ML19340E900
| ML19340E900 | |
| Person / Time | |
|---|---|
| Site: | Crane, Saint Lucie  |
| Issue date: | 01/14/1981 |
| From: | Cutchin J NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | METROPOLITAN EDISON CO. |
| Shared Package | |
| ML19340E901 | List: |
| References | |
| NUDOCS 8101160167 | |
| Download: ML19340E900 (7) | |
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. STAFF 01/14/81 O
UNITED STATES OF AMERICA NUCLEAR REGULATORY Cr$11SSION BEFORE THE ATG11C SAFETY AND LICENSING BOARD G
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Docket No. 50-289
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NRC STAFF'S RESPONSES TO LICENSEE'S TEN INTERR0GATORIES ON REACT;JR VESSEL WATER LEVEL OF DECEMBER 23, 1930 _
Attached are the NRC Staff's responses to Licensee's ten interrogatories on reactor vessel water level, transmitted by the Licensee on December 23, 1980.
These responses were prepared by L.E. Phillips of the NRC Staff whose affidavit is also attached.
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, Ja.. s M. Cutchi., IV Counsel for NRC Staff January 14, 1991 8102260%l g
FET-ED Interrocatory 1 The staff states that the saturation meter does not distinguish between anomolous transients which can drain the pressurizer and cause primary loop saturation due to cooling and shrinkage of primary coolant versus a loss of coolant inventory which could lead to inadequate core cooling.
k' hat need is there for the operator to distinguish between such transients?
Describe any differences in opercior action which would be necessary in responding to these two types of transients.
USNRC Pesponse to MET-ED Interroaatory 1 For an anomolous overcooling transient, the operator should determine the cause of the overcooling and correct it.
For a small break LOCA, the primary system will continue to lose coolant inventory, at a rate and duration dependent on the size and location of the break, until the safety injection make-up flow exceeds the rate of coolant loss.
For some conditions, the time interval from the instant of primary system saturation conditions until the occurrence of super-heat indication on the core exit thermocouples or hot leg RTDs is in excess of 30 minutes.
The superheat condition does not occur until the core is partially uncovered and fuel heat up has begun.
TMI-l Emergency Procedure 1202-6B describes the different operator responses to small break LOCA versus overcooling events which cause automatic high pressure injection. These procedures n'ow require the operator to distinguish between the transients based on indirect indicators from existing instrumentation. Vessel level instrumentation, if available, would permit a much quicker and more reliable diagnosis of the conditions.
For small break LOCA, an orderly cooldown is required, but not necessarily for an overcooling transient.
In both cases, a vessel level meter if available, would provide coordinating information to assist the operator in restoring the water solid primary system (possibly using the upper head vent) and the normal water level in the pressurizer.
The licensee has not submitted any analyses or evaluation addresssing additional operator actions which could be _taken to prevent core uncovery for small break LOCAs if level information were available for prompt i
diagnosis of the condition.
For example, the desirability of various l
operator actions to increase safety injection make-up flow when a slow but continuous loss of coolant inventory is indicated has not been considered.
One such action may be rapid 0TSG depressurization to 400 psig which would tend to increase the cooling rate of the primary system, thus depressurizing l
to make the accumulator coolant injection available.
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MET-ED Interrogatory 2 The staff states that core exit thermocouples do not detect the advent of inadequate core cooling until the level has dropped into the core and fuel heatup has begun. Define " inadequate core cooling" as used by the Staff in its statement, i.nd relate the definition to the limits of 10 CFR 50.46.
USNRC Response to HET-ED Interrogatory 2 Both the staff and the industry have varied definitions of inadequate core cooling. A staff definition is given on page 3 of the evaluation enclosed with our letter of September 24, 1980 to the licensee, and which has been previously filed as testimony at the hearing. The staff considers inadequate core cooling to exist when the fuel cladding tempcrature exceeds the worst case values predicted for the LOCA small break spectrum using an evaluation model which meets the requirements of Appendix K to 10 CFR 50.
