ML20090B439
| ML20090B439 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/19/1982 |
| From: | Speis T Office of Nuclear Reactor Regulation |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082380335 | List:
|
| References | |
| FOIA-91-106 NUDOCS 8204080277 | |
| Download: ML20090B439 (8) | |
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MDORANDUM FOR: Gus C. Lainas Assistant Director for Safety Assessment, DL FRON:
Thesis P. Spets, Assistant Director for Reactor Safety DSI
SUBJECT:
DSI INPUT TO CONGRESSMAN 00ALL'S LETTER OF 2/5/82 Per your request, we have addressed the concerns identified in Congressman Udall's letter of February 5,1982, and which were assigned to DS! in your letter of February 11, 1982.
Specifically, the Reactor Syster.s Branch addressed questions 6,11,15 and 17 part II; and the Accident Evaluation Branch addressed question 10. Be-cause we have an interest in a number of other questions raised by Mr. Udall (e.g., questions 2, 7, 8, 9, and parts of 13,14 and 18), we would be happy to review the drafts to these questions at your pleasure.
Unxical Signed Hy Theutis p,3g,j, Themis P. Spets. Assistant Director for Reactor Safety.
Division of Systems integration QM st t s
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QUESTION 5 What would the course of the incident have been had
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the PORY block valve failed to close partially or fully following failure of the PORY to close fully?
ANSWER Had the block valve failed to close af ter the PORY stuck open, the additional ecolant loss from the primary system would have caused the primary system pressure to continue to decreate below 900 psig.
As the pressure in the reactor system decreased, the combined leak flow through the valve and the rup-ture would decrease and safety injection flow would increase until the flows were approximately equal.
Analyses by Westinghouse in WCAP-9600 (Ref.1) indi-cate that the reactor system pressure would stabilize at approximately 700 psia.
The pressure would then remain relatively constant until the operator took action to depressurize the plant with the intact steam generator.
If the block valve were only partially closed, the combined leak flow and safety injection flow would equalize at a pressure between 700 psia and the 1300 psig pressure which was reached at Ginna after the block valve was fully closed.
Additional leakage out of the reactor system through the broken steam generator tube would not occur for the case of the block valve stuck fully open since the primary-system pressure would be less than the affected steam generator pressure.
If the block valve were only partially closed, the reactor system might repressurize so that some leakage out the broken tube could occur; however, it is expected that the leakage would be less than that which occurred with the block valve fully closed.
The effect on core coolant inventory of a combined PORV leak and steam generator tube leak would be similar to a postulated break in the reactor coolant hot leg with an equivalent break size of about 24 square inches.
The consequences of this tvent on core cooling would be bounded by the spectrum of small break analyses performed for Ginna (Ref. 2).
These analyses demenstrated that the core is adequately protected by the Emergency Core Cooling System in the event of a small break LOCA.
The staff concludes based on the discussions above that the effect of the block valve failing to close or leaking during the event at Ginna would have been a decrease in coolant loss through the steam generator tube anti an increase in coolant loss through the PORV.
Since coolant loss through the FORV is confined within the containment building and coolant loss through the broken tube may be released through the secondary system safety valves, offsite doses would probably have been lessened had the block valve stuck open at Ginna.
Small break LOCA analyses for Ginna indicate that the core would be adequateli cool 5d ha'd'the block valve failed to close.
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REFERENCES l
1.
Report on Small Break Accidents for Wescinghouse NSS System, WCAP_ _9600, Westinghouse Electric Corporation, June 1979.
2.
Letter from LeBoeut, Lamb, Leiby f, MacRae. Attorney for PG&E, to L. Muntzing US AEC, transmitting small break 10CA analyses for Ginna, September 6, 1974,
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t QdESTIONI,0. What consideration has been given the potential for radioactivity escaping PWRs via a path including breaks in steam generator tubes i.
and a stuck open safety valve?
ANSWER Steam generator tube rupture accidents are one of the class of design basis ac.
cidents considered by applicants and staff in each review of PWR license applica-tions.
The staff's Standard Review Plan, NUREG-0800, describes the criteria and procedures used at Section 15.6.3, " Radiological Consequences of Steam Generator Tube Failure (PWR)" (copy attached).
The analysis focuses on the potential release of radioactive noble gases and j
radiciodine both pre-existing in the reactor primary and secondary coolant, and generated concurrently with -the accident.
The former case uses the maximum activity levels permitted by the plant's proposed Technical Specifications.
The latter case postulates activity released from the fuel as a result of the accident, including the potential for furi failures.
