ML20206K162
| ML20206K162 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/07/1987 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20206K124 | List: |
| References | |
| NUDOCS 8704160288 | |
| Download: ML20206K162 (4) | |
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3.8.15 Deleted 3.8.16 Storage in the spent fuel pool shall be restricted to fuel assemblies having initial enrichment less than or equal to 4.1 w/o U-235.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
3.8.17 Storage in Region 2 (as shown on Figure 3.8.1) of the spent fuel pool shall be further restricted by burnup and enrichment limits specified in Figure 3.8.2.
In the event a checkerboard storage configuration is deemed necessary for a portion of Region 2, vacant spaces adjacent to the faces of any fuel assembly which does not meet the Region 2 burnup criteria (Non-Restricted) shall be physically blocked before any such fuel assembly may be placed in Region 2.
This will prevent inauvertent fuel assembly insertion into two adjacent storage locations.
The Irovisions of Specifications 3.0.3 and 3.0.4 are not applicable.
3.8.18 The boron concentration in the spent fuel pool shall be maintained (at all times) at greater than 1600 parts per million.
BASES Detailed written procedures will be available for use by refueling personnel.
These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9.6 of the FSAR incorporating built-in inter-locks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety.
If no change is being made in core geometry, one flux monitor is sufficient.
This permits maintenance on the instrumentation.
Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition.
The requirement that at least one decay heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel at the refueling temperature (normally 140 F), and (2) sufficient coolant circulation is main-tainedthroughthereactorcoretominimizgheeffectsofaborondilution incident and prevent boron stratification The requirement to have two decay heat removal loops operable when there is less than 23 feet of water above the core, ensures that a single failure of the operating decay heat removal loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling, thus in the event of a failure of the operating decay heat removal loop, adequate time is orcvided to initiate emergency procedures to cool the core.
TheshutdownmarginindicatedinSpecification3.8.4willkeeptgcore subcritical, even with all control rods withdrawn from the core Although the refueling boron concentration is sufficient to maintain the core k 5 0.99 if all the control rods were removed from the core, only a few cobol rods will be removed at any one time during fuel shuffling and Amendment No. 17, 56, 57, 76 59a 8704160288 870407 PDR ADOCK 05000313 P
replacement.
The k with all rods in the core and with refueling boron concentrationisaphximately0.9.
Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending unsafe condition detected from the main control board indicators during fuel movement.
The specification requiring testing reactor building purge termination is to verify that these components will function as required should a fuel handling accident occur which resulted in the release of significant fission products.
Because of physical dimensions of the fuel bridt it is physically impossible for fuel assemblies to be within 10 1 of each other while being handled.
Specification 3.8.11 is required as:
- 1) the safety analysis for the fuel handlingaccidentwasggjedontheassumptionthatthereactorhadbeen shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- and, 2) to assure that the maximum design heat load of the spent fuel pool cooling system will not be exceeded during a full core offload.
Specification 3.8.14 will assure that damage to fuel in the spent fuel pool will not be caused by dropping heavy objects onto the fuel.
Specifications 3.8.16 and 3.8.17 assure fuel enrichment and fuel burnup limits assumed in the spent fuel safety analyses will not be exceeded.
Specification 3.8.18 assures the boron concentration in the spent fuel pool will remain within the limits of the spent fuel pool accident and criticality analyses.
REFERENCES (1) FSAR, Section 9.5 (2) FSAR, Section 14.2.2.3 (3) FSAR, Section 14.2.2.3.3 Amendment No. 27, 56, 57, 76 59b
DESCRIPTION OF AMENDMENT REQUEST The proposed amendment request would revise AfD-1 Technical Specification.(TS) 3.8.15 and the related Basis to allow the Auxiliary Building crane to handle a spent fuel shipping cask.
TS 3.d.15 presently states that the spent fuel shipping cask shall not be carried by the Auxiliary Building crane pending the evaluation of the spent fuel cask drop accident and the crane design by AP&L and NRC review and approval.
The related Basis states that upon satisfactory completion of the NRC's review, Specification 3.8.15 shall be deleted.
