ML20205D225

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Summary of 860616 & 17 Meetings W/Util at Plant Site Re Status of Licensing Actions,Rev to Security Plan,Battery Status Alarms & Control Room Habitability
ML20205D225
Person / Time
Site: Oyster Creek
Issue date: 08/01/1986
From: Donohew J
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 8608150306
Download: ML20205D225 (18)


Text

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[ car  % UNITED STATES 8  % NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

\ * * . [l August 1,1986 Docket No. 50-219 LICENSEES: GPU Nuclear Corporation Jersey Central Power and Licht Company FACILITY: Oyster Creek Nuclear Generating Station-

SUBJECT:

APRIL AND MAY 1986 PROGRESS REVIEW MFETING ON LICENSING ACTIONS WITH GPU NUCLEAR PLANT SITE PERSONNEL AND CORPORATE MANAGEMENT' On Monday, June 16, and Tuesday, June 17, 1986, meetings were held at Oyster Cree Station site with GPil Nuclear (the licensee) to discuss the status of station licensino actions. Attachment 1 is the list of the individuals attend-

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ing the meetings. The followino is a summary of the significant items discussed and the actions taken or proposed. References may be made tc Cycle 11 Refueling (Cycle 11R) outage which began in April 1986 and is scheduled to end in October 1986.

Attachment 2 is a marked up copy of the staff's Licensina Actions Reoort Extended (LARE) dated June 14, 1986, for Oyster Creek. The markup, to update the LARE, resulted from the discussion on each item in this mecting. The status of each item is given in the column " STAT" on the right-hand side of the LARE sheets. The. status in that column is the following: "01" means licensee, "02" means staff's reviewer, "03" means staff's Proiect Manaaer, "0L" means action completed and "05" means staff's Project Manager has the licensing action in concurr'ence.

1.0 Reactor Protection System Switch Replacement and SEP Topic VII-1.R The licensee's letter of May 27, 1986, explained that the Static-0-Rina differential pressure (SOR do) switches in the reactor water level low-level ,

function will be replaced by an analog trip system .in the current Cycle 11R outage.--Two of the analog transmitters will be paralleled with differential pressure gauges and the other two transmitters will have remote indication in the control room. This is to (1) provide level indication for a daily check i

and (2) replace the questionable SOR dp switches which have been discussed in I previous meeting summaries dated February 24 and March 14, 1986. SOR dp switches were installed in the reactor water level low-level and double low-level function in the Cycle 10M outage. This was done to meet the schedule and technical requirements of environmentally qualified electrical equipment important to safety in 10 CFR 50.49. Since this outage the SOR dp switches in t% low-level function have experienced significant drift in the switch setpoint.

This letter also applies to the staff's Systematic Evaluation Program (SEP)

Topic VII-1B, Trip Uncertainty and Setpoint Analysis Review of Operating Data Base. In the Oyster Creek Intearated Plant Safety Analysis Report (IPSAR) dated January 1983, the staff stated in Section 4.28, SEP Topic VII-1.B. that the licensee committed to install the General Electric (GE) analog trip system 8608150306 C60801 PDR ADOCK 05000219 0 PDR ,

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' AUG 1 1985 in the Cycle 11R outage. The analog trip system referred to by the licensee in its May 27, 1986, letter for the reactor water level low-level function is this GE system.

The May 27, 1986, letter also chanaes the licensee's commitment to replace the sensors RE-02s (reactor water level double low function), RE-18s (reactor water level triple low function) and RE-23s (low steam pressure to close the main steam isolation valves) with the GE system. The licensee, since its commitment in the IPSAR, has decided to replace the sensors RE-02s, RE-18s and RE-23s with SOR switches instead of the GE system and this replacement in the Cycle 11R outage has been deferred because of the problems experienced by the licensee with the setpoint drift in the SOR switches.

2.0 Licensee's Request to Revise Its Security Plan (TAC 6115?)

The licensee stated that it is withdrawing its request in its letter dated February 24, 1986, to revise its security plan for Oyster Creek. This letter requested a 10 CFR 50.90 review of changes to the station's_ security plan. The licensee ~ stater that, after consultation with the NRC Region and Headquarters, the changes do not decrease the safeguards effectiveness of the plan and, there-fore, will be made pursuant to 10 CFR 'i0.34(d) instead of 10 CFR 50.90. There-fore, the licensee's February 24, 1986, request is not needed and is. withdrawn.

