ML20080P317
| ML20080P317 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 02/28/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20080P306 | List: |
| References | |
| NUDOCS 9503070316 | |
| Download: ML20080P317 (8) | |
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UNITED STATES 1
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NUCLEAR REGULATORY COMMISSION j
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ANENDNENT N0.117TO FACILITY OPERATING LICENSE NO. NPF-10
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AND ANENDNENT N0.106 TO FACILITY OPERATING LICENSE NO. NPF-15 SOUTHERN CALIFORNIA EDIS0N CONPANY j
SAN DIEGO GAS AND ELECTRIC CONPANY a
THE CITY OF RIVERSIDE. CALIFORNIA THE CITY OF ANAHEIN. CALIFORNIA l
SAN ONOFRE NUCLEAR GENERATING STATION. UNITS 2 AND 3
-l DOCKET NOS. 50-361 AND 50-362
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1.0 INTRODUCTION
s By letter dated Decen,er 30, 1993, as supplemented by letters dated June 3,-
i 1994, August 25 1994, January 3 and 19, 1995, Southern California Edison Company, et al. (SCE or the licensee) submitted a request for changes to the Technical Specifications (TS) for San Onofre Nuclear Generating Station (SONGS), Unit Nos. 2 and 3.
These submittals contain the licensee's 1
justification to replace the current TS with a set of TS based on the CE Owners Group Improved Standard Technical Specifications (STS)-issued by the i
NRC Staff as NUREG-1432 in September 1992. The adoption of Owners Group approved TS is part of an industry-wide initiative to standardize and improve
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TS.
SONGS Units 2 and 3 are the lead plants for adoption of the CE Owners 1
Group standardized TS.
l This Safety Evaluation (SE) addresses the licensee's request to implement one j
provision of their new TS program in advance of the implementation of the r
entire program. This provision is a revision to TS 3.9.4, " Containment j
Building Penetrations," and the associated bases to allow both doors of the i
containment personnel airlock to be open at the same time during refueling i
operations provided certain conditions are met. By letter dated January 30, 1995, the licensee forwarded TS pages in the current format for the change l
evaluated by this SE. Based on discussions with the licensee, the TS pages c
i provided in the January 30, 1995, submittal were slightly rearranged for i
better readability.
1 The initial notice in the Federal Reaister included the letters dated
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December 30, 1993, June 3, 1994, and August 25, 1994. The additional information contained in the January 3,19, and 30,1995, letters were
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clarifying in nature, within the scope of the initial notice and did not j
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5 affect the NRC staff's proposed no significant hazards consideration determination.
2.0 BACKGROLNG The containment serves to control fission product radioactivity that'may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR-
.Part 100. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment' atmosphere following accident conditions.
With the primary coolant system above 200*F and pressurized, the containment itself may become pressurized during an accident. Therefore, the containment and its penetrations must be operable by being capable of withstanding the pressure and thus prevent excessive fission product release to the environment.
During refueling operations, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The required condition is referred to as " containment closure" rather than " containment operability." Containment closure means that all potential escape paths are closed or capable of being closed. The closure restrictions are sufficient to restrict fission product radioactivity release from containment due to a fuel handling accident (FHA) inside containment during refueling.
The containment personnel airlock, which is also part of the containment pressure boundary, provides a means for personnel access. The personnel airlock has a door at both ends which are normally interlocked to prevent simultaneous opening when." containment operability" is required. During periods of unit shutdown when neither " containment operability" or
" containment closure" are required, the door interlock mechanism may be disabled, allowing both doors of the airlock to remain open for extended periods when frequent containment entry is necessary.
During core alterations, the current TS require that one personr.el airlock door be closed at all times. This means that personnel must enter one door with the other shut, shut the door just passed, then open the other door.
This operation occurs many times a day during refueling outages and the excessive cycling of the doors results in the need for frequent maintenance of the door hinge pin, the door seals, the packing of the equalizing valve, and other components.
Other licensees have experienced similar difficulties. Calvert Cliffs Nuclear Power Plant submitted an amendment request dated November 5, 1993, which would, in part, allow the personnel airlock doors to remain open during core alterations. The amendment war issued on August 31, 1994, based primarily on the fact that calculated offsite dose and control room operator doses are within acceptable limits with the doors open following an FHA.
