake 26.6 0.51 1.25 4000 cfm intake 20.9 0.53 1.42 MSLB with pre-accident 1.43 2.24 E-4 9.03 E-3 iodine spike MSLB with concurrent 1.68 2.73 E-4 9.48 E-2 iodine spike FHA 0.919 1.87 E-3 0.120 SGTR with pre-accident 16.2 7.4 E-3 0.295 iodine spike SGTR with concurrent 1 .81 6.7 E-3 0.290 iodine spike VCT Rupture 0.594 24.2 WGDT Rupture 0.505 19.7 Allowable Limits 30.0 5.0 30.0

ATTACHMENT 2 Proposed Units 1 and 2 Operating License Amendments

L. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved Nuclear Security Personnel Training and Qualifications Program, including amendments and changes made pursuant to 10CFR50.54(p). The approved Nuclear Security Personnel Training and Qualifications Program consists of a document withheld from public disclosure pursuant to 10CFR2.790(d) identified as "Surry Power Station Nuclear Security Personnel Training and Qualifications Program" dated September 15, 1980. The Nuclear Security Personnel Training and Qualification Program shall be fully implemented in accordance with 10CFR73.55(b)(4), within 60 days of this approval by the Commission. All security personnel shall be qualified within two years of this approval.

M. The design of the reactor coolant pump and steam generator supports may be revised in accordance with the licensee's submittals dated November 5, 1985 (Serial No.85-136), December 3, 1985 (Serial No.

85-136A), and January 14, 1986 (Serial No. 85-136C).

N. The current assessment of control room dose calculations/habitability is documented in the licensee's submittal dated June 1, 1989 (Serial No.89-381 ). The limiting predicted control room doses, which remain within the limits of 10CFR50, Appendix A, General Design Criteria 19, are revised in accordance with the above-mentioned submittal.

4. This license is effective as of the date of issuance, and shall expire at midnight May 25, 2012.

FOR THE ATOMIC ENERGY COMMISSION Original signed by A. Giambusso A. Giambusso, Deputy Director for Reactor Projects Directorate of Licensing Enclosure Appendix A -

Technical Specifications Date of Issuance: May 25, 1972 Surry - Unit 1

e L. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved Nuclear Security Personnel Training and Qualifications Program, including amendments and changes made pursuant to 10CFR50.54(p). The approved Nuclear Security Personnel Training and Qualifications Program consists of a document withheld from public disclosure pursuant to 10CFR2.790(d) identified as "Surry Power Station Nuclear Security Personnel Training and Qualifications Program" dated September 15, 1980. The Nuclear Security Personnel Training and Qualification Program shall be fully implemented in accordance with 10CFR73.55(b)(4), within 60 days of this approval by the Commission. All security personnel shall be qualified within two years of this approval.

M. The design of the reactor coolant pump and steam generator supports may be revised in. accordance with the licensee's submittals dated November 5, 1985 (Serial No.85-136), December 3, 1985 (Serial No.

85-136A), and January 14, 1986 (Serial No. 85-136C).

N. The current assessment of control room dose calculations/habitability is documented in the licensee's submittal dated June 1, 1989 (Serial No.89-381 ). The limiting predicted control room doses, which remain within the limits of 10CFRSO, Appendix A, General Design Criteria 19, are revised in accordance with the above-mentioned submittal.

4. This license is effective as of the date of issuance, and shall expire at midnight May 25, 2012.

FOR THE ATOMIC ENERGY COMMISSION Original signed by A. Giambusso A. Giambusso, Deputy Director for Reactor Projects Directorate of Licensing Enclosure Appendix A -

Technical Specifications Date of Issuance: May 25, 1972 Surry - Unit 2

e e ATTACHMENT 3 Requested Additional Information on Surry Control Room Dose Calculations

NRC Reguest for Additional Information:

For the control room dose calculations submitted by letter Serial No.89-381, please provide the following information:

a. Control room XIQ values (particularly for the LOCA) and how those values were derived

b. Distances between the containment (i.e., activity release points) and the control room intakes

c. Wind speed assumptions

Response

The attached Table 1 identifies the bases for the atmospheric dispersion factors, which were used in the Surry control room habitability calculations. For each accident discussed in our submittal, the following information is provided:

a. Control room atmospheric dispersion factors (X/Q values)

b. Control room occupancy factors

c. Distance between release points and control room ventilation intake

d. Containment projected area (if used in determination of building wake effects)

e. Wind speed used

f. Assumed atmospheric stability class

g. Building wake effects Note that the fifth and sixth entries in Table 1 (Unit II Containment Building and ECCS Leakage and Unit I Containment Building) pertain to the evaluation of the control room doses due to a LOCA. To conservatively bound both units, the worst case atmospheric dispersion factors were used for the control room dose calculations. The limiting control room dose for a LOCA is the cumulative dose due to containment leakage, ECCS leakage, direct containment shine, and sky shine.

The basic reference for the methodology used in developing the control room X/0 values is the Murphy and Campe paper, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19,"

which was presented at the 13th AEC Air Cleaning Conference in August 1974.

In addition, it should be noted that:

1. The control room occupancy factors which were used to adjust the X/Q values were also taken from the Murphy and Campe paper.

2. The distances between the release points and the control room intakes were determined from a plot plan of the Surry Power Station.

