ML20134C862
| ML20134C862 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/28/1997 |
| From: | Reid D VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BVY-97-17, GL-96-06, GL-96-6, NUDOCS 9702040097 | |
| Download: ML20134C862 (6) | |
Text
..
.VSRMONT YANKEE l
NUCLEAR POWER CORPORATION e
Ferry Road, Brattleboro. VT 05301-7002 i
f ENGINE IN OFFICE 580 MAIN STREET BOLTON, MA 01740 (508) 77w-6711 January 28,1997 BW 97-17 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
References:
(a)
License No. DPR-28 (Docket No. 50-271)
~(b)
USNRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment integrity During Design-Basis Accident Conditions," NW 96-151, dated September 30,1996 (c)
Lotter, WNPC to USNRC, BW 96-136, dated October 30,1996 (d)
Letter, WNPC to USNRC, " Amendment No. 22 to License App:lcation dated j
December 22,1996" dated February 25,197 'a (e)
GOTHIC Analysis of a Containment Fan Ccp%r Unit Under LOCA & LOOP Conditions, EPRI, May 1996 (f)
Waterhammer Prevention, Mitigation and Accommodations, EPRI NP-6766, Volume 5, Part 1, July 1992 i
Subject:
Vermont Yankee 120-Day Response to Generic Letter 96-06 i
in Reference (b) NRC notified licensees about safety-significant issues that could affect containment integrity and equipment operability during accident conditions. The NRC requested licensees to evaluate the susceptibility of their plant systems to the identified conditions and report within 120 days the results of their evaluation. The purpose of this letter is to provide the requested information. provides a summary of our evaluation of the information in Reference (b). In pedorming this evaluation, we considered Vermont Yankee's postulated accideat conditions and the scenarios referenced in the Generic Letter. Where appropriate. we have also addressed the operability of related systems.
We trust that the information provided is acceptable. However, should you have any questions or require additional information, please contact this office.
Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION 8
Donald A. Reid
[L7 M p
Vice President, Operations 9702040097 970128 PDR ADOCK 05000271
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United States Nucle:r Rigulatory Commiss!MERMONT YANKEE NUCLEAR POWER CORPORATION January 28,1997 Page 2 of 2 l
Attachment
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USNRC Region 1 Administrator "2003 h
USNRC Project Manager - WNPS
,4 USNRC Resident inspector - WNPS y
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NOTARY
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-s STATE OF VERMONT
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4 PllBLIC k,
WINDHAM COUNTY O
Operations, of Vermont Yankee Nuclear Power Corporation, that he is duly authorized to ex
.yT[c4 Msident, Then personally appeared before me, Donald A. Reid, who, being duly sworn, di
_fds;#dfforegoing if* rsentstherein are docurnent in the name and on the behalf of Vermont Yankee Nuclear Power Corporation, and that the e
true to the best of his knowledge and belief.
Sally A. SIandstru'm, Notary Public My Commission expires February 10,1999
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United Stat s Nucle:r R:gulatory CommissidelERMONT YANKEE NUCLEAR POWER CORPORATION January 28,1997 l
' Attachment 1 ~
Page 1 of 4 i
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Vermont Yankee's 120-Day Response to NRC Generic Letter 96-06:
i Assurance of Equipment Operability and Containment integrity During
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Design-Basis Accident Conditions i
Raouested information i
i Within 120 days of the date of this generic letter, addressees are requested to submit a written summary report stating actions taken in response to the requested actions noted above, i
conclusions that were reached relative to susceptibility for waterhammer and two-phase flow in t
l the containment air cooler cooling water system and overpressurization of piping that penetrates l
containment, the basis for continued operability of affected systems and components as j
applicable, and corrective actions that were implemented or are planned to be implemented. If l
systems were found to be susceptible to the conditions that are discussed in this generic letter, identify the systems affected and describe the specific circumstances involved.
Response
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Executive Summary Vermont Yankee has completed a review of the information provided in Generic Letter 96-06 and performed an assessment of the susceptibility of Vermont Yankee Nuclear Power Station (VYNPS) to the conditions described therein. This assessment considered not only Vermont Yankee's postulated accident conditions but also the scenarios referenced in the Generic Letter. As a result of our assessment, we have determined that: (1) VYNPS is not susceptible to over pressurization of isolated piping since over-pressure protection was installed during the Fall 1996 refueling outage to address this concern, and (2) the section of our Reactor Building Closed Coo!ing Water (RBCCW) system that i
provides cooling water to our containment coolers is potentially susceptible 1o waterhammer conditions.
