ML20215M118
| ML20215M118 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/22/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215M107 | List: |
| References | |
| TAC-63003, NUDOCS 8610300039 | |
| Download: ML20215M118 (6) | |
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UNITED STATES
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 22 TO FACILITY OPERATING LICENSE N0. NPF-29 MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET N0. 50-416
1.0 INTRODUCTION
Sy application dated October 3, 1986, Mississippi Power & Light Company (the licensee) requested an amendment to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1.
The proposed amendment would make a temporary change to Technical Specifications 3.3.2,
" Isolation Actuation Instrumentation," 3.6.4, " Containment and Drywell Isolation Valves," 3.6.6.1, " Secondary Containment Integrity," 3.6.6.2,
" Secondary Containment Automatic Isolation Dampers / Valves," and 3.6.6.3,
" Standby Gas Treatment System" by adding footnotes, which state that secondary containment integrity is not required for control rod movement in a defueled control cell for the time interval of October 3 through October 10, 1986.
2.0 EVALUATION The proposed change to the Technical Specifications (TSs) would provide temporary relief (October 3 through October 10) during the present refueling outage from the requirements for maintaining secondary containment integrity while removing control rods from defueled control cells or while reinstalling the control rods into the defueled control cells. A control cell in the core consists of a control rod with four adjacent fuel assemblies. The removal of control rods from the core and reinstallation of control rods into the core would be performed in accordance with TS 3.9.10.2, " Multiple Control Rod Removal." Such movements of control rods are defined as core alterations by TS 1.7.
TSs 3.3.2, 3.6.4, 3.6.6.1, 3.6.6.2, and 3.6.6.3 require that when the equipment specified therein is inoperable, core alterations must be suspended.
In order to keep the present schedule for the refueling outage, activities requiring breaching of the secondary containment boundary and inoperability of the standby gas treatment system need to be continued while control rods are being removed. The need for the change to the Technical Specifications on an emergency basis is evaluated in Section 3.0 of this safety evaluation.
8610300039 e61022 PDR ADOCK 05000416 P
- The licensee's safety analysis provided in its October 3,1986 application considered reactivity changes involved in such control rod movements and the accidental dropping of a control rod assembly during its transportation out of the reactor, into the upper containment pool and possibly into the spent fuel pool. The NRC staff has reviewed the licensee's application.
Our discussion and conclusions regarding reactivity changes and accidents is given below.
The movement of control rods within the restrictions provided by Technical Specification 3.9.10.2 does not present a reactivity problem. Previous generic analyses reviewed and accepted by the staff have indicated that the removal of the four fuel assemblies in a control cell and subsequent removal of the control rod results in a reactivity decrease both locally (for the control cell) and globally (for the surrounding region and the reactor as a whole). This is the basis for Specification 3.9.10.2 in both Grand Gulf and the BWR Standard Technical Specifications. Thus with the four fuel assemblies already removed from a given control cell there will be no reactivity problem with the removal or withdrawal of the control rod from that control cell. The shutdown margin of the reactor will, if anything, increase following the removal of the fuel assemblies and with the control rod either inserted or withdrawn. Accordingly, the staff concludes that reactivity changes associated with the proposed change are acceptable.
The licensee has provided the results of an analysis of accidentally dropping a control rod assembly on a fuel assembly, while it is being moved out of the reactor. The reactor is presently flooded for refueling activities, so that the movement occurs under water. The control rod weichs about 218
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pounds which is light compared to a fuel assembly that weighs about 1100 pounds. The fuel assemblies remaining in the core are held in position by the core top guide so 'that the fuel assembly lifting bail is the part of the fuel assembly that would be struck by a dropped control rod. The control rod blade contains neutron absorber material enclosed in small diameter tubing. A row of these tubes is enclosed in steel sheaths to form the control rod blade.
Four blades are joined in a crucifom to form the control rod. The licensee assumed, conservatively, that a control rod which drops on the bail of a fuel assembly would damage cladding on all fuel rods in that assembly. This is conservative, as compared to a dropped fuel assembly accident where all rods in the fuel assembly are also assumed to be damaged, because the control rod is lighter than a fuel assembly and the fuel assembly is assumed to be dropped in air rather than water.
In addition to assuming a conservatively large number of fuel rods would be damaged in a dropped control rod accident, the licensee has committed to take compensatory measures during the proposed temporary period when control rods would be handled without secondary containment integrity. Appropriate means will be available to provide temporary closure or sealing of those penetrations and areas where secondary containment is breached. These-closure measures will not provide design level leak tightness but will provide effective closure to minimize air flow and will be executed in the
. event that closure of secondary containment would be needed to mitigate the release of radioactive material. While in the process of removing or re-installing control rods, administrative controls will be used to minimize i
the height of the control rods when they are moved over irradiated fuel assemblies.
The staff performed an independent control rod handling accident analysis for the condition of no secondary containment integrity and assuming that i
the standby gas treatment system (STGS) filters are not available. The exclusion area boundary doses are well within the 10 CFR Part 100 dose limits as defined in the Standard Review Plan (SRP) Section 15.74. The control room operator doses from this accident are within the SRP Section 6.4 guideline values. Accordingly, the staff concludes that the proposed handling of control rods without secondary containment is acceptable for the proposed time interval.
3.0 EMERGENCY CIRCUMSTANCES The licensee's October 3, 1986 application included a discussion and technical bases for the proposed change. The plant is now in a refueling outage which began September 5, 1986, and is scheduled to end November 8, 1986. One of the activities planned for this outage is the removal from the reactor pressure vessel of instrumentation used in the core internals vibration monitoring program. During the removal, certain portions of the instrumentation hardware became loose inside the reactor vessel.
