ML20011F465
| ML20011F465 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/28/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20011F463 | List: |
| References | |
| NUDOCS 9003060074 | |
| Download: ML20011F465 (6) | |
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NUCLEAR REGULATORY COMMISSION 3
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$_AFETY EVALUATION 8Y THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N05.126 AND 110 TO FACILITY OPERATING LICENSE NOS. NPF-4 AND NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION UNITS NO. 1 AND NO. _2 DOCKET NOS.50-338 AND 50-339
1.0 INTRODUCTION
By letters dated March 1, 1989, as supplemented December 22, 1989, Virginia Electric and Power Company (the licensee) proposed changes to Facility Operating License Nos. NPF-4 and NPF-7 for North Anna Units 1 and 2 (NA-1&2), respectively.
The proposed changes add a new license condition to each license stating:
"The limiting dose to the control room operators shall be revised in accordance with the licensee's submittals dated March 1, 1989 (Serial No.89-022) and December 22, 1989 (Serial No. 89-022A)."
To support these changes and for fulfillment of tio licensee's commitment to the NRC, the licensee submitted an engineering eva.uation of the radiological consequer.ces to the_NA-1&2 control room operator during and following an i
accident.
The licensee cominitted to provide such engineering evaluation and any proposed corrective actions, if needed, to ensure that multiple entries to r
the NA-1&2 control room during and following an a:cident will not cause the control room operators to exceed the dose limits specified in General Design Criterion (GDC) 19.
The licensee's commitment was brought about by a control room habitability survey conducted by the NRC in December 1986 and followup NRC inspections in September 1987 and 1988.
The December 22, 1989 letter provided additional information concerning some of the inputs used for the licensee's control room dose calculations.
This additional information did not alter in any way the staff's initial determination of no significant hazards consideration as noticed in the Federal Reaister on August 9, 1989.
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- 2. 0 EVALUATION The NA-1&2 control room habitability system for radiological protection includes a compressed breathing air system and an emergency filtered air system.
The compressed breathing air system is provided to maintain a positive interior control room pressure to ensure outward leakage when the 9003060074 900228 h
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outside air is contaminated.
The compressed breathing air system has been designed to provide 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of positive pressure.
The system is automatically initiated by the safety injection system, and may also be manually activated.
The emergency filtered air system, taking suction from the turbine building through high-ef ficiency particulate air (HEPA) and charcoal filters, is provided to ensure continued outward leakage and to continue the supply of breathing and pressurized air indefinitely upon depletion of the bottled air supply.
The licensee calculated the 30-day radiation exposures to the North Anna control room operators for the following five design-basis accidents:
P (1) loss-of-coolant accident (LOCA)
(2) main steam line breaks (MSLBs)
(3) fuel handling accident (FHA)
(4) steam generator tube rupture (SGTR)
(5) locked rotor accident (LRA)
The licensee's calculated doses indicate that the control room operator doses from these five accidents are all within the limits specified in GDC 19 of Appendix A to 10 CFR Part 50, and within the guidelines provided in Standard Review Plan (SRP) Section 6.4.
In reviewing the licensee's findings, the staff independently calculated the thyroid and whole-body doses for the LOCA, MSLB, and SGTR accidents, which the staff determined to be the most limiting accidents for radiation exposures to the control room operator.
NRC assumptions used in the calculation of the control rocm operator doses are listed in Table 1 and NRC-calculated doses are summarized in Table 2.- In the dose calculation, the staff used iodine protection factors as given in Table 1.
To calculate the consequences of the hypothetical LOCA, the staff used the conservative assumptions of Positions C.1.a through C.I.e of Regulatory Guide 1.4, Revision 2 " Assumptions Used for Evaluating the Potential Radiologice.1 Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors."
The containment structure for NA-1&2 is designed for subatmospheric operation, and during normal operation, the containment structure is maintained at a subatmospheric pressure of about 9-12 pounds per square inch absolute (psia).
For the hypothetical LOCA, the primary containment was assumed to leak at a rate of 0.01 percent per day for the first hour, and because the containment would become subatmospheric within I hour, the leak rate was assumed to be zero after I hour.
In the event of a high-energy line break accident, the containment would be depressurized and a subatmosphere condition would be reestablished within 60 minutes, and this condition would be maintained for at least 30 days following an accident.
The fraction of core inventory available for release was assumed to be 25 percent for iodine and 100 percent for noble gases.
