ML19309H488
| ML19309H488 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 05/07/1980 |
| From: | Vandenburgh D Maine Yankee |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| IEB-80-04, IEB-80-4, WMY-80-72, NUDOCS 8005130374 | |
| Download: ML19309H488 (5) | |
Text
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. MAIRE HARHEE T ~ ~ :F0!?E n" ^ *
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- %,. ' -.....- g May 7, 1980 B.3.2.1 WMY 80-72 United States Nuclear Regulatory Commission Washington, D. C.
20555 Attention: Office of Nuclear Reactor Regulation
References:
- 1. License No. DPR-36 (Docket No. 50-309)
- 2. IE Bulletin 80-04, February 8,1980
Auxiliary Feedwater Systems, dated January 9, 1980
- 4. YAEC-1132, " Justification for 2630 MWT Operation of the Maine Yankee Atomic Power Station," dated July 1977
Subject:
Response to IE Bulletin No. 80-04
Dear Sir:
Reference 2 requested that Maine Yankee review its main steam line break analysis to determine the potential impact on containment pressure and core criticality of continuation of flow to a damaged steam generator from any potential source. The results of this review are provided below.
Q1.
Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater or condensate flow. In your review, consider your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runou't flow.
A1.
The impact of runout flow from the auxiliary feedwater system on containment pressure was provided in Reference 3, where it was determined to be bounded by the Reference 4 analysis. Reference 4 considered continuation of feedwater flow (8788 GPM 31% of total full power flow rate) to the damaged steam generator at flow rates in excess of both the main feedwater system and the runout flow of the auxiliary feedwater
e United States Nuclear Regulatory Commission May 7, 1980 Page 2 pumps. Although auxiliary feedwater was not directly considered, the sensitivity analyses performed as part of this study indicate that the key assumption with respect to containment pressure is how fast the intact steam generators are isolated by the action of the non-return and excess flow check valves and is rather insensitive.to continuous auxiliary feedwater addition. In the worst case the peak containment pressure is 45 psig and occurs 112 seconds into the event.
The Reference 4 analvsis assumed continuation of main feedwater flow to the affected steam generator through the open main feedwater regulating bypass valve along with leakage past the closed main feedwater regulating valve (MFWRV). The continuation of feedwater and condensate flow to the affected steam generator through an open main feedwater regulating valve (i.e., failure of the MFWRV to close on turbine trip) has not been previously addressed. Continued addition of feedwater through a full open MFWRV has the potential for containment pressures above the containment design value of 55 PSIG. For a major break this would occur in approximately four minutes. The plant emergency procedures are being revised to direct the operator to trip main feedwater pumps and close the main feedwater MOVs if the MFWRV to the affected steam generator fails to close.
During a major steam line break accident, the operator has a variety of indications available in the control room for determining and isolating the damaged steam generator.
In the short term, prior to auxiliary feedwater initiation, the damaged steam generator can be easily identified by a rapidly decreasing steam generator pressure and an abnormal level response (high or low).
In addition, if the FWRV and/or FWRV bypass valves have not performed their post-trip function (FWRV closes and FWRV bypass valves open to a pre set position), excess flow would be observed in the loop containing the damaged steam generator.
It should be noted that the analyses in References 3, 4 and 5 include the effects of feedwater flow to the damaged steam generator and that isolation would only be required if the FWRV and FWRV bypass valve failed to perform their functions. In the longer term, post auxiliary feedwater initiation, without operator action, essentially all auxiliary feedwater would be directed,to the damaged steam generator. The results of analyses presented in References 3, 4 and 5 demonstrate that continued auxiliary feedwater flow (following a five minute delay to initiate) to the damaged steam generator does not present a concern for either containment overpressurization or return to power. For decay heat removal from the primary system, the operator would be required to isolate the auxiliary feedwater line feeding the damaged steam generator in order to provide make-up to the intact steam generators, particularly if the main feedwater system were not available. The key indications for directing flow to the intact steam generators would be excess flow in the loop with the damaged steam generator and level in the intact steam generators. An abnormal level and/or pressure would be observed in the
United States Nuclear Regulatory Commission May 7, 1980 Page 3 damaged steam generator versus the intact steam generators. Guidance to the operator for maintaining secondary heat sink by isolating the damaged steam generator is provided in the Maine Yankee emergency procedures.
