ML20078Q361
| ML20078Q361 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 02/09/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078Q325 | List: |
| References | |
| NUDOCS 9502210218 | |
| Download: ML20078Q361 (8) | |
Text
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+4 UNITED STATES g
,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 69 AND 58 TO FACILITY OPERATING LICENSE NOS. NPF-76 AND NPF-80 HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NOS. 50-498 AND 50-499 SOUTH TEXAS PROJECT. UNITS 1 AND 2
1.0 INTRODUCTION
By application dated November 7,1994, Houston Lighting & Power Company, et.
al., (the licensee) requested changes to the Technical Specifications (TSs)
(Appendix A to Facility Operating License Nos. NPF-76 and NPF-80) for the South Texas Project, Units 1 and 2 (STP). The proposed changes would permit both containment personnel airlock doors to be open while moving fuel during refueling operations. Specifically, the amendments would revise TS 3/4.9.4, Containment Building Penetrations, to permit airlocks to be continuously open during fuel movement and core alterations provided that there is 23 feet of water above the reactor vessel flange and a qualified individual is available to close at least one airlock door within 30 minutes or 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, depending on the time since last criticality.
Blocking of the doorways by cables, hoses, etc., would be prohibited, however, removable protective devices would be permitted to be installed on the door seals and sealing surfaces.
The STP facilities are 3800 megawatts thermal Westinghouse-designed pressurized water reactor plants located 12 miles south-southwest of Bay City, TX. The STP containments are of the large, dry, post-tensioned, reinforced concrete type.
2.0 DISCUSSION AND EVALUATION Airlocks The STP containments each contain a personnel access lock (PAL) connecting the containment interior with the Mechanical and Electrical Auxiliary Building.
The PAL is provided for the purpose of permitting personnel to enter and exit the containment while maintaining the integrity of the containment pressure boundary during power operation and certain shutdown operations.
It has two 9502210218 950209 PDR ADOCK 05000498 P
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8-feet high, 5-feet wide, airlock doors with a 11%-foot diameter, 8-foot long personnel chamber between the doors.
Each door is provided with a double inflatable seal. The doors are hydraulically operated.
Electrical and mechanical interlocks ensure that both doors cannot be opened at the same time. When neither core alterations nor movement of irradiated fuel in containment are taking place, the interlock mechanism may be intentionally disabled allowing both doors to be opened at the same time.
Each containment is also provided with an auxiliary airlock.
The auxiliary airlock is 10-feet long and has a diameter of 5\\ feet.
The auxiliary airlock doors are 30 inches in diameter and mechanically interlocked. The proposed amendment applies to the PAL only.
The licensee proposes to revise the TSs to remove the restriction prohibiting the opening of both PAL doors at the same time, for periods when certain conditions are met. This would reduce airlock door wear (eliminating approximately 1500 airlock door open/close cycles per outage) and would facilitate personnel access.
The proposed conditions are that (a) there is at least 23 feet of water over the reactor vessel flange, (b) at least 95 hours0.0011 days <br />0.0264 hours <br />1.570767e-4 weeks <br />3.61475e-5 months <br /> has elapsed since last reactor criticality, (c) a designated individual is available to close the airlock within either 30 minutes or 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> depending on the time since last criticality (the 2-hour condition would apply when the time since criticality is 2165 hours0.0251 days <br />0.601 hours <br />0.00358 weeks <br />8.237825e-4 months <br />).
Standard Graic Reauirements Reaardina Airlock Penetration Intecrity Durino Core Alterattani n
The applicable staff positions regarding opening of airlock doors during Mode 6 (Refueling Operations) are stated in Section 3.9.4 (BASES) of the Improved Standard Technical Specifications (NUREG-1431, " Standard Technical Specifications for Westinghouse Plants" or "ISTS").
Text excerpted from the ISTS states:
The containment air locks, which are part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 operation.
During periods of shutdown when containment cl sure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.
During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, containment closure is required; therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain closed.
The requirements on containment penetration closure ensure that a release of fission pi Muct radioactivity within containment will be restricted from escaping to the environment.
The closure restrictions are sufficient to
e
. restrict fission product radioactivity release from containment due to a fuel handling accident during refueling.
During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, the most severe radiological consequences result from a fuel handling accident.
The fuel handling accident is a postulated event that involves damage to irradiated fuel. Fuel handling accidents include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies. The minimum decay time of [72] hours prior to CORE ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are well within the guideline values spedfied in 10 CFR 100. The acceptance limits for offsite radiation exposure are contained in Standard Review Plan Section 15.7.4, Rev. 1, which defines "well within" 10 CFR 100 to be 25% or less of the 10 CFR 100 values.
As indicated above, the basis for the staff position against simultaneous opening of both airlock doors during core alterations is to limit fission product leakage in the event of a design basis fuel handling accident.
In performing analyses of the radiological consequences of a fuel handling accident, the criteria of Standard Review Plan Section 15.7.4 are used.
If fuel handling is prohibited when the containment is open, radiological consequences need not be calculated.
Standard Technical Specifications thus specify that airlock integrity be continuously maintained during fuel handling in containment.
If the containment will be open during fuel handling operations, automatic isolation by radiation detection instrumentation must be provided for penetrations and calculations must demonstrate acceptable consequences. However, automatic isolation of airlock doors is not practicable. The licensee has shown by analysis that the requirement for airlock closure need not be applied to STP.
STP Fuel Handlina Accident Analysis The licensee performed analyses of the radiological consequences of a fuel handling accident with the PAL doors open.
In performing the analysis the licensee used the assumptions and methodology prescribed by Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors." A 95-hour decay time (subcritical period) was assumed in the Case 1 analysis and a 165-hour decay time in the Case 2 analysis.