The analysis took into account radiological
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decay during holdup in the containment, mixing in the containment, and iodine removal by the containment recirculation system.
To assess the control room operator dose due to the MSLB accident outside containment, the staff assumed that the reactor trip and initiation of steam releases both occur on receipt of a safety injection signal at time zero.
The safety injection signals automatically isolate the control room and activate the bottled air system.
The bottled air pressurizes the control room for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, after which time a filtered outside air intake of 1000 cubic feet per
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minute (ft3/ min) was assumed.
During the first hour, due to the multiple I
entries (10 ft3/ min) to the control room, the staff used an iodine protection factor of 130 from time zero to 60 minutes, and 30 from I hr to 760 hrs.
i The control room operator doses were calculated for a preaccident iodine spike i
to 60 times the NA-1&2 Technical Specifications (TS) limit of 1 microcurie per gram (pCi/gm), and no additional fuel damage was assumed to occur from this accident.
For assessing the control room operator doses from the SGTR accident, the staff assumed the break to be a double-ended rupture occurring at the top of the steam generator tube bundle while the reactor was operating at full power.
The defective steam generator was assumed to be isolated within 30 minutes of the accident. The main condenser was also assumed not available for steam dump because of the coincident loss of offsite power.
The staff further assumed a preaccident iodine spike which raised the primary coolant iodine concentration to 60 pCi/gm dose equivalent iodine-131.
The staff's calculated thyroid and whole-body doses from the hypothetical LOCA, MSLB, and SGTR accidents are summarized in Table 2.
Using the iodine protection factors and the assumptions listed in Table 1, the staff finds that despite the multiple entries to the NA-1&2 control room as described in the licensee's evaluation, control room operators will be adequately protected against an accidental radiological release, meeting the limits specified in GDC 19 of Appendix A to 10 CFR Part 50, 3.0
SUMMARY
On the basis of the above evaluation, the staff concludes that the multiple entries to the NA-1&2 control room will still leave the control room in a safe and habitable condition during and following a design basis accident and provide adequate protection against radiation so that the radiological exposure to the control room operator will be within the limits specified in GDC 19 and within the guidelines provided in SRP Section 6.4.
4.0 ENVIRONMENTAL CONSIDERATION
These amendments involve a change in the installation or use of the facilities components located within the restricted areas as defined in 10 CFR Part 20.
The staff has determined that these amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such finding.
Accordingly, these amendments meet the eligibility criteria for categorical exclusion set g
forth in 10 CFR 51.22(c)(9).
These amendments also involve changes in recordkeeping, reporting or administrative procedures or requirements.
Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.
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5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date:
February 28, 1990 E
Principal Contributors:
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Table 1 Assumptions Used in Calculating the Control Room Operator Doses Parameter Quantity Power level, Hwt 2900 Operating time, yr 3
fraction of core inventory available f or leakage, 5 locine 25 hcble gases 100 0
L Initial 1ocine composition in containment, 's Eleirer. tai 91 Organic 4
Particulate 5
Containment leak rate,1/ cay 0 - 24 hr 0.1 After 24 hr 0
3 6-Free containment volune, f t 1.84 x 10 3
- Atmospheric dispersion tactors, sec/m 0 - 8 hr 3
8 - 24 hr 8.0 x 10'3 6.3 x 10' 7.2x10]
24 - 96 hr 2.5 x 10 96 - 720 hr-3 6
Control room ' net free volune, f t 1.17 x 10 Control rcom unfiltered inleakage (cfm) 10 3
Breathing rate (m /sec) 3.47 x 10'4 Control roco, occupancy factors 0 - 24 hr 1.0 24 - 96 hr 0.6 96 - 720 hr 0.4 Containment spray renoval constants (br'1)
Elemental iocine 10 Organic iocine O
Particulate iocine 0
L-lodine partition factors Covered 100 Uncoverea 1
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.i, Table 1 (cont.)
i Parameter Quantity-Steam release from affectea steam generator (ib) 5 For MSLB accident 3.5B x 10
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For SGTR eccident 8.16 x 104 Iodine protection factors L
0 - 60 min 130 1 760 hr 30 0 - 1.5 min (for SGTR))
0 1.5 - 60 min- (for SGTR 130 E
-Accident Calculated SRP 6.4 Loss-of coolant Thyroid dose 19 30 Whole-body dosc 1.4 5
Main steam line, break Thyroid drse 30 30 Whole-body dose
<1 5
Steam generator tube rupture
-Thyroid dose 29 30 Whole-body dose
<1 5
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