ML20203K382

Revised Monthly Operating Rept for Oct 1997 for Hcgs,Unit 1

ML20203K382
Person / Time
Site:	Hope Creek
Issue date:	10/31/1997
From:	Ritzman R Public Service Enterprise Group
To:
Shared Package
ML20203K366	List: LR-N970798, Forwards MOR for Nov 1997 for HCGS & Revised MOR for Oct 1997 ML20203K375 ML20203K382
References
NUDOCS 9712220347
Download: ML20203K382 (4)

Text

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DOCKET NO,: 50-354

= UNIT: Hope Creek l DATE: 12/11/97 -

COMPLETED BY: R. Ritzman TELEPHONE: (609) 339-1445- ,

SUMMARY

OF CHANGES, TESTS, AND' EXPERIMENTS-FOR THE HOPE CREEK GENERATING STATION MONTH OCTOBER 1997 The following items completed during October 1997 have been evaluated to determine:

1. If the probability of occurrence or the' consequences of an accident or malfunction of equipment'important to safety previously-evaluated in the safety analysis report may be increased; or .

2. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report-may be created; or

3. If the margin of safety as defined in the basis for any 1 technical specification is reduced.

The 10CFR50.59 Safety Evaluations showed that these items did not create a new safety hazard to the plant nor did they affect the

-safe shutdown of the reactor. These items did not change the plant effluent releases and did not alter the existing environmental impact. .The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are. involved.

Desian-Chances Summary of Safety Evaluations 4EC-3579, Pkg. 8, High Pressure Coolant Injection Suppression Chamber Suction Valve Pressure Locking Mitigation. This design change installed'an external by-pass from'the bonnet cavity of a flex wedge gate valve to the Suppression Chamber side piping to prevent pressr.re locking.

The design change does not alter either the component'or the system function. This design change does not reduce the margin of' safety of the High Pressure Coolant Injection system and enhances the reliability of the system by providing an' external relief path for-fluid that may be trapped in the bonnet of the torus suction valve. This design change does not involve an unreviewed safety question.

9712220347 971215 PDR ADOCK 05000354-R- PDR

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.4EC-3644, Pkgs. 1-4, Standby Diesel Generator / Carbon Dioxide-system , Modification. This design change abandoned recirculation.

fan relays and-disconnected the leads frsi the_ carbon dioxide fire suppression' system to the Diesel Gentrator Recirculation

. RoomLfire_ dampers-and' fans. The design change disabled.the .

automatic closure of fire ~ dampers and fan shutdown-from-a-carbon  :

dioxide-actuation. .The fire damper vanes, spring closure ,

mechanismo and elector-thermal links were removed. _As a.part of

' this modification, each diesel fire area will be combined with its respective ventilation recirculation room to form-one fire area.

The design change removed an undesirable interaction between the Diesel Generator Recirculation system and the non-safety related-fire protection systems.

4HE-0075, Pkgs. I and 4x Class 1E, 125VDC Distribution Panel Cable Replacement. This design change replaced 500MCM feeder Ecables.with 350MCM cables in the "A" and "D" Class 1E 125VDC

Distribution Panel. The smaller cable will increase the circuit resistance and reduce fault current contribution. The associated increase in voltage drop is inconsequential and does not affect ,

the operation of any component.

The cable' replacement improves the reliability of the associated equipment. This design change does not create a different type, of accident or malfunction, and does not reduce the margin of safety as described in Technical Specifications. This design change does not involve an unreviewed safety question.

4HE-0076, Pkg. 1, 125 and 250 VDC Non-1E Battery Feeder Circuit.

This design change removed one of the parallel 500MCM cables from the battery feeder circuit of non-1E parallel batteries. This cable change increases the circuit resistance and reduces fault current contribution. The associated increase in voltage drop is inconsequential and does not affect the operation of any component.

.The cable deletion improves the reliability of the associated equipment. This dasign change-does not create a different type of accident or malfunction, and does not' reduce the margin of safety as described in Technical Specifications. This design  !

change does not involve an unreviewed safety question.

