ML20126H144

Summarizes 850319 Meeting W/Util & S&W at Plant Site to Discuss Control Room Habitability,Per NUREG-0737, Item III.D.3.4.Extension to 830314 Confirmatory Order from Cycle 11 to Cycle 12 Refueling to Implement Mods Requested

ML20126H144
Person / Time
Site:	Oyster Creek
Issue date:	06/04/1985
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	Zwolinski J Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-3.D.3.4, TASK-TM NUDOCS 8506100278
Download: ML20126H144 (8)

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y-GPU Nuclear Corporation NggIgf Post Office Box 388 Route 9 South Forked River,New Jersey o8731-o388 609 971-4000 Writer's Direct Dial Number:

June 4, 1985 Mr. John A. Zwolinski, Chief Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Control Room Habitability (NUREG-0737 Item III.D.3.4)

A meeting was held at the Oyster Creek Nuclear Generating Station on March 19, 1985 to discuss Control Room Habitability (NUREG-0737 Item III.D.3.4).

Attending this meeting were representatives from the NRC staff, GPU Nuclear Corporation, and Stone and Webster Engineering Corporation. The objectives of this meeting were to review the drawings of the existing Control Room HVAC System, tour the facility to examine the arrangement and condition of the existing equipment, and discuss the proposed interim system upgrades and the design objectives for the final system modification.

As a result of this meeting, all previous commitments, including the Confirmatory Order dated March M,1983, on Control Room Habitability for the Oyster Creek Nuclear Generating Station were superceded. The new commitments which were agreed to by members of the NRC staff are attached to this letter and reflect our understanding of the NRC minutes of that meeting dated April 16, 1985. - Therefore, the licensee requests an extension to the Confirmatory Order dated March 14, 1983 from the Cycle 11 to the Cycle 12 refueling outage. This request is necessary because of a misunderstanding of the previous licensee correspondence on the part of the NRC staff.

Attachment I outlines the proposed interim measures which will be in place at Oyster Creek Nuclear Generating Station by the end of the Cycle 11 refueling outage. Attachment II provides the proposed final modification design objectives for the Control Room Habitability System. Any final modifications would be planned for completion during the Cycle 12 refueling outage.

Attachment III responds to the six questions on Control Room Habitability contained in the NRC staff's letter dated February 22, 1984.

In addition, GPU Nuclear Corporation understands that the commitments hereby forwarded will satisfy the requirements of Item III.D.3.4 of NUREG 0737.

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GPU Nuclear Corporation is a subsidiary of the General Pubhc Utilities Corporation l }

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- If you have any questions, please contact M. W. Laggart, Manager, BWR Licensing at (201) 299-2341.

Very truly yours,

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F. v. .sedler Vice President and Director Oyster Creek Ir/1594f cc: Administrator Region I U.S. Nuclear Regulatory Conmission 631 Park Avenue King of Prussia, Pa. 19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, N. J. 08731

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Attachment I Interim System Upgrades for, Control Room Habitability at OCNGS l '. 1The 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.

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2. The Licensee will develop and implement a preventive maintenance program on Control. Room HVAC. Ducts and Dampers to ensure system integrity is being maintained and that leakage remains-low.

3. - The Licensee will install weatherstrip material on the two doors which are'not used for normal access into the Control Room.

4. In orde" to override the existing thermostatic controls, the Licensee will ir. tall 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. 'Thi' Licensee will develop 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 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 Regulatory 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 avill 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-iodine source term for the design basis LOCA accident, the thyroid exposure limit will'not be

-addressed.

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• Attachment II Long Term / Final Modification Design Objectives for the Control Room Habitability System at OCNGS

1. The Licensee will perform a Single Failure Analysis of the Control Room HVAC System, address all potential problem areas,.and provide remedial measures. Any modifications will be implemented before the end of the

. Cycle 12 refueling outage and will be consistent with the criteria defined by the NRC staff at the March 19, 1985 meeting.

2. The Licensee will assess existing diesel generator capability in order to provide back up power to the final Control Room HVAC system design.

3. The Licensee will meet the Beta skin dose limits with the final Control Room HVAC system design without protective clothing and goggles.

- 4. The Licensee does not have to meet the natural phenomena criteria.

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Attachment III Response to NRC Request for Additional Information Dated February 22, 1984 NRC Question 1 Justification for all active components of the proposed control room emergency ventilation system (such as fans, isolation dampers, chillers, and radiation monitors) that will not be redundant and/or single failure proof following modification.

Response to Question 1 The single failure analysis for the control room ventilation system will be completed for the Cycle 12 outage as stated in item #1 of Attachment II.