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The steam generator tube f ailure is assumed to be a double ended rupture of a single tube for purposes of calculating the rate of transf4 f primary coulant to the secondary side of the affected steam generator.
Fl ( t ag of the primary coolant is assumed to occur in +.his process with subsequent atomization and trans-fer of activity to the steam phase.
Radioactivity leaving the steam generator is assumed to become airborne immediately and transported directly to the atmosphere via leakage paths not mechanistically specified.
Such leakage could be through a i
stuck open safety valve, an open atmospneric dump valve, or through the condenser vent syst em.
The release is assumed to ha terminated when the primary and second-
'ary coolant system pressures have ecualized.
For FSAR safety analyses, this is usually assumed to occur at about 3u minutes after the _ event initiation.
Exclusion area boundary and low population zone boundary doses are calculated and comparad with the thyroid ar.d whole body dose guideline values cited in 10CFR Part
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100.
Contervative values of site spcci'ic a+stospheric disper sion characteristics are used in these calcclations.
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t QUESTION 11.
Is it generally agreed that if a leak had developed in both steam generators, the operators would have been able to institute the
" feed and bleed" process described in Mr. Spets' January 28 memo-randum?
ANSWER Had a leak developed in the second ("A") steam generator at Ginna, the need to institute the " feed and bleed" process to assure continued core cooling would have depended upon the leak size and total leak rate of primary coolant out of the primary system.
The primary concern associated with two leaking generators is that in order to use the steam generators to cool down the primary system to the residual heat removal (RHR) system entry level, the primary system pressure would have to rett.ain slightly higher than the pressure in both faulted generator secondaries during cooldown.
This would result in continued leakage of primary coolant to the secondary system.
Primary coolant would have to be replaced by the high pressure injection (HPI) system which pumps water from the refueling 4
wa ter storage tank (RWST) into the primary system.
Thus, the allowable leakage depends on the ability to cool the plant to RHR entry conditions prior to deplet-ing the RWST.
For small loss-of-coolant accidents in the primary system, the leaking water will accumulate in the containmen't sumps.
Once the RWST level drops to a preset value, the pump suction is switched from the RWST to the sump and sump water is recirculated through the core.
Decay heat is ultimately removed by the containment heat remoYal system.
For larger tube leaks in both steam generators, which might deplete the RWST in-ventory prior to RHR entry conditions being reached, the operators would be expected
_to open all PORVs to rapidly depressurize the primary system (as well as remove de-cay heat) to below the faulted steam generator secondary pressures, and isolate both stq3m generators. Primary coolant makeup would be accomplished with the HPI pumps.
At Ginna, a two-loop 1300 Mdth Plant, there are two PORVs manufactured by Copes-Vulcan with a relief capacity of 179,000 lb/hr. steam.
Although neither the staff nor the !Icensee has performed any detailed calculations, scoping estimates indi-cate that the Ginna plant can remove decay heat by the " feed and bleed" process.
It is noted that failures in both steam generators are presently not required in the design base for PWRs.
Furthermore, existing emergency procedures, such as those at Ginna at the time of the tube rupture accident, do not provide the oper-ators with explicit guidance on how to cooldown the plant with ruptures in multiple steam generators. However, as a result of the THI accident, the staff's THI Action Plan item I.C.1 requires the industry to upgrade emergency operating guidelines and -
procedures to cover multiple failure events.
One of the specific events cited in NUREG-0737 is tube failures in multiple steam generators. Significant resources to the upgrading of guidelines and procedures have been allocated by both the in-dustry and the staff. We anticipate' approving the new energency procedure guide-lines by the end of-FY 82.
If this goal is met, upgraded procedures should be-implemented-at all operating plants by FY 83.
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Is this a period
- How long did it take to reach cold shutdown?What was the reason for the period QEST1_0N.15_
longer than desirable?What kind of malfunctions during the extended longer than normal?cooldown period might have led to a significant release of r activity to the environment?"
ANSWER The time THe plant was-in cold shutdown the day folloWog the event (6.53 p.m.).
25 minutes, from reactor trip to cold shutdown was 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> In The period from reactor trip to cold shutdown was not longer than desirable.
fact, there was no urgent need to reach cold shutdown conditions after the steam generator tube ieak had been terminated (equaliziag primary pressure with the This faulted steam generator) and the plant was in a stable shutdown in general, it is expected that cooldown with a ruptured tube in one steam generator This slower cooldown is would be significantly slower than a normal cooldown.
because the reactor coolant system pressure is to be equalized to the pressure in the ruptured steam generator to minimize or terminate reactor coolant leak flow Since the direct release of steam from the ruptured steam from the through the rupture.
generator is to be minimized (the steam would contain radioactive product!
primary system), depressurizing the faulted steam generator must be by other less Therefore, the rate at which the faulted steam generator can be depressurized is a limiting factor for the rate of reactor coolant system cooldown direct means.
and depressurization.
