ML19341C616
| ML19341C616 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/31/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0011, NUREG-0011-S04, NUREG-11, NUREG-11-S4, NUDOCS 8103030836 | |
| Download: ML19341C616 (31) | |
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NUREG-0011 Supplement No. 4 Safety Evaluation Report related to the operation of Sequoyah Nuclear Plant, Units 1 and 2 Docket Nos. 50-327 and 50-328 Tennessee Valley Authority U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation January 1981 s
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1 TABLE OF CONTENTS Pag i
1 Introduction and General Discussion................................
1-1 1.1 Introduction..................................................
1-1 22 TMI-2 Requirements.................................................
22-1 j
II.B.7 Analysis of Hydrogen Control..............................
22-1 System Description.................................
22-1 l
i Testing of the 1015.....................................
22-2 Analysis of 1015.....................,..................
22-5 Probability of Core Damage Events.......................
22-6 Hydrogen Generation and Containment Pressures...........
22-10 Survivability of Essential Equipment....................
22-15 Sequoyah Containment and Structural Capacity............
22-18 Conclusions.............................................
22-23 i
TABLES 22.2-1 Core Melt Probabilities...................................
22-7 i
22.2-2 Containment Analysis Sensitivity Studies..................
22-13 22.2-3 Mean and Standard Deviation of Material Properties........
22-22
~ FIGURES l
22.2-1 Estimated Time Sequences for Potential Core Damage i
Events....................................................
22-8 i
22.2-2A Ice Condenser Air Ducts and Insulation.....................
22-20 l
22.2-2B Enclosure of Foam Insulation.............................
22-21 1
APPENDIX D Advisory Committee On Reactor Safeguards Letter.............
D-1 iii
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1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction On September 17, 1980, the Nuclear Regulatory Commission (NRC) issued the facility operating license DPR-77 to the Tennessea Valley Authority for the Sequoyah Nuclear Plant, Unit 1, located in Hamilton County, Tennessee.
The licensa authorized operation of Unit 1 at 100 percent power; however, a license condition regarding the adequacy of the hydrogen control system was included that required resolution by January 31, 1981.
The purpose of Supplement No. 4 to the SER is to further update our Safety Evaluation Reports on the hydrogen 4
control measures (Section 22.2, II.B.7), and to comply with the license condition which is as follows:
"By January 31, 1981, TVA shall by testing and analysis show to the satisfaction of the NRC staff that an interim hydrogen control system will provide with reasonable assurance protection against breach of containment in the event that a substantial quantity of hydrogen is generated."
TVA submitted on December 11, 1980, the first quarterly report on the research program for hydrogen control.
Also, TVA revised volume 2 of the Sequoyah Core Degradation Program Report to incorporate additional information on the overall program.
Section II.B.7 of Supplement No. 4 responds to the license condition.
Each section is supplementary to and not in lieu of discussion in the Safety Evaluation Report and Supplements Nos. 1, 2, and 3, except where specifically noted.
Supplements No. 2 and 3 to the Safety Evaluation Report provided a basis for 4
concluding that the full power licensing of Sequoyah Unit 1 need not await completion of ongoing work on hydrogen control measures.
This supplement concludes that operation of the IDIS for an interim period of one year is appro-i priate.
The ACRS considered this matter and reported its finding in a letter te the Chairman dated January 13, 1981.
A copy of this letter is in Appendix D.
They concur with the NRC staff recommendation to allow interim operation of the IDIS.
1-1
22 TMI-2 REQUIREMENTS 4
11.8.7 Analysis of Hydrogen Control
System Description
The Tennessee Valley Authority (TVA) has installed within the Sequoyah Unit I containment a system of igniters and ancillary equipment designated as the i
interim distributed ignition system (IDIS). The igniters are de3 gned to ensure a controlled burning of hydrogen in the unlikely event that excessive quantities of hydrogen, well beyond the design bases required by 10 CFR Section 50.44, are generated as a result of a postulated severely degraded core accident.
The igniter selected by TVA is a glow plug commonly used in diesel engines that is manufactured as Model 7G by General Motors AC Division.
The igniter L
is powered directly from a 120/14V ac transformer.
