ML19308C743
| ML19308C743 | |
| Person / Time | |
|---|---|
| Site: | Salem, Mcguire, Sequoyah, Diablo Canyon, McGuire  |
| Issue date: | 01/09/1980 |
| From: | Gammill W Office of Nuclear Reactor Regulation |
| To: | Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| NUDOCS 8002010233 | |
| Download: ML19308C743 (1) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION o
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WASHINGTON. D. C. 20555 k..* /
JAn 9 1950
- e MEMORANDUM FOR:
E. L. Jordan, Assistant Director for Technical Programs, Division of Reactor Operations Inspection FR0ft:
W. P. Gammill, Acting Assistant Director for Operating Reactor Projects, Division of Operating Reactors
SUBJECT:
IE DRAFT BULLETIN, OVERPRESSURIZATION OF CONTAINMENT OF A PWR PLANT AFTER A MAIN STEAM LINE BREAK Enclosed is a proposed revision to the subject bulletin.
The proposed revision addresses two additional concerns that were not included in the original draft bulletin:
(1) Each licensee should consider his ability to detect and isolate the damaged steam generater, and (2) Each licensee should review his analysis of the reactivity increase which results from a main steam line break.
The proposed revision includes comments from Plant Systems Branch, Reactor Safety Branch, and Containment Systems Branch.
If you have any questions concerning this proposed revision, contact Elinor Adensam of the Plant Systems Branch on X-28058.
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W. P. Ga il, Acting Assistant rector for Operating Reactor Projects Division of Operating Reactors
Enclosure:
As stated cc w/ enclosure:
D. Eisenhut J. Kerrigan P. Check M. Wilber W. Butler T. Novak D. Shum P. Matthews F. Eltawila
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80 0 2 010 L3.3
UNITED STATES SSINS No.: 6820 NUCLEAR REGULATORY C0KMISSION A:c2ssicn No.:
OFFICE OF INSPECTION AND ENFOR EMENT 7910250504 WASHINGTON, D.C.
20555 Decer.ber
, 1979 Draft IE Bulletin OVERPRESSURIZATION OF CONTAINMENT OF A PWR PLANT AFTER A KAIN STEAM LINE BREAK Description of Circumstances:
Virginia Electric and Power Co. submitted a report to the Nuclear Regulatory Commission dated September 7,1979 that ider.tified a deficien:y in the original analyses of containment pressurization as a result of a stea line break for North Anna Power Station, 'Jnits 3 and 4.
Stone and Webster Engineering Corporation performed a reansiysis of containment pressure following a main steam line break and determined that, if the auxiliary feedwater system continued to supply feedwater at runout condi' ions to the steam generator that had experienced the steam line break, containcent design pressure would be exceeded in approximately 10 minutes.
The lonc tert blowdown o'f the water supplied under runout conditions by tne auxiliary feed ater system had not been considered in the earlier analyses.
On October 1,1979, this information was provided to all holders of operating licenses and construction permits in IE Information Notice 79-24. The Palisades facility did an accident analyses review pursuant to the informaticn in the notice and discovered that with offsite power availabl% the condensate pumps would feed the affected generator at an excessive rate. This excessive feed was not analyzed in the analyses for the main steam line break accident.
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Kerrigan/ REDRAFT 12/19/79 Draft IE Bulletin Actions to be Taken by the Licensee:
For all pressurized water power reactors with an operating license and those reactors listed in Enclosure 1:
1 1.
Review the containment pressure response analyses to detennine if the potential for containment overpressure for a main steam line break inside containment included the impact of runout flow
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r from the auxiliary feedwater system or the impact of other energy sources, such as continuation of feedwater or condensate
.g flow.
In your review, consider your ability to detect and isolate the damaged steam generator from these sources.
2.
Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment. This i
review should consider the reactor cooldown rate and the potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position.
If your previous analysis did not consider all potential water sources (such as those listed in 1 above) and if the reactivity increase is excessive, the report of this review should include:
The boundary conditions for the analysis, e.g., the end a.
of life shutdown margin, the moderator temperature coefficient, power level and the net effect of the associated steam generator water inventory on the reactor system cooling, etc.,
b.
The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system,
-2 c.
The effect of extended water supply to the affected steam generator on the core criticality and return to
- power, 4
d.
The hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the MDNBR values for the analyzed transient.
-l 3.
If the potential for containment overpressure exists o~r the reactor return to power response worsens, provide a proposed corrective action and a schende for completion of the corrective action.
4.
If the potential for containment overpressure exists or the reactor return to power response worsens, and the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action described in Item 3 above is completed.
A report of the above actions shall be submitted within 90 days of the receipt of this Bulletin.
Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Operations. Inspection, Washington, D.C.
20555.
For boiling water reactors with an operating license or a construction permit and all pressurized water reactors with a construction permit, not listed in Enclosure 1, this Bulletin is for information purposes only and no written response is required.
Approved by GAO, B180225 (P.0072); clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for identified generic problems.
P-Enclosure No. 1 Plants with construction permits that are req:. tired to respond to 'the bulletin:
Diablo Canyor.
McGuire Salem 2 Sequoyah If the permit holders have responded to earlier requests from the NRC on some of the items presented in the bulletin, they may respond to the bulletin by reference to the response to the earlier request.
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