Vermont Yankee has determined that the RBCCW system is operable, but will evaluate any necessary long-term actions in accordance with our Corrective Action program.
Backaround j
The containment air coolers at Vermont Yankee are supplied by cooling water from the RBCCW system. The cooling water enters the drywell through a single penetration and is then distributed to the four containment cooling units, the cooling coils for each recirculation pump, and the drywell equipment drain sump cooling coll. The water exits the drywell through another single penetration.
The RBCCW system also supplies cooling water to other components and systems within the Reactor Building. The cooling water is circulated by one pump, with a redundant pump in a standby mode
'which starts automatically on low system pressure. None of the equipment served by the RBCCW system is required to function following a Design-Basis Accident (DBA). A more detailed description of the RBCCW system can be found in the Vermont Yankee Updated Final Safety Analysis Report, Section 10.9.
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United St:tes Nucle:r Rigulat:ry C:mmissidelERMONT YANKEE NUCLEAR POWER CORPORATION I
Janu:ry 28,1997
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4
- Attachment 1 i
Page 2 of 4 j
The four containment air coolers have been upgraded from the original design and are designated as RRUs 1,2,3 and 4. Each RRU has two fans located in common ducting, so that either or both fans will draw air across the common cooling coll. The RRUs at Vermont Yankee do not perform a design-basis post-accident containment cooling function. In fact, an accident signal (high drywell pressure or low reactor water level with low reactor pressure) will trip the fan motors.
i Cooling water circulates continuously through each RRU during normal operation. Power for the RBCCW pumps is from 480 Volt buses that receive emergency electrical power from the standby diesel generators. In the event normal electrical power is lost, the RBCCW pumps are automatically restarted 4
60 seconds after emergency power is restored to the bus. Under DBA conditions, the RBCCW pump i
would restart approximately 73 seconds after the initiation of the accident. However, as previously stated, the RBCCW System is not required to mitigate the consequences of a DBA. The pumps are i
automatically restarted upon restoration of ernergency power as an operator convenience and to minimize the potential negative impact of the loss of cooling water flow to non-essential systoms.
l Discussion i
Thermal Overpressurization Generic Letter 96-06 reported the potential for an overpressurization of isolated piping segments due to the heat up and subsequent thermal expansion of entrapped fluid. This issue was resolved by Vermont Yankee during the Fall 1996 refueling outage with the Installation of pressure relief devices in piping determined to be susceptible to thermal overpressurization. Vermont Yankee installed relief valves in the RBCCW and Radwaste systems and check valves in the Main Steam Drain, RHR Shutdown Cooling, and Nuclear Boller (Sample) systems.
The relief valves were installed in, and relieve into, the primary containment. The check valves relieve overpressure to the upstream side of each respective inboard isolation valve to the reactor pressure vessel. The rate of thermal expansion of isolated fluid, relief requirements as well as the effects of stuck-open relief valves and associated environmental flooding and radiation hazards were considered.
The new valves were, and will continue to be, tested in accordance with 10CFR50, Appendix J.
Two Phase Flow and Waterhammer Generic Letter 96-06 provides a description of circumstances at other nuclear plants where the generation of two phase flow conditions in containment air coolers had the potential to impact the design-basis heat removal capability of the containment coolers. The primary concern identified was i
the degradation of cooling water flow and design-basis heat removal capability. Since Vermont Yankee does not rely on its containment coolers or other RBCCW-cooled components for design-basis heat removal, the generation of two phase flow in the containment cooling piping is not a concern ht Vermont Yankee in the context of heat removal degradation.
Generic Letter 96-06 also provides a description of circumstances at other nuclear plants which concluded that during a design-basis LOCA with concurrent loss of offsite power, cooling water in the containment fan coolers could flash to steam. Subsequent restart of the cooling water pumps could result in rapid collapse of the steam volds, resulting in a waterhammer induced loading condition that could challenge the integrity and function of the fan coolers and the component cooling water system.
