ML20236P437

Application for Amend to License NPF-1,consisting of License Change Application 153,revising Tech Specs to Extend Surveillance Time Period for Verifying Control Rod Insertability from 24 H to 7 Days

ML20236P437
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	11/13/1987
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20236P430	List: ML20236P427 ML20236P437 ML20236P439
References
TAC-66741, NUDOCS 8711180005
Download: ML20236P437 (4)

Text

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PORTLAliD GENERAL ELECTRIC COMPANY EUGENE WATER & ELECTRIC BOARD AND PACIFIC POWER & LIGHT COMPANY l

l Operating License NPF-1 l Docket 50-344 1 License Change Application 153 i

ll This License Change Application requests modifications to Operating l License NPF-1 for the Trojan Nuclear Plant to extend the surveillance time l

period for verifying control rod insertability during control rod worth and shutdown margin tests.

PORTLAND GENERAL ELECTRIC COMPANY

-l By M )

field /

D. W. C Vice esident Nucicar l

Subscribed and sworn to before me this 13th day of November 1987. ,

Jhd Y Notary Pub'lic of Or/gon f L.)

My Conunission Expires: cNc-[ Mk l d j d

8711100005 871113 PDR ADDCK 05000344 P PDR i j

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e LCA 153 page 1 of 3 l

l Description of Change l Trojan Technical Specification (TTS) Surveillance Requirement 4.10.1.2 I requires that each full longth rod not fully inserted be demonstrated l operable by verifying its rod drop time to be no greater than 2.2 seconds within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the shutdown margin to less than the normal operating limits of TTS 3.1.1.1. This pecposed change would allow the surveillance to be performed within the previous 7 days instead of within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Reason for Change Implementing this technical specification change will improve the efficiency of reload physics startup testing by permitting the hot l condition rod drop time measurements (that precede the reload physics j testing program) to satisfy the requirements of TTS 4.10.1.2. A shutdown l and restart of the reactor will be avoided, as well as a time savings of I up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Also, the personnel safety hazard associated with pulling the rod movable gripper power fuses will be eliminated. l Similar changes have been reviewed and/or approved by the Nuclear Regulatory Commission (NRC) as cited in References 1 through 4.

Significant Hazards Consideration Determination The proposed technical specification change does not involve a signifi-cant hazards consideration because operation of the Trojan Nuclear Plant in accordance with this change:

1. Would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The previously analyzed accidents which could be affected by this proposed change are those which involve overcooling of the Reactor Coolant System (RCS). Because of the negative moderator temperature coefficient, RCS cooldown results in an increase in core reactivity.

Thus, a post-trip return to power could be experienced during events involving overcooling of the RCS if insufficient negativo reactivity is inserted by the control rods. As stated in the Bases for Specification 3/4.1.1.1,

" SHUTDOWN MARGIN requirements vary throughout core life. . . The most restrictive condition . . . is associated with postulated steam line break accident and resulting uncontrolled RCS cool-down. . . Accordingly, the SHUTDOWN KARGIN requirement is based upon this limiting condition and is consistent with FSAR accident analysis assumptions."

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• LCA 153 Page 2 of 3 t i

l Section 15.1.5.2 in the Trojan Final Safety Analysis Report (FSAR) analyzes the rupture of the main steam line with subsequent control l rod failure, j Since measurement of control rod worth inherently requires that shutdown margin be reduced, Surveillance Requirement 4.10.1.2 provides added assurance that an adequate amount of negative reactivity is available for insertion should a reactor trip occur.  !

Extending the surveillar.co time period for verifying control rod insertion capability could increase the probability of a control rod failure (ie, stuck control rod cluster). However, the impact on the probability of the previously analyzed accidents due to the increase in probability of a stuck control rod cluster is considered insig-  ;

nificant based upon the fact that the configuration of the components which are used in control rod cluster insertion will not change over the 7-day period. The components considered include the fuel assembly (including foreign material buildup in the gap between the i absorber rods and the guido plates within the guide tube assembly of the guide thimbles within the core region), the drive rod assembly, and the control rod drive mechanism. Also, since the control rod clusters will insert by gravity upon loss of power, the probability of a stuck control rod cluster is not increased due to an electrical l malfunction, if one were to occur-during rod worth testing.

i Although no quantitative probabilistic risk assessment (PRA) of this l proposed change has been performed for Trojan, a PRA performed by Baltimore Gas and Electric Company to support an amendment request to j make a comparable change to the Calvert Cliffs Unit 1 Technical j j Specifications determined that the probability of an overcooling l l

event with a stuck control element assembly increases insignificant 1y  !'

(from 1.1 E-7 to 4.8 E-7) when the requirement for trip verification is increased from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 7 days during low power testing.

(Reference 1) )

l Since the system design and installation, operating modes or safety system setpoints have not been changed, the consequences.of any previously unanalyzed accident will not be increased.

2. Would not create the possibility of a new or different kind of accident from any previously evaluated.

This proposed change does not create the possibility of a new or different kind of accident occurring, since the FSAR already assumes a hypothetical overcooling event combined with a stuck control rod cluster and since the proposed change does not result in any change to the facility.

3. Would not involve a significant reduction in the margin of safety.

( A significant reduction in the margin of safety will not result from the proposed change, since the affected TTS 3/4.10.1 provides that a minimum amount of control rod worth is immediately available when j tests are performed for rod worth measurement. i l

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• l LCA 153 Page 3 of 3 In the March 6, 1986 Federal Register (51 FR 7750-51), the NRC provided certain examples of amendments that are considered not likely to involve significant hazards considerations. Example (vi) is a change which either may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan, eg, a change resulting from the application of a previously used calculational model or design method.

As discussed above, this proposed change would result in an insignificant increase in the probability of a previously analyzed accident and does not affect the acceptance criteria of any safety-related system or l

component.

1 Therefore, based on the above discussion, this change has been determined not to involve a significant hazards consideration.

References

1. 50 FR 15061 (April 19, 1985), Baltimore Cas and Electric Company; Consideration of Issuance of Amendment to Facility Operating License and Proposed No Significant Hazards Consideration Determination.

l 2. U.S. NRC, " Safety Evaluation by the Nuclear Reactor Rogulation Supporting Amendment No. 40 to NPF-10 and Amendment No. 29 to NPF-15, l Southern California Edison Company, et al. , San Onofre Nuclear l Generating Station, Units 2 and 3", December 12, 1985.

l l 3. 51 FR 22584 (June 20, 1986), Arkansas Power and Light Company, l Consideration of Issuance of Amendment to Facility Operating License l and Proposed No Significant Hazards Consideration Determination.

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4. 51 FR 37514 (October 22, 1986), Louisiana Power and Light Company; consideration of Issuanco of Amendment to Facilit'y Operating License and Proposed No Significant Hazards Consideration Determination.

Safety / Environmental Evaluation Safety and environmental evaluations were performed as required by 10 CFR 50 and the Trojan Technical Specifications. This review deter-mined that an unreviewed safety question does not exist since Plant operations remain consistent with the Updated FSAR, adequate surveillance is rcaintained, and there is no significant adverse impact upon the environment.

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l ARA /kal 5938k.1187 w __ _ _ _