IR 05000277/1987031

Insp Repts 50-277/87-31 & 50-278/87-31 on 871102-06.No Violations or Deviations Noted.Major Areas Inspected:Nrc Bulletin 84-03 & IE Info Notice 84-93 Accessible Portions of Units 2 & 3

ML20239A719
Person / Time
Site:	Peach Bottom
Issue date:	12/09/1987
From:	Blumberg N, Rebelowski T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20239A718	List: IR 05000277/1987031 ML20239A717
References
50-277-87-31, 50-278-87-31, IEB-84-03, IEB-84-3, IEIN-84-93, NUDOCS 8712300045
Download: ML20239A719 (7)

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. 4 U.S. NUCLEAR REGULATORY COMMISSION

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50-277/87-31 Report N /87-31 DPR-44 Docket N DPR-56 i DPR-44 License N DPR-56 Licensee: Philadelphia E'ectric Company 2301 Market Street Philadelphia, Pennsylvania 19101 Facility Name: Peach Bottom Atomic Power Station Units 2 and 3  ;

Inspection At: Delta, Pennsylvania Inspection Conducted: November 2-6, 1987 .)

Inspectors: M # 4 Theodore Rebelowski, Senior Reactor Engineer

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Approved by: d/4 4g e M [ ( / / 7

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Nb'rmgfi J. Blumberg, Chief Operational Programs Section, 0B, DRS /datf Inspection Summary: Routine Unannounced Combined inspections on November 2-6, 1987 (Report Nos. 50-277/87-31 and 50-278/87-31)

Areas Inspected: Inspection of NRC Bulletin No. 84-03 and IE Information Notice 84-93 of accessible portions of Unit 2 and Results: No violations or deviations were identifie /

8712300045 871217 f PDR ADOCK 05000277 0 .DCD _ _ _ _

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DETAILS )

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1.0 Persons Contacted Peach Bottom Atomic Power Station Mr. H. R. Abendroth, Independent Safety Engineering Group Mr. D. A. Anders, Engineer Mr. C. H. Anderson, Asst. Supt. Instrumentation and Control Mr. J. T. Budzynski, Reactor Engineer Mr. P. L. Bushek, Independent Safety Engineering Group

• Mr. J. B. Cotton, Superintendent Plant Services

• Mr. A. A. Fulvio, Technical Engineer Mr. S. J. Mannix, Shift Manager Mr. M. Mickley, Technical Assistant j Mr. M. S. Sattler, Engineer 4 Mr. W. Sawruk, Structural Engineering

• Mr. D. M. Smith, Manager, Peach Bottom Atomic Power Station US Nuclear Regulatory Commission

• Mr. T. P. Johnson, Senior Resident Inspector, Peach Bottom Atomic Power  !

Station {

• Mr. R. J. Urban, Resident Inspector, Peach Bottom Atomic Power Station

• Mr. L. E. Myers, Resident Inspector, Peach Bottom Atomic Power Station Other licensee and contractor employees were also contacte l

• Present at exit interview on site and for summation of preliminary-finding .0 Refueling Cavity Water Seals (Inspection and Enforcement Bu11etin84-03)

Background j At another facility, during a plant refueling operation, the refueling cavity water seal failed allowing the cavity water level to be reduced to the level of the reactor vessel head flange. I&E Bulletin 84-03 titled " Reactor Water Cavity Seals" addressed this concern to all licensees. A review of the Peach Bottom response (9-17-84) to this event was the subject of this inspectio .1 Initial Review and Bellows Seal Description The licensee does not utilize the pneumatic seals in the refueling cavity seal design. Stainless steel bellows are used for both-refueling and drywell seal The bellows as described in licensee's response and the Peach Bottom Atomic Plant Station Manual follows:

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The refueling bellows forms a seal between the reactor vessel and the drywell to permit flooding of the space (reactor well) above the vessel during refueling operations. The refueling bellows assembly consists of a type 304 stainless steel bellows, a backing plate, a l spring seal, and a removable guard ring. The backing plate surrounds the outer circumference of the bellows to protect it, and is equipped to monitor seal leakage. The self-energizing spring seal is located in the area between the bellows and the backing plate, and is designed to limit water loss in the event of a bellows rupture by yielding to make a tight fit to the backing plate when subjected to full hydrostatic pressure. The guard ring attaches to the assembly and protects the inner circumference of the bellow The assembly is welded to the reactor bellows support skirt and the reactor well seal bulkhead plate. The reactor bellows support skirt is welded to the reactor vessel shell flange, and the reactor well seal bulkhead plate bridges the distance to the drywell wall. The bulkhead plate is penetrated by six ventilation ducts. These penetrations are sealed by six watertight, hinged covers during refuelin .2 Analysis of Postulated Event The licensee's analysis shows that a postulated drywell bellows seal failure would result in the greatest refueling cavity water volume displacement and could result in a maximum of a 3300 gpm leak rate from the refueling cavity. With an assumption that a fuel assembly was being moved without makeup water, the fuel assembly in movement would uncover in forty-two minutes. The licensee's review of the

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movement of a fuel assembly indicates that a fuel assembly can be

! , taken from the core and moved to storage in spent fuel pool in less than 10 minutes. Adding ten minutes for operator recognition of a leakage event, the fuel assembly could be in a safe position in twenty minutes. Thus the fuel assembly would not be uncovered. The inspector reviewed the calculations and found them to be acceptabl .3 Makeup Water Capabilities The licensees analysis shows demineralized water makeup to the re-fueling cavity exist at 1885 gpm. Thus, upon operator recognition of a leak, this make up capability could be utilized to reduce the refueling cavity level drop rat Based on the licensee's analysis of the drywell bellows seal failure which is postulated as the greatest refueling cavity water volume displacement, indicates that a 3300 gpm leak rate would occur. Thus, the loss 1415 gpm (3300 - 1885 gpm) of inventory to uncover' fuel assembly would be 88 minutes. Since this 88 minutes exceeds the time of 20 minutes for placement of a fuel assembly in safe position, the makeup systems are adequate. An additional makeup of RHR is avail-able after one hour and would be 10,000 gpm thus, level loss in cavity would not occu These calculations were reviewed and found acceptable, w__ -- _ _ - _ _ _ _ - ._- -

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During the postulated loss of drywell bello'ws seal, sufficient time -

exists to safely store a fuel assembly that was in transit without 4 uncovering fue .4 Licensca Response Review Based on the Bulletin 84-03 specificity the licensee response was~ l directed to the creditability of a bellows failure. Based on the inspector's review of diagrams and observations on both refueling floors, bellows catastrophic failure is not a creditable even .5 Instrument Review Monitoring Seal Bellow Leakage Two flowmeters per unit monitor bellow seal leakage. A review of calibration data for Unit 2 flowswitches (FS) 2692 and 2704 and Unit 3 FS 3692 and 3704 identified that an electrical continuity check is performed, which tests the local and control room alarm The flow switch is a Ball Co. Serial D Model 2500S which is actuated by the mechanical movement of a paddle in the flow stream, which is j not presently examined or tested. The licensee stated they will-

. examine internals of all four flow switches prior to heat up to

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verify the paddle-mechanism does move the magnetic slug and actuates the bellow leakage alarms. Based on this examination a frequency of surveillance will be determine The inspector has no further questions on this ite Additional Surveillance / Calibration review of refueling cavity level calibrations and area radiation alarms has been conducted. The licensee does have the means to recognize level changes during refuel-ing operations. No deficiencies.were identifie .6 Control Room Procedures That Address Seal Failure The control room has multiple alarms that are activated from seal rupture drain high flow alarm (1(A-1)) flow switches FS 2704 and 3704 and reactor vessel bellows seal leak alarm (2(2A)) flow switches FS 2692 and FS 3692. The operator alarm actions do not address specific corrective actions to mitigate these events. The licensee plans action to address operator guidance to institute makeup water and address any fuel assembly. movements that would require fuel assembly safe storage during a leakage even .7 Bellows Seal Inspection No program presently exists to monitor degradation of the bellow seals and their supports. The examination of the bellow seals l'

require the removal of a portion of ring guards. The licensee will review the need and ability to remove the ring guard (bolted) and perform and document the bellow seal conditio i

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2.8 Summary of Findings (Bulletin No. 84-03)