The igniter assembly consists of a steel box with 1/8-inch-thick walls, which houses the trans-j former and all electrical connections and partially encloses the igniter.
The glow plug side of the box is covered by a spray shield; and the glow plug face of the igniter assembly has a copper heat sink.
The igniters are powered from the standby lighting system which has normal and alternate ac power supply from offsite sources.
In the event of a loss of offsite power, the igniters would be powered from the diesel generators.
The 1015 was not designed as a safety grade system and as such is not a seismic Category I system; but it is seismically designed not to damage other safety-related equipment inside containment.
This is accomplished by securing the igniter assemblies with a steel cable attached to an anchoring bolt.
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The interim distributed ignition system presently installed in the Sequoyah Nuclear Plant Unit 1 consists of 32 igniter assemblies distributed throughout r
I the upper, lower, and ice condenser compartments.
There are a total of 20 igniters in the lower compartment and 3 igniters, which are suspended 35 feet from the top of the containment, in the upper compartment.
There are a total l
of 9 igniters located in the ice condenser compartment; 5 in the lower plenum j
and 4 in the upper plenum of the ice condenser.
As a supplement to the 32 igniters presently installed, TVA has committed to install 13 additional igniters, all to be located in the upper compartment, before the end of the first refueling outage.
The interim distributed ignition system is designed such that it can be manually actuated following the start of an accident and can remain actuated until the unit reaches cold shutdown.
The system is to be manually initiated by switching on three lighting circuits at the standby lighting panel located in the auxiliary building.
To ensure that the interim distributed ignition system will function as intended, TVA has proposed a preoperational and surveillance testing program.
Preoperational testing, to be performed upon installation of the system, will 22-1
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verify:
(1) that the output voltage of the transformer is greater than or equal to 12 volts and is less than or equal to 14 volts and (2) that the temperature of the igniter is at least 1500 F.
During the preoperational tests the current in each circuit will be measured and the results used as the base-line for future surveillance tests.
The igniter system will be subjected to periodic surveillance testing which will consist of energizing the IDIS at the standby lighting panel and taking cerrent readings of the circuits.
If the current readings do not compare favorably with current measurements taken during preoperational testing, then all igniters will be visually and otherwise checked to ensure their operability.
Testing of the IDIS TVA has conducted two testing programs to obtain information pertinent to the performance characteristics of the glow plug igniters.
Preliminary screening dnd qualification testing was performed at TVA's Singleton Laboratory.
Com-bustion tests using the TVA igniter were performed by Fenwal, Inc. to study igniter performance under various environmental conditions.
The principal purpose of the testing at Singleton Laboratory was to evaluate igniter surface temperatures and determine the effects of overvoltage condi-tions and extended operation.
Test ng of a GM plug was conducted at 12, 14, and 16 volts ac with surface tempeNtures of 1480, 1550, and 1650 F as measured by a thermocouple.
Since thermocouple heat losses were estimated to be signif-icant, the surface temperature of another plug at 14 volts was measured with an optical pyrometer obtaining at least 1800 F.
Subsequent testing on several GM glow plugs has been performed to ensure adequate surface temperatures at the minimum design voltage.
The minimum acceptance temperature of 1500 F was reached within 1 minute, with ali plugs reaching 1600 F in 3 minutes.
Voltage tests on 5 plugs verified re'iable operation at 14 volts, but two plug failures were recorded after a minute's operation at 16 volts.
Initial endurance tests were conducted on a plug by operating it continuously at 14 volts for 148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br />.
After the successful completion of the test, the plug was then used in hydrogen burning tests.
After the GM glow plug was selected, TVA began more extensive endurance testing to select the igniters which would be installed inside containment.
From a lot of 302 plugs which were subjected to a screening test to eliminate those with manufacturing defects, 50 plugs were randomly sampled.
These plugs were then tested by cycling the plug and then by continuous testing for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to verify successful operation.
Based on the results of these tests, we find that the selected igniters should perform satisfactorily in terms of reliably achieving the desired surface temperature of at least 1500 F for the required duration of 3 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I To evaluate the efficacy of the igniters, TVA in cooperation with Duke Power, American Electric Power, and Westinghouse developed a two phase testing program, l
which was conducted by Fenwal, Inc.
