ML19340C526
| ML19340C526 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 10/28/1980 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Nichols T SOUTH CAROLINA ELECTRIC & GAS CO. |
| References | |
| NUDOCS 8011170774 | |
| Download: ML19340C526 (10) | |
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UNITED STATES
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.., 'i NUCLEAR REGULATORY COMMISSION
. f,f.,'C WASHINGTON, D. C 20555
%j 5.s, OCT 2 81980 Docket No. 50-395 Mr. T. C. Nichols, Jr.
Vice President and Group Executive Nuclear Operations South Carolina Electric and Gas Company P. O. Box 764 Columbia, South Carolina 29281 Dea. Mr. Nichols:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION As a result of our meeting on October 8,1980 we have resolved 22 of the 33 open issues in the Reactor Systems Branch review of your application for an operating license for Virgil C. Summer Nuclear Station, Unit 1.
We are still reviewing your response to Branch Technical Position RSB 5-1.
The remaining 10 open issues are addressed in the enclosed nine requests for additional information, 211.124 thru 221.132. These requests were discussed with your representatives at the October 8, 1980 meeting.
Your responses ;hould be provided not later than October 31, 1980.
Please contact the staff's assigned project manager if you require any clarification.
Sincerely, i, (
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- Y cA c 2 Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
Request for Additional Information cc: See next page 80212707 79
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I Mr. T. C. Nichols, Jr.
Vice President & Group Executive OCT 2 8 79g9 1
l gr Nuclear Operations South Carolina Electric & Gas Company P. O. Box 764 Columbia, South Carolina 29281 cc: Mr. William A. Williams, Jr.
Vice President South Carolina Public Service Authority 223 North Live Oak Drive Moncks Corner, South Carolina 29461 J. B. Knotts, Jr., Esq.
Debevoise & Liberman 1200 17th Street, N. W.
Washington, D. C.
20036 Mr. Mark B. Whitaker, Jr.
Group Manager - Nuclear Engineering & Licensing South Carolina Electric & Gas Company P. O. Box 764 Columbia, South Carolina 29218 Mr. Brett Allen Bursey Route 1, Box 93C Little Mountain, South Carolina 29076 Lasident Inspector / Summer NPS c/o U. S. NRC Route 1, Box 64 Jenkinsville, South Carolina 29065 4
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ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION 211.124 Overpressurization of Reactor Vessel at Low Temoerature and Pressure I
For protection of overpressurization of the reactor vessel at low temperature and pressure you have provided a seismically qualified nitrogen (N ) supply system to serve each of the PORVsjthe system is j
2 i
sized to assure that no operator action is required to terminate the i
transient in 10 niinutes. Provide a justification for this 10 minute limit and why it is enough for the operator to identify and terminate l
the cause of the transient.
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211.125 Identification of Indicators and Alarms Provided in the control Room for Leakage Detection Provide a table of all indicators and alarms in the control room associated with leak detection instrumentation for all three types of leak detectors.
i 211.126 Loss of CVCS or CCW to Reactor Coolant Pumps or Motors In response to request 211.123 concerning loss of CVCS or CCW to reactor coolant pumps you stated that two RCP motors have been tested with interrupted CCW flow, and-that the test demonstrates that the RCP motor can withstand loss of CCW flow for 10 minutes without pump r
damage. Verify that the loss of CCW to both RCP motor bearings and the thennal barrier heat exchanger will not have a worse effect on the i
l RCP than only the result of the lost CCW to the pump motor bearings as simulated in your test. Also, provide a sununary of your pump tests.
e l 221.127 Overpressurization of Internal Body Cavity of Gate Valves in the ECCS System We have been notified of a potential design deficiency regarding double seating gate valves which are used in the ECCS systems of some PWR plants. The concern is that when fluids, trapped in the internal body cavity of the valve, are heated due to the increased temperatures of adjacent piping systems or of the environment, sub-stantial pressure increases may result in these cavities that could i
rupture the valve.
Provide information which addresses this poten-tial valve problem as it applies to the Virgil C. Summer Station.
211.128 Credit for Operator Action Your response to requests 211.108 and 211.120 have only identified three events that require operator manual action to mitigate the consequences of an accident. The response should be expanded to specifically identify the need 2nd the time for operator action for each Chapter 15 event.
211.129 Submittal of Revised LOCA Analyses with Corrected Metal '..'ater Reaction.
And, Additional Small Break Analyses to Insure Identification of Worst Case Small Break You have recently revised the input to the small break LOCA mouel.
This revision resulted from a QA audit which uncovered an input error in modeling the reactor vessel downcomer. The correction reduced the area 2
of the downcomer by a factor of 2, from 52 to 26 ft. This input correction resulted in a predicted peak clad temperature decrease of 125'F for the 3-inch break (1833 to 1708 F). The staff is presently evaluating this modification. However, we require that you formally document the corrections made to your small break LOCA model, and
. revise the analyses presented in the FSAR.
In addition, you should i
discuss why the limiting small break size is not less than 3 inches in diameter (yet greater than 2 inches, which is the size capable of being mitigating by the charging pumps along).
211.130 Isolation of Lines Between MSIV and Turbines Stop Valves on ESFAS Table 10.3-3 of your FSAR indicated that several steam line valves downstream of the MSIVs will remain open on an ESFAS.
Confirm that the assumed steam release from unaffected steam generators following a main steam line break accident as listed in Table 15.4-23 i
has included the steam released from the open valves identified in i
i Table 10.3-3 and the steam supply to the turbine driven auxiliary feedwater pump.
