ML19263D287
| ML19263D287 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 03/23/1979 |
| From: | Parker W DUKE POWER CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7903270336 | |
| Download: ML19263D287 (3) | |
Text
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DUKE POWEH COMI%NY I'OWEH Ut!!1. DING 422 Sourn Citt:ucn STurET, CnAHIOTTe, N. C. 2a242 W 8 L LI AM O. PA R K E R, J R.
9 Ver Persiof NT TCL(PHONE: Aeta 704 51r ame PaoovcTiog 373-4003 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Mr. Robert L. Baer, Chief Light Water Reactor Project Branch No. 2
Subject:
McGuire Nuclear Station Unit 1 and 2 50-369, 50-370
Dear Mr. Denton:
Please find attached the status of the outstanding review matters trans-mitted by your letter of February 8, 1979. You will note that much of the requested information is still under preparation and will be submitted at a later date.
7 V y truly yours,/
/
un (d. eto b.,
William O. Parker, J GAC:ses Attachment PROGapE OOk Sip
/q g
i th Anniversary 7003370.33(,
ATTACHMENT 1 OUTSTANDING REVIEW MATTERS - STATUS MCGUIRE NUCLEAR STATION - UNITS 1 AND 2 1.
Provide your evaluation of the degree of conformance of the McGuire design with the Commission's Regulatory Requirements Review Committee Category 2 and 3 matters.
Our evaluation of each of these items is in progress. A response to each of these items will be provided by April 6, 1979.
2.
Provide a complete description of the load follow strategy which has led to the increase in the allowable axial flux difference control band from
+ 5 percent to 3 percent to -12 percent. This change appears in your proposed Technical Specification 3.2.1.
Reference 6, Section 4.3.5 of the FSAR only supports use of a + 5 percent axial flux difference control band.
Also, provide the results of analyses you have performed justifying the wider axial f]ux difference control band.
These analyses are in progress and are expected to be available by April 20, 1979.
3.
Provide analyses to identify the maximum containment temperature response for a postulated main steam line break accident.
(See letter dated reb-ruary 27,1979.)
Westinghouse is currently performing typical plant analyses which will identify the limiting steam line breaks. A response to this item will be provided by April 30, 1979.
4.
Provide Class IE environmental qualification documentation for safety related equipment.
Additional information requested by the Staff in a November 30, 1978 tele-phone conversation will be provided by April 30, 1979.
5.
Using the Alden Research Laboratory test results, provide the NPSH for the emergency core cooling system pumps.
The scale model testing performed by Alden Research Laboratory adequately demonstrated the hydraulic performance of the McGuire containment sump.
However, this scale model testing could not be used to verify adequate NPSH for the containment spray pumps on residual heat removal (RHR) pumps.
A preoperational test was performed simulating the injection mode of Contain-ment Spray System operation in order to verify the NPSH results obtained via theoretical calculations.
During the test, pump flow rate, Refueling Water
ATTACHMENT 1 Page Two Storage Tank level, water temperature and pump suction pressure were obtained. This information was then utilized to determine the actual NPSH availabic for the containment spray pumps. The preoperational test values for NPSH are compared to the theoretical values below.
NPSH (ft)
Flow (GPM)
I pump (test) 120.5 3400 1 pump (theoretical) 117.1 3400 2 pumps (test) 111.2 3450 2 pumps (theoretical) 106.5 3450 This comparison verifies that our theoretical method is conservative for calculating available NPSH. This method was used to calculate the NPSH values shown in the FSAR for the RHR and Containment Spray pumps.
6.
Provide a summary of plant procedures demonstrating core cooling capability for postulated failure in a residual heat removal line.
This procedure is currently under development and is expected to be avail-able by April 30, 1979.
7.
Provide additional LOCA analysis information; Q 212.110 (See letter dated December 1, 1978).
This information will be included as part of Revision 36 to the McGuire Final Safety Analysis Report. This revision will be filed by April 20, 1979 8.
Provide additional information regarding augmented inservice inspection for pipe rupture protection as agreed upon at the January 18, 1979 meeting (See letter dated February 9, 1979).
This information was transmitted by letter dated March 22, 1979 from Mr.
William O. Parker, Jr. to Mr. Harold R. Denton.
9.
Provide acceptable radiological effluent technical specifications (See letter dated February 7, 1979).
This information was transmitted by separate letter dated March 23, 1979 from Mr. William O. Parker, Jr. to Mr. Harold R. Denton.
10.
Provide additional information relating to the spent fuel cask drop accident analysis (See letter dated February 6, 1979).
This information was transmitted by letter dated 'tarch 2,1979 from Mr. William O. Parker, Jr. to Mr. Harold R. Denton.