The testing was conducted using a single i
igniter assembly in a spherical vessel approximately 6 feet in diameter (134 i
3 ft ).
Features of the test assembly include external electrical heaters, internal fan, and water spray.
The vessel atmosphere pressure and temperature conditions were measured, as was the vessel surface temperature.
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Instrumentation was also added to measure the surface temperature of equipment samples located in the test vessel.
The hydrogen concentration was determined by taking samples for gas chromatograph analysis.
The phase 1 tests were conducted to determine the ignition efficacy of the igniter for varying hydrogen concentrations under various test vessel atmos-phere conditions of pressure, temperature, steam concentration, and turbulence.
Turbulence was simulated by the internal mixing fan to produce estimated flow velocities of 5 and 10 feet per second past the igniter.
The phase 2 portion of the test program was designed in four parts to investi-gate separate effects which may affect igniter performance and the response of typical equipment in the Sequoyah plant.
The part 1 tests were performed to evaluate the igniter performance in the lower hydrogen concentration range.
The part 2 tests were conducted to determine igniter efficiency under tran-sient conditions whereby steam and hydrogen were introduced continuously into the test vessel.
The part 3 tests were conducted to evaluate the effects of water spray on igniter performance.
The part 4 tests were designed to provide information on the effects of a hydrogen burn on typical equipment, e.g.,
limit switch and solenoid valve, by measuring the equipment temperature response.
The results of the Fenwal test program were generally consistent with the applicable published information on hydrogen combustion.
The tests confirmed that limited combustion occurs over hydrogen concentrations of 6-8 percent although the completeness of combustion is influenced by the ability to pro-mote mixing thereby exposing the igniter to sources of fresh atmospheres.
In the regime of hydrogen concentrations of 8-9 percent, test results indicate the combustion process is altered; this range of concentrations represents a transition zone where combustion may proceed to a nearly complete reaction.
Again, this is consistent with published data and findings regarding the general limits of upward and downward flame propagation.
At the higher hydro-gen concentrations of 10-12 percent that were tested, the results indicate that the likely scenario is that the combustion process will proceed to con-sume all the hydrogen present in the atmosphere.
The test vessel atmosphere pressure measurements reflected the completeness of combustior, at the various hydrogen levels.
Pressure measurements showed an increase of as little as approximately 1 psi for single hydrogen burns at 6 percent concentration and an increase of approximately 70 psi for the burning of a mixture of 12 percent concentration.
The phase 2, part 2, portion of the test program dealing with transient hydrogen injection with the plug pre-energized resulted in a series of eight burns for the test with continuous injection of both steem and hydrogen.
The injection flow rates were scaled to simulate the flow rates calculated to enter the lower compartment of an ice condenser containment for a small-break loss-of-coolant accident (LOCA).
The results of that test indicate relatively small pressure increases of approxi-mately 2-7 psi for each of the eight discrete burns.
These lower pressure excursions result from the burning of hydrogen at lower concentrations and allowing for heat removal via heat transfer to heat sinks during the transient.
It is to be reasonably expected that this type of sequence simulates the transient that would occur if hydrogen were to be introduced inside Sequoyah in a like manner.
22-3
l The effects of spray operation on igniter performance appear to be minimal in terms of affecting the ability of the igniter to initiate combustion.
There is some evidence, however, that spray operation, by promoting mixing and t
turbulence, may actually improve the efficiency of the 101, in burning lean hydrogen mixtures.
Similarly, the effect of fan flow acrosi the glow plug, within the considered flow rates, indicates minimal alteration of the ignition time and some improvement in the ability to burn a larger fraction of the hydrogen in lean hydrogen mixtures.
Based on the results of these tests as described in the applicant's submittal, we find that the glow plug igniter will serve its intended function to initiate combustion of flammable mixtures under various conditions.
To independently evaluate the efficacy of the IDIS, the NRC staff arranged for the testing of the GM glow plug igniter at Lawrence Livermore National Labora-tory.
The test program was designed to examine the performance of the igniter under a spectrum of test conditions.
The principal parameters of concern in this testing were varying hydrogen and steam concentrations.
The Livermore tests were conducted in an insulated pressure vessel having a 3
volume of 10.6 ft.
