ML20147E481

Documents Util Position & Forwards Addl Info Re NUREG-0737, Item III.D.3.4, Control Room/Technical Support Ctr, Including Brief Discussion of Leakage Tests Results & Method & site-specific Data Used in Determining Cr/Tsc X/Q Value

ML20147E481
Person / Time
Site:	Rancho Seco
Issue date:	02/25/1988
From:	Andognini G SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Miraglia F Office of Nuclear Reactor Regulation
Shared Package
ML20147E485	List: ML20147E481 ML20147E499 ML20147E520 ML20147E536
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-3.D.3.4, TASK-TM GCA-88-104, NUDOCS 8803070076
Download: ML20147E481 (6)

Text

. - _ -

. - ~ ,

hSMUDSACRAMENTO MUNICIPAL UTIUTY DISTRICT O P. O. Box 15830, Sacramento CA 95852-1830,(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA FEB 2 51988 GCA 88-104 U. S. Nuclear Regulatory Commission Attn: Frank J. Miraglia, Jr.

Associate Director for Projects 11555 Rockville Pike Rockville, MD 20852 Docket No. 50-312 Rancho Seco Nuclear Generating Station License No. DPR-54 CR/TSC HABITABILITY, NUREG 0737 ITEM III.D.3.4

Dear Mr. Miraglia:

A meeting was held with the NRC staff on February 11, 1988, in Rockville, Maryland concerning the District's CR/TSC habitability calculations. During this meeting, presentations were made by the District with respect to the CR/TSC X/Q, post-accident radiation dose and habitability after an ammonia spill. Subsequent to this meeting, several discussions and calculational iterations occurred on the subject of CR/TSC habitability. This letter documents the District's position and provides additional information requested by the staff.

During the meeting, the staff indicated some concern regarding the location of the CR/TSC Essential HVAC intakes. The District will address this concern by pursuing a wind tunnel study to verify the current intake locations. If the results of this study indicate the current locations require modifications, the intakes will be relocated to a position that will result in acceptable CR/TSC doses. This effort will be initiated after plant restart, and any modifications which may result will be installed prior to returning to power from our next refueling outage.

Review of the original leak rate test (ANSI /ASME N510-1980) of the CR/TSC Essential HYAC System indicated leakage in Train A in excess of the calculated inleakage value (35 cfm) used in the dose analysis. As a result, the dose calculation has been revised using the actual values obtained from the test. Attachment 1 contains a brief discussion of the leakage test results. Leakage testing and repair, if needed, will be performed on the suspect air handling units (AH-A-545 A&B) to bring the leakage within acceptable values prior to returning to power from our next refueling outage.

~

8803070076 8'80225 PDR P

ADOCK 05000312 8I l P DCD RANCHO SECO NUCLEAR GENERATING STATION O 1444o Twin Cities Road. Herald, CA 95638-9799;(209) 333 2935

e f

l GCA 88-104 l Frank J. Miraglia, Jr. FEB 2 51988 i i

Attachment 2 contains a discussion of the method and site specific data used by the District in the determination of the CR/TSC X/Q value. The resultant X/Q value is based on the current location of the CR/TSC Essential HVAC intake. This X/Q value was used as input to the post-accident CR/TSC-radiation dose calculation. Assumptions and results of the calculation are presented in Attachment 3.

Calculations indicate that the dose limits will be met in both the Control Room and the Technical Support Center if protective action is taken to limit the beta and thyroid radiation exposure. The District considers the dose results to be acceptable for the interim, given the conservatisms involved, the aforementioned commitments and the actions that would be implemented in the event of an accident via our Emergency Plan. The current Emergency Plan-and its associated procedures provide for the availability and administration of potassium iodide (KI) as well as requiring periodic CR/TSC radiation surveys to determine the radiation environment and protective measures (e.g., protective clothing), if any, for assuring continued habitability of these areas.

At the February 11, 1988 meeting, the District presented two analyses for the ammonia spill. These analyses (250C and 380C) indicated that there was adequate time (>2 minutes) to take protective actions prior to incapacitation of the operators. In the 380C case the incapacitation model presented in NUREG/CR-1741 was used to determine the time to incapacitation. It is the District's understanding that the use of incapacitation models has been used by other licensees whose control rooms have been found to be acceptable. The District has provided the results of several additional ammonia cases (Attachment 4) for the staff's review and consideration.