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-3.0' EVALUATION l
.The licensee proposes to modify TS 3.9.4 to allow both doors of the' 1
containment personnel airlock to be open at the same time during refueling.
operations provided certain conditions are met. These conditions are:
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one personnel airlock door is OPERABLE, and l.
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the plant is in MODE 6 with 23 feet of water above the fuel in the
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reactor vessel, or j
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defueled configuration with fuel in containment (i.e., fuel in refueling machine or upender).
Operability of the personnel airlock door is defined in the TS bases, and it i
requires that the door is capable of being closed; that the door is unblocked and no cables or hoses are being run through the airlock; and that a designated individual is continuously available to close the airlock door.
t This individual must be stationed at the outer airlock door.
Standard requirements regarding airlock penetration integrity during core alterations: The applicable staff positions regarding opening.of airlock-doors during Mode 6 (Refueling Operations) are stated in Section 3.9.3 (BASES) of the Improved Standard Technical Specifications (NUREG-1432, " Standard l
Technical Specifications for Combustion Engineering Plants" or "ISTS").
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excerpted from the ISTS states The containment airlocks, which are part of the containment pressure boundary, provide a means for f
personnel access during MODES 1, 2, 3, and '4 operation. During periods of shutdown when.
containment closure is not required, the door t
interlock mechanism may be disabled, allowing both doors of an airlock to remain open for extended periods when frequent containment entry is necessary.
During CORE ALTERATIONS or movement of irradiated fuel i
assemblies within containment, containment closure is required; therefore, the door interlock mechanism may remain disabled, but one airlock door must always remain closed.
l The requirements on containment penetration closure ensure that a release of fission product radioactivity within containment will be restricted from escaping to I
the environment. The closure restrictions are sufficient to restrict fission product radioactivity l
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
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v radiological consequenc'es resultifrom a fuel h'andling accident. The fuel handling accident-is a postulated-event-that involves damage to irradiated fuel..- Fuel j
handling accidents include dropping single :
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> irradiated fuel' assembly.and hand 1"ng too or a heavy l
object onto other irradiated fuel: assemblies. The minimum decay time of.[72] hours prior.to CORE l
ALTERATIONS ensure [s].that the release of fission j
product radioactiv'ity subsequent to a fuel handling :
9 accident, results in doses that are well within the-1 guideline values specified in 10 CFR Part 100. The
.j acceptance limits for offsite radiation exposure are' i
contained in Standard Review Plan Section 15.7.4,-
i Rev. 1, which defines "well within" 10 CFR Part 100 to j
be 25% or less of the 10 CFR Part 100 values.
-l As noted above, the basis for the staff position against simultaneous opening of both airlock doors during core alterations is to limit fission product j
leakage in.the event of.a Fuel Handling Accident.
In performing analyses-of-the radiological consequences of a Fuel Handling Accident, the containment isolation criteria of Standard Review Plan (SRP) Section 15.7.4 are used.
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fuel handling is prohibited when the containment is open, radiological consequences need not be calculated.
If the containment will be open during fuel handling operations, automatic isolation by radiation detection j
instrumentation must be provided for penetrations and calculations must' l
. demonstrate acceptable consequences. However, automatic isolation of airlock.
1 doors is not practicable.
Standard Technical Specifications thus specify that 1
airlock integrity be maintained.during fuel hand 99 in containment. However, the licensee has shown by analysis that the StauW Technical Specification-i requirement need not be applied to San Onofre Uniti 4 and 3.
San Onofre 2/3 Fuel Handling Accident Analysis: The licensee performed an analysis of a fuel handling accident with the personnel airlock doors open.
l In performing the analysis the licensee used the assumptions and methodology prescribed by Regulatory Guide (RG) 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-i 72-hour decay time was assumed in the analysis. The licensee's analysis demonstrated that the 0-2 hour site boundary thyroid dose would be less than-i 72.7 Rem and the 0-2 hour whole body site. boundary dose would be less than 0.3 Rem. These consequences are within the SRP limits of 75' Rem thyroid and 6 Rem Whole Body (WB).