Unlike the vessel used in the Fenwal te:t ;'rogram, this vessel was unheated, which allowed a faster steam condensation rate and heat transfer to the environs.
Primary data for the Livermore tests included atmosphere pressure and temperature measurements e d gas concentration analysis.
1 Gas analysis was accomplished by drawing samples from the vessel; samples were taken before and after eact h drogen burn.
Mass spectrometric analysis was
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the method by which volume -
entrations were determined.
The test me;rix included dry tests with hydrogen and air mixtures ranging from 6 to 16 percent hydrogen, and steam tests with 30 and 40 percent steam fractions, each with varying hydrogen concentrations.
A total of 43 tests was run.
The Livermore combustion tests confirmed the ability of the proposed TVA l
igniter to ignite gas mixtures over a range of conditions.
In the dry air j
tests Livermore was able to partially burn the hydrogen at lower concentra-tione (7-8 percent) and completely burn the hydrogen at higher concentrations, simiiar to what was seen in the Fenwal test program.
For the tests with 30 and 40 percent steam concentrations, it appears that the flammability limit for downward flame propagation was snifted upward to higher hydrogen concen-trations.
The results of a test at 30 percent steam and 10 percent hydrogen indicated a partial burn.
However, the igniter never failed to initiate combustion for any of the 30 and 40 percent steam fraction tests.
The measured pressure increases for tests with hydrogen concentrations over the range of 8-12 percent varied from approximately 1 psi to 65 psi, showing close agree-ment to the measured pressure increases during the Fenwal tests.
There was no detectable deterioration in the ability of the glow plug to initiate combustion through the test series using the same plug for all tne tests.
It consistently initiatai hurning in dry mixtures at plug temperatures of 1310-1370 F for the dry tests and at 1360-1480 F for the steam tests.
As previously discussed for the Fenwal test program, it was apparent that promotion of mixing or turbulence enhanced combustion at lowi_r concentrat ions.
This phenomenon, while not specifically addressed in the Livermore tests, was apparently manifested as secondary burns for tests when the circulating fan 22-4
was activated after the initial burn and the igniter temperature remained high enough to reinitiate combustion.
Several tests were run with a nominal steam fraction of 50 percent.
Although the degraded core accident scenario considered in the Sequoyah review would not result in such high steam concentrations, Livermore performed tests to determine what steam fractions would be necessary to prevent ignition of the mixture.
In all the tests with a 50 percent steam fraction, Livermore was unable to initiate combustion.
This steam concentration which was observed to effectively inert the vessel atmosphere also approximates the values quoted in the reference literature.
Two tests which began at nominal steam concentrations of 50 percent, with no initial combustion, were allowed to continue with the steam fraction being gradually reduced by condensation on the vessel wall.
Even though the steam fraction was eventually reduced to levels where combustion should have occurred, no substantial pressure increase, as a result of hydrogen burning, was recorded.
Livermore is unable at this time to conclusively resolve why a pronounced burn did not occur and plan to continue the investi-gation of this matter.
Because the initial steam concentration is outside the spectrum of conditions calculated for the Sequoyah plant during the accident used as the basis for review, we see no immediate cause to consider these particular test results as a bases for rejection of the TVA ignition system as an interim solution to hydrogen control for degraded core accidents.
The statf is planning to continue the Livermore tests for several months to investigate the effects of containment spray operation on igniter performance and to further study hydrogen combustion in steam environments. We will report on the continuation of this testing in a future supplement to the Safety Evaluation Report.
Analysis of 101S To evaluate the role of igniters in accident mitigation, TVA has initiated an analytical effort to determine the effectiveness of distributed ignition systems in reducing the threat to containment integrity due to th combustion of that hydrogen generated following a spectrum of postulated degraded core accidents.
It is expected that thorough analyses including sensitivity studies on critical parameters for a range of accident scenarios will continue for approximately one year.
Currently, TVA has provided the results of analyses for a single degraded core accident scenario, designated $20 in WASH-1400, which is a small-break LOCA accompanied by the failure of emergency core cooling injection.
The S2D sequence leads to the production of hydrogen from the zirconium-water reaction as a result of the degraded core conditions, i.e., lack of core cooling.