4HE-0122,-Pkg. 1, Additional Guardrails Around' Cask Loading Pit at Elevation 201' Reactor Building.--This design change added removable guardrails with-toe plates around the cask loading pit to augment the existing removable guardrail. The guardrail was

- added-as a personnel safety feature.

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Because.the guardrails are: lightweight, the consequences of

. dropping,a guardrail section into the Spent-. Fuel Pool is bounded-by t'ae fuel handling accident. This design change does not create a different type of accident or: malfunction, and does'not

. reduce the margin of safety as described in Technical-Specifications. This design change does not1 involve an

~ unreviewed safety question.

4HE-0300, Pkg. 28, Rework Closed Motor Operated Valve Position Indication on the Residual' Heat Removal Reactor Pressure Vessel Head Spray Isolation Inboard Valve.. This design change provided a jumper to bypass a torque switch during normal plant operations. The jumper ~is necessary to allow the resetting of an open indication limit switch. Resetting the limit switch provides position indication by maintaining the illumination of the valve open light until the valve reaches the full closed position.

The affected limit switch is not a control element and the

-elimination of the opening torque feature will not prevent the

- valve from performing its intended functions. This design change 4

does not create a different type of accident or malfunction, and does not reduce the margin of safety as described in Technical Specifications. This design change does not involve an unreviewed-safety question.

4HE-0300, Pkg. 29, Rework Closed Motor Operated Valve Position Indication on the Residual Heat Removal Reactor Pressure Vessel Head Spray Isolation Outboard Valve. This design change provided a jumper tt bypass a torgde switch during normal plant operations. The jumper is necessary to allow the resetting of an open indication limit switch. Resetting the limit switch provides position indication by maintaining the illumination of theLvalve open light until the valve reaches the full closed position.

The affected limit-switch is not a control element and the elimination of the opening torque feature-will not prevent the valve from performing its intended functions. This-design change does not create a different type of accident or malfunction, and does not reduce the margin of safety as described in Technical Specifications. This design change does not involve an unreviewed safety question.

Tomoorary Modifications Sn==arv of Sa(ety Evaluations 1

There were no changes in this category implementeg during October 1997.

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' Procidur_gs Summarv of Safety Evaluations HC'.IC-GP.SF-0001(Q), Revision:7, Reactor Manual Control System One-Rod-Out Interlock Bypass and Rod Position Indication System-Bypass to Reactor Manual Control System. This procedure revision Ladded steps to install and remove electrical ~ jumpers to bypass the. Refueling Platform "all rods full in" and "one rod out" interlock signals. These steps facilitate.the use of the platform for Reactor Pressure Vessel inspection activities.

The bypass is only_ permitted with licensed operator approval-and-with the reactor core de-fueled. The Refueling Platform control interlocks reinforce refueling procedures to prevent inadvertent criticality during fuel and control rod movements. The interlocks are not necessary when no fuel is in the reactor vessel. The interlock circuits do not interface with any other plant system. This procedure revision does not involve an Unreviewed Safety Question.

HC .MD-FR .KE-0 003 (Q) , Revision 16, Reactor Pressure Vessel Head Insulation-Package Removal. This procedure revision provided rigging instructions to relocate the Reactor Pressure Vessel insulation package from the existing primary storage location to the secondary storage location _on the refueling floor to allow for the reinstallation of the Reactor Pressure Vessel Cavity Shield Plug.

This procedure revision uses the same lifting devices and continues to meet the single failure proof guidelines. The new safe load path established by this procedure revision does not go over the Spent Fuel Pool or the Reactor Pressure Vessel cavity.

Therefore, this proposal does not increase the probability or consequences of the load drop accident or Polar Crane malfunction, does not create a different type of accident or malfunction, and does not reduce the margin of safety as described in Technical Specifications. This procedure revision does not involve-an Unreviewed Safety Question.

UFSAR Chance Notices Summary of Safety Evaluations There were no changes in this category implemented during October 1997.

Deficiency Reports Sn===rv of Safety-Evaluations There were no changes in this category implemented during October 1997.

Other Summarv of Safety Evaluation There were no changes in this category implemented during October 1997.