This condition is normally more restrictive than the 10 CFR 50.46 limits (2200*F PCT) which are approached in the LOCA large break analyses.
Inadequate core cooling, as used in the statement, is referring to the occurrence of an anomalous condition of core uncovery whereby the fuel clad temperature is increasing in an uncontrolled heat up.
This condition also occurs for analyzed accident transients, namely Loss of Coolant Accidents (LOCA) of break size and location which result in core uncovery.
For a stylized LOCA transient, i.e., a transient whereby the system response is consistent with assumptions in the accident analyses, inadequate core cooling may occur but automatic systems will respond to, prevent the fuel cladding from exceeding the limits of 10 CFR 50.46 (2200F PCT, etc.).
Note that the B&W Owner' ; Group, in B&W Document No. 86-1120838-00, issued August 15, 1980, defines inadequate core cooling as follows:
"In a depressurization event, the reactor coolant system (RCS) must first reach saturation conditions before there is any danger of inadequate core cooling.
Subsequently if the RCS inventory is reduced and uncovery of the core begins, temperatures in the uncovered region will increase causing super-heating.
It is important to note in this discussion that inadequate core cooling does not begin until reactor vessel (RV) water inventory falls below the top of the core thus resulting in an increasing fuel clad temperature."
Using the Owners's Group definition, Core exit thermocouples do not detect inadequate core cooling uritil it already exists.
MET-E9 Interrogatory 3 The Staff states that if level instrumentation were available, the effect-iveness of HPI in recovering the system would provide valuable diagnostic l
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information on the nature of the transient before the level drops into the core. What additional actions would the operator take, upon evaluating the effectiveness of HPI via level indication, apart from verifying that:
the HPI pumps are operating; the valves are open; and design basis HPI flow is present?
USNRC Response to MET-ED Interrooatory 3 The NRC staff has not evaluated the possible actions the operator could take based on core level instrumentation but has required that the licensees describe how emergency (procedures should be modified after level instrum is incorporated (Item 7) and (8) of Documentation Required in II.F.2 of NUREG-0737).
MET-ED Interrogatory 4 Phat is the Staff's basis for the statement that the trend of level-indicatien would provide valuable diagnostic information on the nature of the transient before the level drops into the core? Would not a large LOCA also result in core uncovery?
USf!RC Pesponse to MET-ED Interrogatory 4 Pater level indication vould provide positive and direct information that a LOCA condition exists and, based on the rate of loss of coolant inventory, wculd provide early indication whether core uncovery is likely. Without level instrumentation, there is no warning of core uncovery until the core has partially uncovered and superheat is indicated on the core exit thermocouples. As indicated in the response to Interrogatory No.1, the water level warning could precede the thermocouple warning by as much as thirty minutes. In fact, more complete small break spectrum data on other reactors indicate that time intervals up to three hours of continuous inventory loss could exist before core uncovery is detected by superheat conditions.
Pater level information would also provide indication when safety injection makeup is adequate to prevent core uncovery and would provide pnsitive information when primary system inventory recovery commenced.
Pater level instrumentation is not expected to be of value during the blowdown phase of a large break LOCA. Core uncovery for the large break condition occurs in a much shorter time interval, on the order of one minute or less, which is insufficient time for the operator to evaluate and take actions based on level information.
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With respect to diagnostics, the primary coolant inventory versus time data could be compared to event scenarios considered in the safety analyses to infer the type and location of the break, or to provide evidence of anomalous behavior which could otherwise be undetected.
MET-ED Interrogatory 5 What information for diagnosing the effectiveness of core recovery actions would be provided by water level indication that is not provided by hot leg temperature, cold leg temperature, reactor coolant system pressure and temperature saturation margin? What would be done differently by the operator based on such information?
USNRC Respcnse to MET-ED Interrogatory 5 Coolant inventory information based on water level indication would provide direct evidence concerning the adequacy of the safety injection flow to the primary system.