In Ginna, the ruptured steam generator was drained to the reactor coolant system Additional cociing and depressurization was provided through the ruptured tube.
by cold auxiliary feedwater which replaced part of the drained water.
If there has been no steam release from the ruptured steam generator in the early stage of the event, it is reasonable to expect that the cooldown perloc would have For a large initial steam space in the ruptured steam generator, a limiting factor for steam generator draining is +ne need to keep steam generator been longer.
Should the steam come in direct contact with the tubes, rapid condensation would occur resulting in a rapid depressu zation of the ruptured steam tubes covered.
i generator secondary side and re-initiation of reactor coolant leakage back through the ruptured tube.
During the extended cooldown period at Ginna, the ruptured stea The reactor coolant All other steam valves from the steam generator were secured.
system was controlled similar to a normal cooldown, except for measures (increase letdown, boration) to accommodate the leak flow to the primary system coming from the secondary side.
10, potential releases of radioactivity As indicated in the response to questionto the environs during the short term o Such leakage could bc through a stuck malfunctions in the faul ted steam generator. valve flow path or through the condenser vent syste For FSAR open safety or rel.
during the first 30 minutes
_ radiological safety analyses, such releases are assumedof the event, after wh m
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QUEST 10N 17.
(Part II)
Does the Commission believe that conditions might develop in PWRs calling for the use of remotely controlled valves for the purpose of venting steam?
ANSWER In PWRs with inverted U-tube steam generators (i.e., Westinghouse and Combustion Engineering reactors), high point vents are required to be located on the vessel head.
This requirement was added for the purpose of providing a vent path for non-condensible gases that could accumulate in the primary system under degraded core cooling conditions. Although these vents could be used to vent steam which might accumulate in the vessel upper head after saturation conditinns are reached in parts of the vessel, it is not expected they would be used for this purpose, nor is it recommended that they be used to vent steam.
Steam in the upper head of Westinghouse and Combustion Engineering reactors does not pose a direct threat to continued core cooling.
If the steam bubble were to expand to the hot leg out-lets, it would most likely condense as it came into contact with subcooled water exiting the core.
If, for any reason, the water exiting the core was saturated, the steam would enter the hot leg pipes and travel to the steam generators, where it would be condensed. '
For events such as the one at Ginna, the method we prefer for removing steam which accumulates in the upper head of the vessel is to restart a reactor coolant pump.
The pump will force subcooled water into the upper head region and condense the steam bubbl e.
The operators at Ginna demonstrated the capability to do this following the formation of a steam bubble in the upper head.
~Ir PWRs with once-through steam generators (OTSGs) (i.e., B&W reactors), a steam bubble in the upper head of the vessel has the potential to temporarily interrupt natural circulation if it expands and is able to enter the hot leg outlets without condensing.
These plants will have high point vents installed on the top of the hot leg inverted U-bends.
In addition, some utilities with B&W reactors are in-stalling vents on the top of the vessel head.
Analyses by B&W have indicated that interruption of natural circulation is a temporary phenomenon.
The analyses show that system repressurization following the interruption of natural circulation will ultimately produce thermal-hydraulic conditions in the primary system which restore natural circulation.
The staff is still reviewing the capability of the B&W analysis methods to properly predict the relevant thennal-hydraulic phenomene.
B&W has recently recommended use of the hot leg high point vents to vent steam which may accumulate during the recovery phase of a small break loss-of-coolant accident (SBLOCA).
During the accident phase of a SBLOCA, B&W has recommended the " bumping" of the reactor coolant pumps to sweep any steam trapped in the hot leg high points into the steam generator.
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, QUESTION 17. (Part II)
ANS'WER(continued)
The use of the high point vents to vent steam in B&W reactors, as wall as the acceptability of the B&W calculational models to properly predict the thermal-hydraulic behavior of the primary system under two-pha se condit;:ns, is under active staff review.
At this point in the review, it is our preliminary con-clusion that the use of the vents in B&W reactors to remove steam which accu-mulates at primary system high points may be the preferred method of steam re-moval if a reactor coolant pump cannot t.e restarted and run continuously.
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