The NRC Project Manager has discussed this with the Division of Safeguards, Office of Nuclear Material Safety and Safeguards, and this is acceptsble.

3.0 Battery Status Alarms (TAC 494101 This is IPSAR Section 4.32, SEP Topic VIII-3.B. DC Power System Bus Voltage Monitoring and Annunciation. The licensee explained the following:

(1) procedures (tour sheets) require that every shift it is verified that the breaker, between the de bus and de batteries, is closed.

(2) procedures (634.2.002/3) require a weekly and monthly check of the battery cell specific gravity, which indicates if the batteries need to be charged, and the batteries recharged if it is needed.

(3) procedures (643.2.00?/3) require weekly and monthly visual inspection of battery cell-to-cell and terminal connection, cell plates and cell condition.

(4) procedure (634.2.001) requires a battery capacity test and a cell-to-cell resistance test every 18 months. The licensee stated that an additional check of the resistance through the breaker can be added to this procedure.

(5) The breakers are a molded case breaker. They are hemetically sealed in a cabinet which the manufacturer recommends not opening.

MG 1 1995 The A and B battery room was visited by the NRC Project Manager in a tour of the Reactor Building. The room was clean. The A and R batteries are easily visible for inspections of the cell-to-cell connections and the terminal con-nections. The breaker are in cabinets and the breaker itself is not visible.

An indication near the lever arm, which is used to pump the breaker open or closed, shows the breaker is open or closed. This is the indication inspected in each shift tour (see item 1 abnve). The cell-to-cell connections and terminal connections were clean, in good condition and well connected.

4.0 Isolation Condenser Makeup Pump The licensee's letter of April 21, 1986, was discussed. This lotter provided additional information on the licensee's request to defer its comitment to install the isolation condenser makeup pump in the Cycle 11R outage. This was reouested in the licensee's letter dated July 26, 1985.

The licensee explained that procedures require a minimum of 20 feet or 250,000 gallons in the Condensate Storage Tank (CST). The intake tour sheet requires a minimum of 350,000 gallons in the Fire Water Storage Tank (FWST). The high wind conditions for emergency procedure 2000-ABN-3200.31 are the followinc;:

(1) tornado watch or warning, (2) hurricane watch or warning, (3) tornado funnel cloud in the area and (4) sustained wind speeds greater than 74 mph, This procedure requires the CST to be filled to 43 feet or 537,500 gallons and the Isolation Condensers to be filled (50,000 gallons). The licensee stated that it could if needed bring in a fire truck and pump water into the Isolation Condensers using an alternate connection from the fire water main.

There are no technical specifications on the FWST and the deminerialized water tank.

The NRC Project Manager toured the areas surrounding the CST, FWST and deminerialized water tank. These tanks are outside buildings near the western side of the Turbine Building or near the intake canal. See Figure 1, Site Plan, from the Updated Final Safety Analysis Report. The tanks and pumps to the Isolatien Condenser are at plant grade, 23 feet 6 inches PSL. The pumps are in steel frame, metal siding buildings on a concrete mat.

The licensee, in the April 21, 1986, letter stated that if the water level at the intake structure cannot be verified to be less than 4.5 feet above MSL a plant shutdown would commence. The NRC Project Manager toured the intake structure to see the vertical column used to measure the intake water level. The column is easily visible for determining the water level. The NPC Project Manager verified that this had been added to Procedure 2000-ABN-3200.31.

5.0 Control Room Habitability The NRC Project Manager and the Oyster Creek Licensina Manager reviewed the licensee's implementation of its commitments for the Cycle 11R outaae for control room habitability. These are in Attachment I to the licensee's June 4, 1985, letter. These are in Table 1 to this report. The results are the follow-ing in tems of the list in Table 1:

1. The chorine monitoring capability and alam in the control room has been installed. This was seen by the NRC Project Manaaer in the tour of the site.

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?. The licensee stated that it has implemented the preventative maintenance program.