Control Room Habitability considerations: General Design Criteria (GDC) 19-specifies that adequate radiation ~ protection is to be provided to permit-access and occupancy of the control room under accident conditions without personnel exposures in excess of 5 Rem WB or its equivalent to any part of the body for the duration of the accident. The SRP limits are 5 Rem W8 and 30 Rem Thyroid for the control room operator. The licensee's analysis demonstrated that the control room operator whole body dose would be less than 0.2 Rem and the thyroid dose would be less than 19.5 Res.
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Staff Confirmatory Analysis: The staff performed an independent analysis to confirm the licensee's results. The staff evaluated the radiological consequences resulting from a postulated fuel handling accident using accident source terms given in RG 1.4, " Assumptions Used for the Potential Radiological Consecuences of a Loss of Coolant Accident for Pressurized Water Reactors,"
assumptions contained in RG 1.25, and the review procedures specified in SRP Section 15.7.5.
The staff assumed an instantaneous puff release of noble gases and radiciodine from the gap of the broken fuel rods as gas bubbles pass up through the 23 feet of water covering the fuel. All airborne radioactivity l
reaching the containment atmosphere is assumed to be exhausted within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l
into the enviror. ment. All radioactive material in the fuel rod gap is assumed to have decayed for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The staff computed the offsite doses for the San Onofre exclusion area boundary using the above assumptions and the NRC computer code, ACTICODE. The control room operator doses were computed using methodology given in SRP Section 6.4.
The computed offsite doses are within the acceptance criteria given in SRP 15.7.5 and the computed control room operator doses meet the dose limits set forth in GDC 19. The resulting values of the offsite and control room operator doses calculated by the staff are listed in Table 1 and the assumptions used are listed in Table 2.
Based on the evaluation performed by the licensee and the staff's confirmatory analysis, the staff concludes that the radiological consequences of not having a closed containment following a fuel handling accident are acceptable. The steps taken by the licensee to optimize the ability of plant personnel to close the personnel airlock, if needed, provide assurance that offsite radiological consequences will be minimized to the extent practical. Based on t
these reasons, the staff approves the proposed changes to TS 3.9.4.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
1 The amendments change a requirement with respect to the 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 considera-tion, and there has been no public comment on such finding (59 FR 49434).
Accordingly, the amendments meet 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.
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6.0 CONCLUSION
i The Commission has concluded, based on the considerations discussed above, t
t!iat (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 activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public, t
Principal Contributors: John L. Minns t
Mel B. Fields Attachments:
1.
Calculated Radiolgical Consequences, Table 1 2.
Assumptions used for Calculating Radiological Consequences, Table 2 Date:
Febric? :c 28, 1995
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t Table 1 CALCULATED RADIOLOGICAL CONSEQUENCES
- Exclusion Area Boundary Dgit SRP Limits.
Whole Body 0.3 Rem-6 Rem Thyroid-75 Rem 75 Rem-Control Room Doerator 0g11 GDC-19 Limits i
'i Thyroid 30 Rem Equivalent to 5 Rem WB*
- Section 6.4 of the SRP defines the dose limit to the thyroid as 30 Rom.
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. TABLE 2 l
ASSIBIPTIONS USED FOR CALCULATING RADIOLOGICAL CONSE00ENCES l
Parameters Guantity Power Level, Mwt 3390 Number of Fuel Rods Damaged 236 Shutdown Time, hours 72 Power Peaking Factor 1.65 t
Fission-Product Release Fractions, %
Iodine 10 i
Noble gases 30 Pool Decontamination Factors Iodine 100 Noble gases 1
Iodine Forms, %
Elemental 75 Organic 25 3
Atmospheric Re'4ative Concentration, sec/m 2.72E-4 Fission-Product Release Duration, hours 2
Dose Conversion Factors ICRP-30 CONTROL ROOM 3
Atmospheric Relative Concentration, sec/m 3.lE-3 Filter Recirculation Rate, cfm 2.97E+4 Unfiltered Inleakage, cfm 10 Filter Efficiency, %
95 Iodine Protection Factor 20.4 Geometry Factor 18 Cor. trol Room Volume, cubic feet 2.44E+5 i
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