Both analyses assumed that there was 23-foot water coverage above the reactor vessel flange. The licensee's analyses indicated that for Case 1, radiological dose consequences acceptance criteria are met if the airlock doors are closed in 30 minutes.
For Case 2, the dose criteria are met with the airlock doors open for the entire assumed 2-hour release period.
The staff did not review the licensee's analysis, but instead performed an independent analysis of the potential radiological consequences of a fuel handling accident that bounds both the licensee's Case 1 and Case 2 analyses, based upon the conditions of the proposed TS change. The staff's analysis was performed to determine conformance with the requirements of 10 CFR Part 100 (for offsite doses) and General Design Criterion (GDC) 19 of Appendix A to 10 CFR Part 50 (for doses to control room operators). The staff's analysis utilized the accident source term given in Regulatory Gu
1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," the assumptions contained in Regulatory Guide 1.25, and the review procedures specified in Standard Review Plan (SRP) Sections 15.7.4 and 6.4.
The staff assumed an instantaneous puff release of noble gases and radiciodines from the gap and plenum of the broken fuel rods as gas bubbles pass through 23 feet of water covering the fuel. All airborne activity reaching the containment atmosphere is exhausted to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. As stipulated in the proposed TS change, the gap l
activity is assumed to have decayed for a period of greater than 95 hours0.0011 days <br />0.0264 hours <br />1.570767e-4 weeks <br />3.61475e-5 months <br />.
The staff computed the offsite doses for STP using the above assumptions and NRC computer code ACTC0DE. Control room operator doses were determined using the methodology in SRP Section 6.4.
The computed offsite doses and control room operator doses are well within the acceptance criteria given in SRP Section 15.7.4 and GDC 19. The assumptions used in calculating those doses and the resulting calculated values are provided in Attachments 1 and 2.
3.0
SUMMARY
l The proposed changes to the TSs will result in delayed containment closure in the event of a fuel handling accident. However, the staff has confirmed that the containment closure conditions and limitations established by the proposed TSs assure acceptable dose consequences. Accordingly, the licensee's proposal is acceptable.
1
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Texas State official was notified of the proposed issuance of the amendments.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements.
The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (59 FR 63123). Accordingly, the amendment meets the eligibility criteria for
. categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has 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, (2) such 1
activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Attachments:
- 1. Calculational Assumptions
- 2. Calculated Radiological Consequences Principal Contributors:
W. Long S. Boynton Date:
February 9, 1995 1
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CALCULATIONAL ASSUMPTIONS I
Parameter Value Source Term Variables Core Thermal Power (MWt) 3r,00 Total Number of Fuel Rods 60,952 Number of Damaged fuel Reds (1 assembly + 50 rods) 314 Power Peaking Factor 1.7 Time Since Shutdown 95 hours0.0011 days <br />0.0264 hours <br />1.570767e-4 weeks <br />3.61475e-5 months <br /> Release Fractien
- 7.odine 0.10 doble Gases 0.30 Pool Decontamination Factors *
!adine 100 Jeble Gases 1
i lodine forms
- Elemental 75%
Organic 25%
Fission Product Release Duration
- 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> e
Receptor P_giat Variables EXClysion ArqA_Bogndaty i
3 Atmospheric Relative Contontration, y/Q (sec/m )**
0-2 hours 1.3 x 10
Low Population Zone L
3 Atmospheric Relative Concentration, y/Q (sec/m )**
0-8 hours 1.6 x 10~5 8-24 hours 1.1 x 10~5 l-4 days 4.3 x 10
4-30 days 1.2 x 10
Control Room 3
Atmospheric Relative Concentration, y/Q (sec/m )***
0-8 hours 8.15 x 10
8-24 hours 4.92 x 10
l-4 days 1.56 x 10
4-30 days 8.72 x 10'5 3
Control Room Volume 200,000 ft
l Parameter Value Control Room (Continued) 3 Maximum Infiltration Rate 10 ft / min i
Pressurization Makeup Air Inflow 3
Flow Rate 2000 ft / min ESF Filter Efficiency Elemental Iodine 99%
Organic Iodine 99%
Particulate Iodine 99%
Recirculation Air Flow 3
Flow Rate 10,000 ft / min ESF Filter Efficiency Elemental Iodine 95%
Organic Iodine 95%
Particulate Iodine 95%
Geometry Factor 18 Iodine Protection Factor 380 F
Core Fission Product Inventories (TID-148d4_}.
Isotope Inventory (Ci/MWt)
Isotone Inventory (Ci/MWt)
I3' 2.51 x 10' Xe' 'm 2.60 x 10*
32 3
I 3.80 x 10' Xe' 3m 1.38 x 10 I'33 5.63 x 10' Xe' 3 5.62 x 10' 1'*35 6.57 x 10' Xe'35m 1.56 x 10' l
5.11 x 10' Xe'35 5.36 x 10' Xe'37 5.10 x 10' i
UN 4.15 x 10 Xe 4.78 x 10' 3
ssa u&
l.30 x 10' i
Kr5 2
4.10 x 10 ar Kr 2.34 x 10' Kr 3.20 x 10' as Kr '
3.98 x 10' 8
Note: Dose conversion factors from ICRP-30 were utilized for all calculations Reg Guide 1.25
- South Texas Project FSER (NUREG-0781, April 1986)
r CALCULATED RADIOLOGICAL CONSEQUENCES l
(rem)
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Exclusion Area Bpundary Q211 SRP 15.7.4 Limits Whole Body 0.2 6
Thyroid 36 75 i
Control Room Doerator Q2ig GDC-19 Limits l
Whole Body
<0.1 5
Thyroid 0.6 Equivalent to 5 rem whole body
- 1 Section 6.4 of the Standard Review Plan defines the dose limit to the thyroid as 30 rem.
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