In the event of a steam-line rupture upstream of the excess flow check valve (EFCV), it is assumed that the operator isolates flow to the rupture stream generator within 10 minutes. As a result, the auxiliary feed pumps ( AFW) may experience cavitation due to pump runout for a period 5 minutes. Maine Yankee has evaluated the effects to the auxiliary feed pump operating under these runout conditions and concludes that there will be no consequential loss of safety function capability.
At Maine Yankee, if the auxiliary feed pump is operating at runout conditions while discharging to a depressurized steam generator, moderate cavitation is expected to occur at the eye of the impeller and along the trailing edge of the impeller vane. Without pre-heat, there is no possi'ility of forming and sustaining large voids in the suction pipe and losing pump suction as a result. Since the pump is cooled by the water pumped, there is no threat of overheating. As the cavitational voids increase in size, the problem becomes self-correbting, because there is a rapid drop in pump efficiency, or flow, which in itself eliminates the voids. The result can be a surging flow condition but not a loss of flow so long as the water pumped is cold. The effects of surging and collapsing voids are not expected to cause damage since these forces are significantly less than the design capabilities of a boiler feed pump which, according to the manufacturer, are experienced at shut-off conditions.
In conclusion, if the MY AFW pumps operate at runout; we expect noisy operation, a fall off of pump performance but no damage to the pump.
Q2.
Review your analysis of the reactivity iricrease which results from a main steam line break inside or outside containment. This 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 greater than previous analysis indicated, the report of this review should include:
a.
The boundary conditions for the analysis, e.g.,
the end 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, c.
The effect of extended water supply to the affected steam generator on the core criticality and return to power,
United States Nuclear Regulatory Commission May 7, 1980 Page 4 d.
The hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the analyzed transient.
A2.
The revised steam line break analysis submitted in Reference 5 included the effects of automatic initiation of auxiliary feedwater and continuation of main feedwater flow through the main feedwater regulating valve bypass valve. No return to power is predicted to occur for either condition. The Reference 5 analysis did not include the continuation of main feedwater flow to the affected steam generator through an open MFWRV or MFWRV bypass valve should either fail to respond to its post-turbine trip position. Continued feedwater addition in either mode would result in a return to criticality due to the excessive reactor cooldown and negative moderator temperature coefficient at end of cycle.
As previously described in the response to Item 1, emergency procedures are being revised to direct the operator to trip all pumps in the main feeedwater system should the MFWRV and/or MFWRV bypass valve fail to close or respond to their post-trip positions.
Q3 If the potential for containment overpressure exists or the reactor return-to-power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action.
If the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action is completed.
A3 Failure of the MFWR7 and/or MFWRV bypass valve to close or respond to their post turbine trip positions results in the potential for overpressurization of the containment and/or return to criticality following a main steam line break. As a result, the following design change has been initiated as a corrective action with implementation scheduled around June 1, 1981.
The design change would provide a safety grade closure signal from the low steam generator pressure excess flow check valve (EFCV) closure signal to both the MFWRV and the MFWRV bypass valves of the affected steam generator. This would isolate all main feedwater flow to the break. Redundancy would be provided by a safety-grade signal to trip all pumps in the main feedwater system (MFW, condensate, and heater drain pumps) on receipt of a coincident SIAS and any low steam generator pressure EFCV closure signal. In conjunction with these changes, the auxiliary feedwater system would be modified to provide a safety grade closure signal from the low steam generator pressure EFCV closure signal to the associated auxiliary feedwater flow control valve in order to direct flow to the intact steam generators. These changes would prevent containment overpressurization and return to criticality for any steam line break transient.
As an interim measure, an additional closure signal from a safety-grade source, the low steam generator pressure 9FCV closure singal, has been provided to the E/P converters controlling the MFWRV and MFWRV bypass
United States Nuclear Regulatory Commission May 7, 1980 Page 5 valves, as a back-up to the turbine-trip override signal. This signal will close both the MFWRV and the MFWRV bypass valve associated with the affected steam generator following a steam line break.
Interim action that will be implemented as soon as practicable involves upgrading the emergency procedures to direct the operator to trip the main feedwater pumps and close the main feedwater MOVs in the event that the MFWRV or MFWRV bypass valves fail to close or respond to their post-trip positions. Operator action to isolate main and auxiliary feedwater to a broken steam generator and direct feedwater flow to the intact steam generators is already included in the Maine Yankee emergency procedures.
We trust you will find this submittal satisfactory; however, should you desire additional information feel free to contact us.
Respectfully submitted, MAINE YANKEE ATOMIC POWER COMPANY y
D. E. Vandenburgh Vice President
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