211.131 Analyses of Baron Dilution Events from Hot Standby ar.d Cold Shutdown a.
It is required by the Standard Review Plan that you analyze unplanned boron dilution events. Since the sequences of events that may occur depend on plant conditions at the time of the unplanned moderator I
dilution, the analyses should include conditions at the time of the unplanned dilution, such as refueling, startup, power operation, hot standby and cold shutdown, i
Your Chapter 15 analyses did not include analyses of hot standby and cold shutdown. He request that you include theseanalyses in your FSAR.
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l b.
What are the assumed causes of an unplanned reactivity insertion j
during refueling, startup, and at power? What are the necessary i
l actions to be taken by the operator to mitigate each of these events?
't Identify the actions to be taken by the operator in the event of the worst single failure postulated in the mitigating system, and show that the time available to the operator to mitigate the event including the effects of the single failure is sufficient.
211.132 Containnent Sump and its effect on long term coolino followina a LOCA During our reviews of license applications we have identified concerns related to the containment sump design and its effect on long term cooling following a Loss of Coolant Accident (LOCA).
i These concerns are related to (1) creation of debris which could i
potentially block the sump screens and flow passages in the ECCS and i
the core, (2) inadequate NPSH of the pumps taking suction from the 1
containment sump, (3) air entrainment from streams of water or steam which can cause loss of adequate NPSH, (4) fonnation of vortices which can cause loss of adequate NPSH, air entrainment and suction i
l of floating debris into the ECCS and (5) inadequate emergency pro-I j
cedures and operator training to enable a correct response to these probl ems.
Preoperational recirculation tests performed by utilities j
have consistently identified the need for plant modifications.
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The NRC has begun a generic program to resolve this issue.
- However, more immediate actions are required to assure greater reliability of safety system operation.
We therefore require you take the following actions to provide additional assurance that long term cooling of the reactor core can be achieved and maintained following a postulated LOCA.
i 1.
Establish a procedure to perform an inspection of the containment, j
and the containment sump area in particular. to identify any materials which have the potential for becoming debris capable of blocking the containment sump when required for recirculation of coolant water.
Typically, these materials consist of: plastic bags, step-off-pads, health physics instrumentation, welding equipment, scaffolding, metal chips and screws, portable inspection lights unsecured wood, construction materials and tools as well as other miscellaneous loose equipment.
"As licensed" cleanliness should i
be assured prior to each startup.
1 This inspection shall be performed at the end of each shutdown as j
soon as practical before containment isolation.
2.
Institute an inspection progran according to the requirements of Regulatory Guide 1.82, item 14. This item addresses inspec-i tion of the containment sump components including screens and i
intake structures.
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Develop and implement procedures for the operator which address both a possible vortexing problem (with consequent pump cavitation) i and sump blockage due to debris. These procedures should eddress i
all likely scenarios and should list all instrumentation available j
to the operator (and its location) to aid in detecting problems which may arise, indications the operator should look for, and i
2 operator actions to mitigate these problems, i
l 4.
Pipe breaks, drain flow and channeling of spray flow released below l
1 or impinging on the containment water surface in the area of the sump can cause a variety of problems; for example, air entrainment, i
cavitation and vortex formation.
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l Describe any changes you plan to make to reduce vortical flow in the l
neighborhood of the sump.
Ideally, flow should approach uniformly from all directions.
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5.
Evaluate the extent to which the containment sump (s) in your plant meet the requirements for each of the items previously identified; namely debris, inadequate NPSH, air entrainment, vortex formation, and operator actions.
The following additional guidance is provided for performing this i
j evaluation.
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l 1.
Refer to'the recommendations in Regulatory Guide 1.32 (Section C)
I which may be of assistance in performing this evaluation.
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. 2.
Provide a drawing showing the location of the drain sump relative
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to the containment sumps.
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3.
Provide the following information with your evaluation of debris:
a.
Provide the size of openings in the fine screens and canpare this with the minimum dimensions in the pumps which take suction from the sump
, the minimum dimension in any spray nozzles and in the fuel assemblies in the reactor e
core or any other line in the recirculation flow path whose size is ccmparable to or smaller than the sump screen mesh size in order to show that no flow bloc'Kage Will occur at any point past the screen.
b.
estimate the extent.to which debris could block the trash I
r rack or screens (50 percent limit).
If a blockage problem is identified,, describe the corrective actions you plan to take (replace insulation, enlarge cages, etc.).
c.
For each type of thermal insulation used in the containment,
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provide the following information:
i, type of material including composition and density, ii. manufacturer and brand name, j
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111. method of attachment,
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location and quantity in containment of each type, i
v.
an estimate of the tendency of each type to form particles
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small enough to pass through the fine screen in the suction j
lines.
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j d.
Estimate what the effect of these isulation particles would a
be on the operability and performance of all pumps used for recirculation cooling. Address effects on pump seals and j
bearings.
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Additionally, previous in-plant sump tests did not accurately replicate l
expected post-LOCA conditions, and thus did not demonstrate acceptable
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sump performance under ECCS recirculation conditions. Specifically, l
the plant test only pulled suction from a single line, when there are two lines in each of two sumps. This resulted in approach flow velocities which were lower than would be expected during a LOCA.
Additionally, various flow approach directions were not. investigated to determine if undesirable rotation could be induced in the sump i
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area, which could lead to vortex formation.
Finally, sump screen blockage due to debris entrainment was not considered, with the correspondingly higher screen velocities which 1
l also could aggrevate vortex formation.
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