There was some question by the staff on the use of the incapacitation model, and the District agreed to evaluate the 250C case and determine at what point in the year the 250C would be expected to be exceeded more than 5% of the time. Attachment 5 contains a discussion of that data review. The review indicates that 250C will not be exceeded more than 5% of the time until after June 1988. The District will continue to pursue resolution of the ammonia issue with the staff to ensure closure of the issue prior to July of this year.

During conversations with the staff we have provided the staff with updated information on the air balance data. Attachment 6 provides a description of the revised data and the supporting data sheets, i

a

r-

.. 3 GCA 88-104 Frank J. Miraglia, Jr. FEB 2 51988 Please. contact me if you have any questions. Members of your staff with questions requiring additional information or clarification may contact Mr. John Atwell at (209) 333-2935, extension 4917.

Sincerely, W'#*

. Carl Ando ni Chief Executive Officer, Nuclear Attachments cc: G. Kalman, NRC, Rockville A.~D'Angelo, NRC, Rancho Seco J. B. Martin, NRC, Walnut Creek

, s.

GOA 88-104-Frank J. Miraglia, Jr. 'bc:

w/atch w/o atch X' General Manager MS 41 X T ' Chief Executive Officer, Nuclear MS 209 X Executive Assistant MS 204 X AGM, Nucl. Tech. & Adm. Services MS 206 AGM, Nuclear Power Production MS 254 X Director, Nuclear Quality MS 271 Director, Nuclear. Ops. & Maint. MS 257 T Director, Plant Support MS 258 Director, Sys. Rev. & Test Prgm. MS 259 Modifications Manager MS 201 T Manager, Nuclear Training .MS 296 X Manager, Nuclear Licensing MS 298 X Manager, Nuclear Engineering MS 208-6 Manager, Cost Control Services MS 270 X Public Information MS 299 Manager, Maintenance MS 254 Manager, Operations MS 255 T Manager, Env. Protection MS P.92A Manager, Rad. Protection MS 244 Manager, Nuclear Chemistry MS 244 Manager, Plant Performance MS 258 Risk Manager (D. Nears) MS 40 MSRC Secretary (J. Palmer) MS 274 Surveillance Coordinator MS 278 X IIRG MS 298 T NAC (6) MS 204 T..Baxter F. Burke (B&W)

LER Files MS 286 T Licensing Files MS 286 X PRC Files MS 286 X RIC Files MS 224 Special Report Files MS 286 Tech. Spec./PA Files MS 298 N0Y/N00 Files MS 286 Licensing Verification MS 286 X Elizabeth Gebur MS 286 T Jerry Delezenski MS 298 X Bob Little (Bethesda)

e 3 ATTACHMENT 1 As discussed with the staff, the District has reviewed the leak rate testing performed on the CR/TSC Essential HVAC System and found that the tested leakage exceeds the 35 cfm assumed in the dose analysis presented on February 11, 1988. Test results indicate that Train "A" leakage was approximately 42 cfm while Train "B" was measured at approximately 18 cfm; thus, the leakage from Train "A" exceeds the assumed inleakage value of 35 cfm. Detailed review of test results indicates that the majority of the leakage was from the air handling units AH-A-545A & B (40 cfm and 14 cfm, Train A and B, respectively).

In addition, the testing was performed at 8.1 inch w.g. while the system is expected to operate at a pressure of 5.25 inch w.g. Taking this difference of pressures into account and conservatively adding both trains results in an )

, inleakage of 50 cfm plus the 10 cfm for access and egress for a total inleakage of 60 cfm. This value has been used in the dose analysis provided in Attachment 3.

The testing was reviewed and determined to be done in a manner consistent with ANSI /ASHE H510-1980 with the exception of the air handling units discussed above. The air handling units were tested using a calibrated orifice plate rather than the gas meter or decay methods described in the ANSI criteria.

The District intends to retest these units using the gas decay method to determine leakage of the units and to repair, as needed, to bring the leakage within acceptable values.

Another concern was raised with respect to the TSC being within the CR/TSC envelope and the potential for excessive inleakage. In the 1987 Emergency Drill and Exercise Report an observation was made that the CR/TSC envelope was being broken and that the doors should be marked to make people aware that opening the doors results in breaking the envelope. The District will resolve this issue by preparing signs. These will be posted on the doors to the envelope boundary during an emergency and indicate the doors as CR/TSC boundary doors.

Considering the additional inleakage assumed in the calculation of a continuous rather than intermittent 10 cfm, and the positive pressure inside the envelope, the District considers the potential leakage as a result of the TSC access / egress to be sufficiently minimized and bounded by the 10 cfm assumed in the calculation.

1

l. .

ATTACHMENT 2 X/Q Methodology

. . -. - _ - . . -