3. For the main steam line break, steam generator tube rupture, fuel handling, and LOCA accidents, the building wake dispersion correction factors were taken from NRC Regulatory Guides (e.g., Figure 1 of Regulatory Guide 1.4) using a previously calculated containment projected area. This area, determined from drawings of the Surry containment structures, was used in previous calculations documented in letter Serial No. 335, dated June 7, 1982.

For the waste gas decay tank rupture and volume control tank rupture accidents, a factor of 3 credit was applied for the building wake effects, as recommended by Murphy and Campe for calculations involving point sources and point receptors.

4. The wind speeds were determined from site meteorological data 'for the 1974 to 1987 time period in accordance with the referenced Murphy and Campe paper. Pasquill-F atmospheric stability was assumed throughout the calculation of the control room X/0 values, also as recommended by Murphy and Campe.

I TABLE 1 METHODOLOGY USED TO CALCULATE ACCIDENT X/Q VALUES FOR ONSITE CONTROL ROOM HABIT ABILITY

~*

Distances Between Release Points Accident X/Q Values Control Room and Control Room Containment Wind Stability Building Condition (sectm 3 ) Methodology Occupancy Factors Ventilation Intakes Projected Area Speed Class Wake Effects MAIN STEAM 0-8 hr 3.79 X 10-3 0-8 hr 1.00 38.57 m 1456m2 .94 mis Pasquill-F A/2 LINE BREAK 8-24 hr 3.09 X 10-3 8-24 hr 1.00 1-4 day 1.05 X 10-3 (1) 1-4 day 0.60 FUEL 4~Qday 0-8 hr 2.49 X 1Q-4 2.98 X 10-3 4-3Q~

0-8 hr Q.4Q 1.00 51.76 m 1456m2 .72 mis Pasquill-F A/2 e

HANDLING 8-24 hr 1.79 X 10-3 8-24 hr 1.00 1-4 day 6.26 X 10-4 (1) 1-4 day 0.60 4-30day 8.94x 10-5 4-30day 0.40 WASTE GAS 0-8 hr 4.35 X 10-2 0-8 hr 1.00 57.85 m Not used .89 mis Pasquill-F Factor of 3 DECAY TANK 8-24 hr 2.57 X 10-2 8-24 hr 1.00 credit 1-4 day 8.70 X 10-3 (1) 1-4 day 0.60 4~Qday :1.a o-1~ 1 3 ~Qday Q.40 VOLUME 0-8 hr 4.83 X 10-2 0-8 hr 1.00 57.85 m Not used .80 mis Pasquill-F Factor of 3 CONTROL 8-24 hr 2.90 X 10-2 8-24 hr 1.00 credit TANK 1-4 day 9.66 X 10-3 (1) 1-4 day 0.60 4-JQday 1.45~ jQ-3 4-3Qday Q,40 UNITII 0-8 hr 4.02 X 10-3 0-8 hr 1.00 46.5 m 1456m2 .76 mis Pasquill-F A/2 CONTAINMENT 8-24 hr 2.41 X 10-3 8-24 hr 1.00 BUILDING AND 1-4 day 8.44 X 10-4 (1) 1-4 day 0.60 ECCS LEAKAG!; ~Qday 1,61 X lQ 4-3Q~ Q.40 UNITI 0-8 hr 4.07 X 10-3 0-8 hr 1.00 38.1 m 1456m2 .89 mis Pasquill-F A/2 CONTAINMENT 8-24 hr 2.52 X 10-3 8-24 hr 1.00 BUILDING 1-4 day 8.55 X 10-4 (1) 1-4 day 0.60

~~Qday l.63X lQ 4-3Q~ Q,40 STEAM 0-8 hr 3.79 X 10-3 0-8 hr 1.00 38.57 m 1456m2 .94 mis Pasquill-F A/2 GENERATOR 8-24 hr 3.09 X 10-3 8-24 hr 1.00 TUBE 1-4 day 1.15 X 10-3 (1) 1-4 day 0.60 RUPTURE 4-30 day 2.49 x 1o-4 4-30 day 0.40 Refecences: (1) Murphy, K. G., and Campe, K. M., "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19,"

Presented at the 13th AEC Air Cleaning Conference.

L_ ----

-~

ATTACHMENT 4 Basis for No Significant Hazards Determination

• , e BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION It has been determined that the reevaluation of the Surry control room doses does not involve a significant hazards consideration as described in 10 CFR 50.92. The results of this determination can be stated as follows:

1. There is no significant change in the probability or consequences of an accident previously evaluated. The revision is to the analytical evaluation of limiting control room doses by consideration of a broader spectrum of accidents and recent updates in meteorological assumptions based on onsite measurements.

There are no system changes which increase the probability of an accident occurring. Although the limiting doses to the control room were found to increase, the increases are not considered to be significant because the revised doses remain well below the limits of 10CFR50, Appendix A, General Design Criteria 19 and meet the guidelines of NUREG-0800 (Section 6.4).

2. No new accident types or equipment malfunction scenarios have been introduced. The revision is analytical, not physical, and, therefore, the possibility of an accident of a different type than any evaluated previously in the UFSAR is not created.

3. There is no significant reduction in the margin of safety. The revised dose calculations for the control room continue to meet the requirements of General Design Criteria 19.