I Both of the nuclear units discussed in Generic Letter 96-06 depend on the fan coolers for post accident containment heat removal.
p United Stat:s Nucirir R:gulat:ry CcmmissiaMERMONT YANKEE NUCLEAR POWER CORPORATION Janugry 28,1997
' Attachment 1 Page 3 of 4 l
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The containment cooling units at Vermont Yankee do not perform a design-basis post accident containment cooling function. However, the portion of the RBCCW system that is within the drywell is considered a closed system in accordance with 10CFR50, Appendix A, General Design Criterion 57.
This part of the system is considered an extension of the containment and as such is maintained as 1
a seismic Category 1, Safety Class 2 system. The RBCCW piping inside the drywell is credited as one of the two required barriers; the isolation valves on the supply and return lines constitute the second barrier. Therefore, a waterhammer induced load has been evaluated at Vermont Yankee because of its potential to affect the integrity of one of the two containment barriers.
Vermont Yankee determined that for boiling to occur in the coils the temperature of the cooling water I
must reach 272*F. The driving force for heat transfer to the cooling water is the ambient condition in the drywell following a LOCA. The Design-Basis LOCA, a guillotine break of a recirculation line l
(UFSAR Section 14.6.3.3) with a coincident loss-of offsite-power (LOOP), causes drywell temperature to exceed 272'F for approximately 15 seconds (UFSAR Figure 14.6-6). Drywell temperature drops 3
i below 272* F 16 seconds into the event and remains below 272* F thereafter. A heat transfer study has I
shown that the cooling water temperature will lag the increase in containment temperature. Since the j
containment temperature will be decreasing below 272'F when the cooling water temperature reaches j
saturation, bolling of the containment cooling water will not occur. Therefore, for the DBA LOCA there is reasonable assurance that waterhammer will not occur when RBCCW flow through the containment j
coolers is restored 73 seconds into the event.
i Vermont Yankee also considered a main steam line break (MSLB) with a coincident I.OOP. Reference J
2 (d reported the analysis results of a number of steam line t' eaks ranging in size from 0.02 ft to 0.5 j
ft}.
The analysis demonstrated that drywell temperature would remain below the saturation temperature of the containment cooling water for the first 73 seconds of the accident. Therefore, for this spectrum of breaks, containment cooling water flow would be restored prior to the onset of void j
formation within the containment coolers, providing reasonable assurance that waterhammer would not l
occur for these scenarios.
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Vermont Yankee does not have a plant specific analysis for MSLB >0.5 ft and therefore has conservatively assumed that waterhammer could occur for such an event. In its analysis of i
waterhammer, EPRI postulates that the loads associated with the reintroduction of cooling water flow would be mitigated by the presence of hot water on either side of the void region [ Reference (e)]. That is, the hot water would slow down the condensation rate of the steam volds and thus reduce the l
pressure differential for accelerating flow. Additionally, the pressure in the steam region would not i
i drop below the saturation pressure of the heated water. Thus, a steam void would remain to mitigate waterhammer impact loads.
In attempting to quantify the anticipated waterhammer loads, Vermont Yankee has used EPRI guidelines, Reference (I), for estimating loads for water filling a voided line or water column rejoining.
The estimated load relative to the water column velocity is between 25 psi per 1 ft/sec and 50 psi per 1 ft/sec. The water velocity in the cooling coil t@es during normal operation is about 4 ft/sec. The estimated loads would thus be between 100 and 200 psi. If the water slug is assumed to be accelerated by the steam bubble collapse, the impact velocity would increase relative to the initial velocity. Again from Reference (f), if the initial column length in the cooling coil tube is at least 1% of the total length, the velocity could be increased to 1.5 times the initial velocity, in this case, the loads would be between 150 and 300 psi. The containment cooling coils at the VYNPS were pneumatically tested at 200 psig and hydro tested at 300 psig. Therefore, the expected waterhammer induced loads on containment cooling piping are within piping design limits and operability is maintained.
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United Stit s Nucle:r RIgulatr y CommissohfERMONT YANKEE NUCLEAR POWER CORPORATION
, Janupry 28,1997 Page 4 of 4 Conclusion Vermont Yankee has installed modifications that will preclude everpressurization of isolated piping at VYNPS and has concluded that there is reasonable assurance that waterhammer will not occur upon restoration of containment cooling following a DBA LOCA with a coincident LOOP. Additionally, in the event of a steam line break inside containment (MSLB) Vermont Yankee has determined that containment operability would also be maintained. Vermont Yankee will continue to follow industry activities in this area and to refine its analyses to better quantify the likelihood of waterhammer and the severity of its effects on the containment cooler piping.
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