3. The licensee stated that it had installed the weather strip material.
4. The licensee stated that the switch had been installed.
5. The licensee stated that it will propose the appropriate Technical Specifications (TS) before restart from the Cycle 11R outage.
6. The licensee stated that it will develop the procedures and train the operators before the restart from the Cycle 11R outage.
7. The licensee submitted this on August 16, 1985.
8. The licensee submitted this on June 17, 1985.

As part of submitting the appropriate TS, the licensee must submit the results of testing the control room for the air inflow during the minimum air inflow mode. A minimum air inflow of 450 cfm was assumed in the licensee's submittals for items 7 and 8 above.

The licensee stated that the 1-ton chlorine tanks in the chlorine facility are to be replaced by sodium hypochlorite by April 1987. There will be a small tank of chlorine gas stored near the intake structure for the new radwaste service water. This will be farther away from the control room ventilation intake than the chlorine facility.

6.0 Time Delay to Open the SGTS Inlet Valves In its letter dated April 18, 1926, on the licensee's request to cancel the modification to install a pressure relief vent in the SGTS ductwork, the staff stated that the licensee had installed the 5-second delay to start openina the SGTS filter inlet valves following the loss-of-coolant acci, dent (LOCA). This was a provision made by the licensee to protect safety-related equipment located outside the containment against loss of function from the environment created by the escaping air and steam during the LOCA. The only safety-related equip-ment located downstream of the containment ventilation isolation valves is the SGTS.

A test was run of this delay on June 17, 1986 at the request of the NRC Pro.iect Manager. The SGTS was initiated by inputting a false sional in one of the radiation monitors in the Reactor Building ventilation exhaust duct. The inlet valves of the two SGTS opened after 5 seconds following the initiation of the SGTS.

7.0 Suppression Pool Temocrature for the LOCA In its letter dated May 5,1986, the staff stated that the initial suppression ,

pool temperature for the design basis LOCA shall be consistent with the most conservative pool temperature allowed by the TS for any extended period

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. 1 mag of time. Usirg this initial temperature in the LOCA analysis, there should be net positive suction head (NPSH) for the operation of any safety system pump drawing from the suppression pool. The licensee explained that TS 3.5.A.7 requires that the suppression pool temperature must be less than 95"F or reactor shutdown is initiated and the reactor shall be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. -

Based on this the initial suppression pool temperature for the LOCA analysis should be 95 F and the licensee's calculations for this temperature show that the core spray pumps have NPSH. See Section 17.0 of the meeting sumary dated May 22, 1986.

The NRC Project Manager has discussed this with the Reactor Systems Branch in the Division of RWR Licensing. The staff agrees that based on TS 3.5.A.7 the initial suppression pool temperature for the LOCA analysis is the 95 F. The fact that TS 3.5.A.1.C(2) allows the pool temperature to exceed the 95 F limit by no more than 10 F for no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, for testing which adds heat to the pool, does not change this conclusion.

8.0 Undated NRR Licensina Action Report Extended (LAREl Dated 04/20/86 Attachment 2 has the updated LARE for Oyster Creek. The updating was done during the discussion on each licensing action in this meeting. The licensing actions are listed by TAC number (left hand column of LARE).

The LARE is a print out from the staff's PC licsasing act~ ion tracking system.

The LARE contains references to future licensing actions to be submitted by the licensee. These future actions have TAC number OCXXX.

9.0 Next Meetina The June 1986 Progress Review Meeting is expected to be held at the 1(censee's headquarters in Parsippany, Ne ersey on July 30, 1986, ac N. onoh Jr., Project Manager RWR Project Directorate No. 1, l Diirision of BWR Licensjna Attachments:

1. List of Attendees
2. Staff's Updated LARE dated 6/14/86 cc w/ attachments:

R. Bernero/R. Houston G. Lainas/B. D. Liaw D. Vassallo/C. Grimes W. Hodges L. G. Hulman M. Srinivasan BWR#1/DP JPL RWR#1/0PL CJamerson Don ew wolinski 07/3)/86 '7g/86 pq/86

l cc:

Mr. Ernest L. Blake, Jr. Resident inspector Shaw, Pittman, Potts and Trowbridge c/o U.S. NRC 1800 M Street, N.W. Post Office Box 445 Washington, D.C. 20036 Forked River, New Jersey 08731 J.B. Liberman, Esauire Commissioner

. Bishop, Liberman, Cook, et al.