The inspector has identified no significant leakage paths past the bellow seals other than those identified by the ' license The licensee has stated that to enhance his surveillance program, the will add, additional surveillance / calibration inspections to verify that the flow switch paddle activators are mechanically free and are r able to indicate leakage flow. The licensee is also reviewing the frequency of removal of part of bellows guards prior to a refueling outage to allow for visual examination of the bellows. Alarm responses to bellow seal leakage will be revised to present the operator with guidance to institute makeup water on a bellow seal failur The inspector has determined that the licensee's Bulletin No. 84-03 i response has properly addressed NRC concerns. This item is close l IE Information Notice (IEIN} 84-93 j As additional information was obtained from PWR and BWR licensees in re-sponse to IE Bulletin 84-03, additional areas of concern were identified in IEIN 84-93, " Potential For Loss Of Water From The Refueling Cavity" that could potentially reduce reactor refueling cavity water inventor were identified, for example,'in the use of pneumatic seals'and at the penetrations for main steam lines. The licensee addressed with their internal reviews the main steam lines and concluded that the plugs used to block steam line penetrations in the vessel were not a creditable inventory loss pat The inspector reviewed procedures that addressed the following areas:

M-4.108 Main Steam Line Plug Maintenance M-4.216 Installation of the Main Steam Line Plugs a M-4.205 Removal of the Reactor Vessel Head Pipin No deficiencies were identifie In addition, the licensee reviewed alternate leakage paths and found the reactor cavity drains were not subject to inadvertent water-inventory losses. The licensee reviewed the recommendations of General Electric (GE) Service Information Letter No. 388 concerning RHR valve misalignment

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during a time when shutdown cooling operation was in progress. Changes to RHR cooling procedures were made; the Nuclear Review Board reviewed the actions taken relative to the IE Information Notice including tha

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e recommended procedural changes and found them satisfactor ___ _ __ - ___- -_____

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Additional Items-Alternate itakage paths investigated by the inspector included:

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Possible dislodgement of ventilation header (18"); and,

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Possible dislodgement of main steam line plug Both areas of interest resulted in a determination that both of the above concerns did not support a viable cavity leakage pathwa At the time of this inspection, the licensee's corporate mechanical engineering staff had not completed their review of the radwaste tanks regarding acceptance of refueling cavity water at.an analyzed potential leakage rate of 3300 gp Findings:

The licensee's internal review of viable leakage paths was thorough and found no additional water pathways that would drain the refueling water cavit This item is close . Previous Identified Outstanding Items (Closed) 50-277/84-BU-03 - Refueling Cavity Water Seals (Closed) 50-278/84-BU-03 - Refueling Cavity Water Seals The items are closed based on contents'of this repor . Management Exit Meeting An exit meeting was held on November 6,,1981 to discuss the inspection scope and findings, as detailed in this report (see paragraph 1.0 for attendees).

At no time was written material given to the licensee. The licensee determined that no proprietary information was utilized during this inspection.

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ATTACHMENT A Procedures / Documents Reviewed

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R.T. 7.10 Periodic Calibration Check of ARMS - Rev. 3

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Surveillance Test 2.25.10 Calibration of Flow Switch-Level FS-2692, LT 2695, LS 2675, LS 2696, LS 2697 and FS 2704 (Unit 2)

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Surveillance Test 2.30.10 Calibration of Flow Switch-Level FS-3692, LT 3695, LS 3596, LS 3697, FS 3701, FS 3704 (Unit 3)

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GP-12 Core Cooling Procedure (Unit 2 & 3)

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M-4.216 Installation of the Main Steam Line Plugs (Units 2 and 3)

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PEC0/ Mechanical Engineering Division Actions on items identified in Info Notice 84-93

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Report of ADHOC Committee to Investigate Inadvertent Loss of Inven' tory from Reactor Vessel to Suppression Pool

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PECO Calculations - Time to store assembly in transit

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PBAPS Fuel Handling Procedure FH-5 Appendix A - Evacuation of fuel floor I

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Licensee Response to IE Bulletin 84-03 (9-17-84)

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SOER 85-011 - Recommendations ,

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E-5 High Radiation - Refueling Floor Exhaust

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Calibration RX Vessel Water Level LT-3-02-3-061 M-4.108-

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Main Steam Line Plug Maintenance l

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Calibration RX Vessel Water Level LT 2-02-3-061 Unit 2 i i

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