ML20037A607
| ML20037A607 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/17/1977 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Goller K Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20037A606 | List: |
| References | |
| TAC-08377, TAC-8377, NUDOCS 8003160312 | |
| Download: ML20037A607 (5) | |
Text
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UNITED STATES J
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NUCLEAR REGULATORY COMMIS'"9N f
Yjj WASHINGTON, D. C. 20566
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ggg 171977 MEMORANDUM FOR:
Karl R. Goller, Assistant Director for Operating Reactors, DDR FROM:
Darrell G. Eisenhut, Assistant Director for Operational Technology, DOR
SUBJECT:
UNREVIEWED SAFETY QUESTION RELATED TO BORON DILUTION INCIDENT AT CRYSTAL RIVER UNIT NO. 3 A moderator dilution incident occurred at Crystal River Unit No. 3 (CR-3) on February 7,1977 due to the opening of a single valve while the unit was in cold shutdown. Approximately 600 gallons from the Na0H tank were injected into the primary system 'thus diluting the boron concentration in the moderatpr.
This caused an increase in f approximately.52% AK/k from an initial subcriticality reactivity p/k.
of -5.07% A4 The injection was quickly terminated and the core remained subcritical by a large margin.
The event was reported in Licensee Event Report 77-17, dated March 1,1977.
The B&W evaluation of this event indicated the possibility of an un-reviewed safety question as defined by 10 CFR 50.59(b) in that the unterminated injection of the NaOH tank contents into the Reactor Ccolant System (RCS) cculd result in core criticality with all rods inserted.
The B&W analysis was performed conservatively by using the most severe conditions allowed by technical specifications:
1)
Initial boron concentration,1488 ppm l
- 2) Reactor Vessel (RV) depressurized at T <1000F
- 3) RV volume of liquid drained to approximately RV outlet height 3
(s3000 ft ) This condition occurs only before or after removal of reactor vessel head or if certain types of maintenance work is being done.
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- 4) All control rods in (maximum worth CRA conservatively assumed stuck out)and beginning of life rod worth conditions, i
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- 5) Na0H assumed to be demineralized water for the purposes of moderation dilution.
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AUG 171977 Karl R. Goller.
Calculations were performed assuming unterminated dilution at various injection rates.
Time to criticality was calculated to be approximately 18-3/4 min when injecting at 400 gpm; the rate at which B&W calculated the actual NaOH injection occurred.
Time to criticality was calculated to be approximately 2-3/4 min if injection occurred at the full design flow rate of the decay heat pump (3000 gpm).
However, as discussed below, this high flow rate cannot occur.
The B&W analysis concluded that the emptying of a NaOH tank as dilutant into the reactor coolant presents an unanalyzed moderator dilution event.
Fi'rther studies made by Gilbert Associates, Inc. concluded that the greatest possible flow frcm the NaOH tank could be only 350 gom due to the 3" pipe from that tank. Using this flow rate, the time to criticality would be just over 22 minutes, again assuming the worst conditions allowed by technical specifications.
The specific criteria for operator action as given by the Standard Review Plan are that freu the tine an alarm makes the operator aware of an unplanned moderator dilution, the following minimum time intervals must be available before a loss of shutdown margin occurs:
(1) During refueling:
30 min.
(2) During startup, cold shutdown, hot standby, and power operation:
15 min.
During injection there would be several alarms and indications of dilution of reactor coolant to the operator:
(1)
Position of the valve whose opening caused the problem (an alarm on this valve is being considered).
(2) High and low level indicators and alarms on the NaOH tank.
(The technical specification limits on level are 11,190 gallons and 12,010 gallons, so the loss of 600 gallons, as happened in the February 7 incident, would not necessarily cause an alarm. How-ever, a loss of 820 gallons or more from the NaOH tank would cause an alann. An addition of approximately 8000 gallons to the RCS is necessary for criticality.)
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O AUG 17 577 Karl R. Goller (3) The nuclear instrumentation would indicate a change in reactivity before the reactor were critical Due to the length of dilution time necessary before criticality and the number of indications and alarms -to the operator, Florida Power is planning to use administrative controls to avoid this type of dilution. They are presently evaluating the entire chemical-additive system and will include an evaluation of this unreviewed safety question as part of this analysis which is due on or before Sectember 3,1977.
The accident analyses section of the Standard Review Plan does seem to ignore dilution sources (other than the chemical and volume control system) such as the NaOH tank which might be injected into the primary system and result in a dilution accident.
Other piants have been checked and this type of dilution is possible en at least some of these-plants.
(Rancho Seco and Arkansas Unit 1). We recommend that a letter ce sent to all
-licensees asking them to investig te the potential for boron dilution and report back in 60 days.
A dr ft of such a etter is enclosed.
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' MM l-r varrell G. Eisennut, Assistant Director for Operational Technology Division of Operating Reactors
Enclosure:
As stated cc:
V. Stello B. Grimes J. Donohew R. Baer F. Coffman S. Weiss C. Berlinger M. Chatterton H. Vander Molen J. Crooks P. Check J. Guibert C. Nelson
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O SAMPLE LETTER Licensee's Address 4
Gentlemen:
RE:
Facility name Recently at an operating PWR facility, a limited boron dilution incident occurred due to the inadvertent injection of a portion of the contents of the NaOH tank into the primary coolant system while the reactor was in tne cold shutdown condition. This injection event occurred subsequent to the performance of surveillance testing (valve cycling) on the isola-tion valve for the NaOH tank.
In the above-mentioned case, only a limited arount of NaOH (accroximately 600 gallons) wss injected and the reactor remained subcritical by a large margin. Hov:ever, this event highlighted the fact that a postulated single failure at this facility (i.e., misposition of the isolation valve for the NaOH tank) could result in a moderator dilation incident which had not been previcusly considered.
Subseouent analysis by the licensee and his vender revealed that, for certain conservative assumptions (e.g.,
4 reactor in the cold shutdown condition, vessel temperature less than 100*F, beginning of core life characteristics, vessel drained to a level aoproximately ecual to the height of the outlet nozzle, lowest initial boron concentration allowed by Technical Specifications, the maximum worth control rod stuck in the fully out position, and no credit assumed for operator action), the unterminated injection of the Na0H tank contents into the primary coolant system due to the misposition of a single isola-tion valve could have resulted in reactor criticality with the control rods inserted.
Based upon our review of this particular incident, we have determined that the assumption that operator action would not be taken in sufficient time to terminate the event prior to reactor criticality is overly con-servative. This determination was influenced by the length of the dilution time necessary before return to criticality and by the number of indications and alarms available to the operator at this facility. However, as a result of plant-specific system design and instrumentation differences, we are requiring that each licensee of a PWR facility provide an analysis of the potential for and consequences of a similar boron dilution incident at their facili ty.
Consequently, you are hereby requested to perform and submit the results of such an analysis within 60 days of receipt of this letter.
Your analysis should be based upon conservative assumotions consistent with conditions allowed by your facility technical soecifications and should
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_2 include the assumption of the most limiting single failure (e.g., mis-position of a valve.which performs an isolation function for a potential source of a moderator dilutant). The analysis should also include an assessment of the factors which affect the capability of the operator to take corrective action which would terminate the postulated event prior to return to reactor criticality.
If, based on the results of this analysis, you determine that corrective actions (design or procedural) are required to preclude the occurrence or mitigate the consequences of postulated boron dilution accidents, your response should include proposals for such actions.
e
, Chief Operating Reactors Branch No.
Division of Operating Reactors
.