  • New Jersey Department of Energy 1155 Avenue of the Americas 101 Commerce Street New York, New York 10036 Newark, New Jersey 07102 Mr. David Scott, Acting Chief Regional Administrator, Region I Bureau of Nuclear Engineering U.S. Nuclear Regulatory Commission Department of Environmental Quality 631 Park Avenue 380 Scotch Road King of Prussia, Pennsylvania 19406 Trenton, New Jersey 08628 RWR Licensing Manager P. 8. Fiedler  ;

GPU Nuclear Vice President & Director 4 100 Interpace Parkway Oyster Creek Nuclear Generating Parsippany, New Jersey 07054 Station -

Post Office Box 388 Deputy Attorney General Forked River, New Jersey 08731 State of New Jersey Department of Law and Public Safety 36 West State Street - CN 112 Trenton, New Jersey 08625 Mayor Lacey Township 818 West Lacey Road Forked River, New Jersey 08731 Mr. D. G. Holland Licensing Manager i Oyster Creek Nuclear Generating Station Post Offl.e Box 388 Ferked River, New Jersey 08731 l

TABLE 1 Interim System Upgrades for Control Room Habitabiilty at OCNGS .

1. The Licensee will install chlorine monitoring capability which will provide an alarm in the Control Room to alert operators in the event of a chlorine leakage condition.
2. The Licensee will develop and implement a preventive maintenance program on Control Room HVAC Ducts and Damcers to ensure system integrity is I being maintained and that leakage remains low.
3. The Licensee will install weatherstrip material on the two doors which I are not used for normal access into the Control Room.
4. In order to override the existing thermostatic controls, the Licensee will install a switch which will allow operators to either isolate the Control Room or place the Control Room HVAC system into the Recirculation Mode.
5. The Licensee will propose appropriate Technical Specifications for the Control Room HVAC System.
6. The Licensee will devalop radiation and chlorine alarm response procedures for the control room operators to take the appropriate actions in response to either of these alarms.
7. The Licensee will provide a Chlorine Transport Analysis to demonstrate that the control room operators will have at least two minutes to respond l to a chlorine leak alarm. This analysis will be submitted to the NRC by August 15, 1985.
8. The Licensee will provide calculations and analysis for whole body and beta skin doses using Ragulatory Guide 1.3 source term. If necessary, procedural guidance for protective measures to be taken by the control room operators such as the usage of protective clothing and goggles will be developed. The results of these calculations and the assumptions and models for the analysis will be submitted to the NRC.by June 14, 1985.

Because the NRC staff is presently reviewing the lodine source term for the design basis LOCA accident, the thyroid exposure limit will not be addressed.

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Attachment 1 APRIL-MAY 1986 PROGRESS REVIEW MEETING June 16-17, 1986 Ovster Creek Site Name Affiliation J. Donohew NRC/NRR? DBL M. Laagart GPUN*

J. Kowalski GPUN V. Foglia GPUN D. Pino GPUN

  • GPU Nuclear Corporation i

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t' i Distribution for Meeting Sumary Dated:

L Facility: Oyster Creek Nuclear Generating Station *

(Beshot4 File lJ Q NRC PDR Local PDR -

BWD1-Reading R. Bernero R. Houston J. Zwolinski J. Donchew C. Jamerson OGC-Beth (For info only)

E. Jordan B. Grimes ACRS (10)

G. Lainas B. D. Liaw W. Hodges D. Vassallo C. Grimes G. Hulman M. Srinivasan DC Files R. Manili

  • Copies sent to persons on facility service list

e.

Distribution for Meeting Summary Dated:

Facility: Oyster Creek Nuclear Generating Station

  • Docket File (50-2191 NRC PDR Local PDR BWD1 Reading R. Bernero R. Houston J. Zwolinski J. Donohew C. Jamerson OGC-Reth (For info only)

E. Jordan B. Grimes ACRS (10)

G. Lainas B. D. Liaw W. Hodges D. Vassallo C. Grimes G. Hulman M. Srinivasan OC Files R. Manili

  • Copies sent to persons on facility service list i

J l

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