TS 3.8.15 assures that the spent fuel cask drop accident cannot occur prior to completion of the NRC staff's review of this potential accident and the completion of any modifications that may be necessary to preclude the accident or mitigate the consequences.
NRC review of this particular issue was incorporated into the staff's resolution of the generic issue (A-36) related to control of heavy loads near spent fuel.
AP&L has completed all actions and submittals required by the issuance of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants", as evidenced by NRC Safety Evaluation-Report dated October 11, 1984 and Generic Letter 85-11 dated June 28, 1985.
Generic Letter 85-11 indicated that application for license amendment may be submitted to delete related license conditions citing GL 85-11 as the basis.
This amendment request is therefore submitted to delete TS 3.8.15 and its associated Basis.
BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The NRC has provided standards (10CFR50.92(c)) for determining whether a significant hazards consideration exists.
AP&L has evaluated the proposed TS amendment and, on the basis of these three standards as discussed below, determined that the proposed amendment does not represent a significant hazards consideration.
A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; AP&L's r cedures, load paths, crane equipnent certification and operator training and other related heavy load handling topics were evaluated as part of the c'ntrol of heavy loads issue and found acceptable.
Spent fuel cask handling is discussed in Section 9.6.2.6 of the ANO-1 SAR, which shows that the cask will never travel over spent fuel.
Although cask handling is presently prohibited by TS 3.8.15, ANO-1 SAR Section 9.6.2.6 further evaluates the incredible event of a cask drop accident and shows that the consequences are acceptable.
Deletion of TS 3.8.15 and the related. Basis to' allow handling of a spent fuel shipping cask by the Auxiliary Building crane will.have no actual impact on the cask drop or any other previously analyzed accident. This amendment request therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2) create the possibility of a new or different kind of accident from any accident previously evaluated; The proposed amendment request will allow handling of a spent fuel shipping cask by the Auxiliary Building crane where previously prohibited due to pending NRC review and approval of the control of heavy loads issue.
The cask handling methods and cask drop accident are discussed and evaluated in ANO-1 SAR Section 9.6.2.6.
The evaluation of the incredible event of a cask drop accident has included all possible equipment failures and shown the consequences to be within acceptable bounds.
No new accident scenarios can be identified related to the proposed amendment request, therefore this change is bounded by the current analysis in the SAR.
The proposed amendment request will therefore not create the possibility of a new or different kind of accident from any accident previously evaluated.
(3) involve a significant reduction in a margin of safety.
Although allowing use of the Auxiliary Building crane where previously prohibited could increase the possibility of a cask drop accident, the margin of safety is preserved in that the acceptable consequences of the cask drop accident evaluation in SAR Section 9.6.2.6 are not affected by this change.
The proposed amendment request will not adversely affect the adequacy and conservatism of the cask drop accident evaluation.
The spent fuel cask has been issued and continues to hold an NRC Certificate of Compliance for radioactive materials packages, and the procedures, load paths and equipment to be used for cask handling have been reviewed and approved by the NRC with the resolution of the control of heavy loads issue.
Therefore, the cask handling issue at ANO-1 continues to exibit an acceptable margin of safety and this amendment request will not involve a significant reduction in a margin of safety.
The NRC has also provided certain examples (51FR7750) of TS amendment requests which are likely to involve no significant hazards considerations.
This TS change request most closely matches example (iv):
A relief granted upor.
demonstration of acceptable operation from an operating restriction that was imposed because acceptable operation was not yet demonstrated.
In this case, demonstration of acceptable operation applies to the acceptable resolution of the NRC review of the control of heavy loads issue for ANO-1.
It should also be noted that a one-time exemption to existing TS 3.8.15 was allowed by the NRC to allow cask handling shipment of some test burnable poison assemblies during Cycle 3 operation at ANO-1 (TS Amendment 36, dated October 5, 1978).
Based on the above discussion and evaluation, AP&L has concluded that the proposed TS amendment request meets the standards for determining that no significant hazards consideration is involved and therefore concluded that this amendment application involves no significant hazards considerations.
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