The level indication would also provide evidence that the core is covered during recovery from a TMI-2 type flow blockage ccndition, even though superheat may persist at the core exit thermocouples.
None of the cited process parameters would provide equivalent information on a continuous basis.
The licensee has not yet provided a description of proposed level instrumentation and the supporting analyses and procedures based on level information for the staff to evaluate.
MET-ED Interrogatory 6 Describe the " unsafe operator actions" the Staff alleges could have occurred during the June 11, 1980 St. Lucie event, and provide the post-event evaluations upon which this statement is based.
USNRC Response to MET-ED Interrogatory 6 The statement referenced is based on the report by E. V. Imbro, Office for Analysis and Evaluation of Operational Data, subject:
" Report l
on The St. Lucie I Natural Circulation Cooldown on June 11, 1980."
A copy of this report is attached.
The staff has not evaluated the type and consequences of unsafe actions which might have been taken because the plant response to l
this event (the formation of a steam bubble in the upper head during l
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. natural circulation cooldown) initially puzzled the operators. However, one mis-operation associated with the confusion caused by the event was the restart of a Reactor Coolant pump without adequate seal cooling.
MET-ED Interrogatory 7 Describe how vessel level indication could prevent the unsafe operator actions described in your response to Question 6 above.
USNRC Response to MET-ED Interrogatory 7 Vessel level information would have indicated the void formation in the upper head and thus alleviated any concern about water - solid operation.
Eetter understanding of the event would have prevented operator confusion which contributes to the likelihood of unsafe actions. Additionally, vessel level indication would provide a safe bases for manual shut off of HPI to avoid overflow to containment.
MET-EC Interrogatory 8 Describe the basis for the statement that vessel level indication is "important and possibly essential" to proper use of the reactor vessel head vent.
In particular, what actions would the operator take based upon vessel level indication which would be different than actions based upon information from available alternate instrumentation?
USNRC Response to MET-ED Interrogatory 8 Vessel level information would indicate the existence of a void in the upper head so that the need for vessel venting could be evaluated.
Vessel level indication should provide a direct indication of the voided condition so that unnecessary opening of the vessel head vent could be avoided.
The NRC staff has not evaluated the conditions for which the head vent should be opened and has requested that procedures for use of the vent be provided by the licensee. However, the staf f is aware of at least one vent design for which the designer insists on the coupling of level infcrmation to the safe operation of the vent.
MET-ED Interrogatory 9 Has the Staff evaluated how water level varies with time for a spectrum of LOCA's, and determined at what point or under what conditions the operator should take new or different actions based on water level?
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USNRC Response to MET-ED Interrogatory 9 The NRC Staff has evaluated the variation of the reactor vessel two-phase level with time for the spectrum of small break sizes for the purposes of determining compliance with 10 CFR 50.46. The NRC staff has also evaluated the variation of the reactor vessel level for certain smaller break sizes that were used to establish smali break LOCA Guidelines.
These evaluations are discussed in the NRC response to UCS contention 8.
The NRC staff '-is not evaluated the variation of water level with time for the purpose Jf determining the level meter response on the actions to be taken by the operator. However, some actions wh,ich should be considered by the licensee for incorporation into emergency procedures are suggested in our response to the preceding Interrogatories.
In addition, it may be prudent to include level information as one of the indicators for other existing emergency actions which are now dependent on existing signals. Level information provides direct indication giving protection against anomalous behavior which is not always available from other indicators that must be interpreted assuming stylized behavior.
The NRC has required that procedures and supporting analyses for utilizing the reactor vessel level indication be provided by the licensee in January, 1981.
MET-ED Interrogatory 10 Does the staff dispute the statement that one of the bases underlying the requirement for vessel level indication is the ability to use such information for Post-Accident Evaluation?
USNRC Response to Met-ED Interrogatory 10 The staff agrees that the vessel level indication will aid post accident evaluations.
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