Plans to inspect and retrieve loose hardware involve the removal and rein-stallation of multiple control rods as provided for by Specification 3.9.10.2.
The four fuel assemblies surrounding each of 14 control rods have been removed to facilitate inspection. The movement of a control rod in a defueled core cell is a core alteration as defined in Specification 1.7.
Other outage activities involving the breaching of secondary containment integrity are important to the overall outage schedule and are constrained 4
from execution while core alterations are in progress.
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With the issuance of the reouested license amendment changing the GGNS Technical Specifications, the secondary containment work and control rod removal, inspection, and installation work can proceed in parallel, minimizing impact to the outage schedule. Without the requested change, rod removal and installation would have to be completed prior to the 1
commencement of secondary containment work, thus extending the overall j
outage by approximately six days, i
The need for this change was not identified until the discovery on September 28, 1986 of loose vibration monitoring hardware in the pressure vessel. All planned core alterations for this phase of the outage had been otherwise completed. The need for control rod removal for the purpose of inspection and retrieval of loose hardware was not anticipated.
Therefore, the need for prompt action to obtain this temporary amendment to the Technical Specifications could not have been avoided.
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. The staff has reviewed the circumstances associated with the licensee's request and agrees that performing the inspection and retrieval of loose hardware from the reactor vessel in series instead of in parallel with activities involving the secondary containment, could extend the refueling outage. The requested amendment is, therefore, needed to avoid a delay in the scheduled restart of Grand Gulf Nuclear Station, Unit 1, and thus constitutes a valid emergency situation. The staff has also concluded that the licensee has provided a sufficient basis for finding that the emergency situation could have not been avoided by prior application.
Therefore, in accordance with 10 CFR 50.91(a)(5), a valid emergency existed.
3.1 FINAL NO SIGNIFICANT HA7ARDS CONSIDERATION DETERMINATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amend-ment to an operating license for a facility involves no significant hazards considerations if operation of the facility in accordance with a proposed amendment would not:
(1) Involve a significant increase in the probability or consecuences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in.a margin of safety.
The licensee has provided an analysis of significant hazards considerations in its reouest for a license amendment. The licensee has concluded, with appropriate bases, that the proposed amendment meets the three standards in 10 CFR 50.92 and, therefore, involves no significant hazards considerations.
The NRC staff's evaluation of the licensee's submittal (Section 2.0 above) provides the basis for staff's conclusions regarding these three standards.
A summary of staff's conclusions and bases follows.
The proposed changes to the Technical Specifications would allow movement of the control rods out of the reactor and reinstallation into the core without secondary containment integrity and without the standby gas treat-ment system operable. The relevant accident to consider for these changes is damage to irradiated fuel rods and release of fuel rod gap activity as a result of a dropped control rod. The previously analyzed accident was a dropped fuel assembly accident, assuming filtration of gaseous activity released from damaged fuel rods through the standby gas treatment system.
The analysis of a dropped control rod without filtration in the standby gas treatment system yields slightly higher doses to control room operators and higher offsite doses than the previously evaluated accident. The lack of filtration is nearly offset by the long decay time of irradiated fuel in the present core (about 30 days) compared to that assumed in a dropped fuel assembly accident (immediately after shutdown). However, the doses are still within the acceptance criteria in the Standard Review Plan and the Comission's rules in 10 CFR Part 50, Appendix A, General Design Criterion 19 and 10 CFR Part 100.
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. i The changes to the Technical Specifications do not involve a significant increase in the probability of an accident previously evaluated because planned movement of control rods during the refueling outage has not
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_ changed. Although the consequences of an accident are greater than the consequences of a previously evaluated accident (i.e. damaged fuel rods),
they are not significantly greater because calculated doses are within applicable regulatory acceptance criteria. The changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because dropping of loads on fuel assemblies during refueling activities has been previously evaluated. The proposed changes do not involve a significant reduction in a margin of safety because although work will be in progress which breaches secondary containment, administrative procedures will be in effect to make temporary closures of these breaches should an accident occur and accident doses are within regulatory acceptance 1
criteria, even assuming no secondary containment.
Accordingly, the proposed amendment does not involve significant hazards considerations.
3.2 STATE CONSULTATION
The staff consulted with the State of Mississippi by telephone on October 3, 1986. The state expressed no concern either from the stand-point of safety or of the determination of no significant hazards considerations.
4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the i
installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the i
amendment involves no significant increase in the amounts, and no signifi-cant change in the types, of any effluents that may be released offsite and that there is no significant increase in indivioual or cumulative occupational radiation exposure. The Commission has made a final finding that this amendment involves no significant hazards consideration. Accord-ingly, this amendment meets the eligibility criteria for categorical j
exclusion set forth in 10 CFR 51.22(c)(9). Pursuantto10CFR51.22(b),no environmental impact statement or environmental assessment need be prepared i
in connection with the issuance of this amendment.
5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above that:
(1) the amendment does not (a) significantly increase the probability or l
consequences of an accident previously evaluated, (b) create the possibility of a new or different kind of accident from any previously evaluated or (c) significantly reduce a safety margin and, therefore, the amendment does not involve significant hazards considerations; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (3) such activities will be i
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6-conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and the security or to the health and safety of the public.
Principal Contributors:
Angela Chu, Plant Systems Branch, DBL H. Richings, Reactor Systems Branch, DBL L. Kintner, BWR Profect Directorate, DBL s
Dated: October 22, 1986
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