NRC Question 2 Identify proposed ~ technical specification requirements for periodic leak testing of the control room emergency zone, filter bypass dampers, outside air dampers and/or valves, and ESF leakage outside containment or provide justification if these leak tests are not proposed.

Response to Question 2 The licensee will propose the appropriate Technical Specifications for the Control Room HVAC System in accordance with item #5 of Attachment I.

NRC Question 3 Identify the locations of radiation release points from design basis accidents other than the LOCA, relative to control room outside air intake locations.

To support your assessment, provide your bases and relevant layout drawings.

Response to Question 3 The following design basis accident (DBAs) are considered in determining the control room radiological habitability. In addition, their associated airborne radiation release points are indicated and provided below for each DBA.

DBAl - Large Line LOCA Inside Primary Containment (SRP 15.6.5) Release Points

1. Turbine / Condensers

2. Main stack Bases-

1. Main steam isolation valve (MSIV) bypass leakage travels through the main steam lines and is released to the Turbine Building at the Turbine / condensers (reference FSAR Figure 1.2-40 and Branch Technical Position CSB 6-3).

2. Primary containment leakage and ESF equipment leakage are collected and processed by the standby gas treatment system (SGTS) and released through the main stack (reference FSAR Figure 6.5-1A, SGTS ,

description - FSAR Section 6.5, and Appendices A and B to SRP 15.6.5).

DBA2 - Control Rod Drop (SRP 15.4.9)

Release Points:

Turbine / condensers Bases Assuming that the main steam system maintains integrity during this accident, the activity travels from the reactor vessel to the turbine / condensers by means of the main steam lines (reference Appendix A to SRP 15.4.9).

DBA3 - Small Line LOCA Outside Primary Containment (SRP 15.6.2)

Release Points Reactor building Bases The_ loss of coolant outside the primary containment is conservatively assumed to increase the pressure in the reactor building to greater than

-0.25 WG, negating secondary containment. The airborne fission products therefore are released at ground level from the reactor building (reference Branch Technical Position CSB 6-3).

DBA4 - Main Steam Line Break Outside Containment (SRP 15.6.4)

Release Points Main steam tunnel (MST)

Bases The line break occurs in the MST with the tunnel maintaining structural integrity. The fluid released from the ruptured line follows the path of least resistance which would lead to a release at the turbine building side of the main steam tunnel (reference FSAR Section 15.1.5.2).

DBA5 - Fuel Handling Accident (SRP 15.7.4)

Release Points Main Stack Bases-The activity released from the reactor vessel or spent fuel pool is exhausted throught the main stack by means of normal reactor building ventilation prior to isolation and SGTS initiation due to high radiation in the exhaust stream. After isolation, the activity continues to be-released through the main stack after processing by the SGTS. (reference FSAR Figure 6.5-1A and SGTS description - FSAR Section 6.5).

NRC Question 4 Information which shows that the control room emergency ventilation system is designed to function properly in the event of a loss of offsite power or pipe breaks in areas adjacent to the control room.

Response to Question 4 Details for powering the CR ventilation system will be provided for the Cycle 12 outage and will include single failure analysis considerations in accordance with items #1 and #2 of Attachment II. Pipe breaks in the areas adjacent to the Control Roon will be examined during the single failure analysis as a potential mechanism which may initiate a single active failure.

NRC Question 5 Analysis of control room operator doses following postulated design basis accidents. It may be necessary to assess DBAs other than a LOCA if radiation release points are closer to the control room air intake than that for a LOCA. The. analysis should include a detailed listing of data and assumptions used, as well as the results.

Response to Question 5 Using information from the 1983 Oyster Creek Integrated Plant Safety Assessment (NUREG-0822) and considering the appropriate source terms, release points, and relevant design data, the number of DBAs requiring an analytical comparison to determine the limiting case was narrowed to two: the large line LOCA and the small line LOCA outside the primary containment.

Considering the present control room HVAC operating data, the doses for the two accidents were calculated using standard design basis assumptions and source terms (i.e., Reg. Guide 1.3 source terms for the large line LOCA). The results showed that the large line LOCA is the limiting case DBA for gamma and beta doses.

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A detailed listing of the data, assumptions, and results will be available upon finalization of the control room habitability analyses. This will be submitted by June 14, 1985 in accordance with item #8 of Attachment I.

NRC Ouestion 6 Analysis wnich shows that the rate of increase of toxic gas concentration in the control room will be slow enough to allow the control room occupants sufficient time to don respiratory equipment.

Response to Question 6 The-licensee will perform a Chlorine Transport Analysis which will demonstrate that the Control Room operators will have at least two minutes to respond to a chlorine leak alarm. This analysis will be submitted by August 15,1985 in accordance with item #7 of Attachment I.

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