TVA has concluded as a result of studies to determine accident sequence proba-bilitie that a small-break LOCA followed by a failure of emergency core cooling injection is one of the more probable sequences among the spectrum of serious degraded core accidents.
TVA has cited calculations to indicate that the S2D sequence is a more likely event than a large-break LOCA. Another sequence initially considered for the purposes of evaluating the 1015 was a loss of feedwater transient followed by a loss of offsite power; however, TVA has concluded that the good availability history for offsite power near the Sequoyah plant lessens the likelihood of this accident.
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Probability of Core Damage Events The sequences of interest in evaluating a hydrogen control system were limited to those that result in only core damage with associated significant hydrogen generation from the zirconium-water reaction, as opposed to complete core melt events with associated pressure vessel failure.
In general, these sequences result from arrested core melt events where vital equipment / functions are restored after their initial failure or result from dynamic situations caused by man-machine interactions similar to the TMI-2 accident.
The frequency of these core damage sequences has not been extensively reviewed previously t
because of the difficulty in discriminating between core damage and core melt events.
TVA has presented estimated probabilities for core melt sequences based on an earlier study of Sequoyah performed by Sandia under the Reactor safety Study Methodology Applications Program sponsored by the Office of Nuclear Regulatory Research.
A list of potential events where operator action may limit the event to only core damage by restoring failed equipment is presented in Table 22.2-1.
These events were selected because of the relatively long time to total core melt (1-1/2 to 2-1/2 hours) which could allow potential beneficial operator action.
In order to have core damage with significant hydrogen generation (and not total core melt), vital equipment would have to be restored in a short tio period (15 to 45 minutes) prior to complete core melt as shown by the shaded areas in Figure 22.2-1.
Earlier recovery would result in small hydrogen generation which would not jeopardize containment integrity while later recovery would be too late to preclude complete core melt and pressure vessel failure.
The probabilities of core melt in Table 22.2-1 are estimates by Sandia and the licensee.
Without vouching for the correctness of either set of estimates, the total probability of potentially rectifiable core melt sequences (i.e.,
limited to core damage) is approximately 50 percent of the probability of all core melt sequences.
The conditional probability for restoring a vital component / system function by unplanned operator action in the time frame noted above given a potentially rectifiable event has not been determined.
Such a determination would require a decomposition of the component / system unavailabilities into their root causes and a probabilistic analysis of their rectification under a temporal frame of reference.
Mean repair times for failed pumps, valves, diesels, and instrumentation range from 6 to 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> in WASH-1400.
On the other hand, opening valves locally or jury rigging electrical connections or bypasses may only take minutes to correct misaligned systems.
Considering the above repair times, a crude estimate of the conditional probability for restoring vital component /sv tem function in the 15 to 45 minute window shown by the shaded 1
area in Figure 22.2-1 is assumed to be 0.1 to 0.2.
Combining this assumed conditional probability with the probability of having a rectifiable event, we l
would estimate that the probabil.ity of having an event with core damage and significant hydrogen generation (i.e., an aborted core melt) is approximately 5 to 10 percent of the total core melt probability.
A second mode of achieving core damage with significant hydrogen generation is for operator interruption (error of commission) of vital functions and subse-quent late restoration of the functions in a dynamic man-machine interaction 22-6 1
I TABLE 22.2-1 Core Melt Probabilities
9 x 10~6 4 x 10~6 (S 0) 2 Small LOCA, loss of ECCS recirculation **
2 x 10-2 x 10 c 7
Small LOCA, loss of ECCS and spray **
3 x 10 recirc; no containment drain effect (S HF) 2
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~6 Loss of all feedwater with AC power **
7 x 10 7 2 x 10 available (TML, TB, TLD) 2 Station Blackout **
3 x 10 7 8 x 10 9 (TMLB')
All other events 2.8 x 10 5 4 x 106 Total 5.P x 10 s 1.2 x 10 s
^Per reactor year.
- Potential core damage events with significant hydrogen generation.
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NOTE: SilADED AREAS INDICAYE COHE DAMAGE
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WITH SIGNiflCANT l
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IIYDHOGEN GENEHATION STANT COHE COHE UNCOVEHY MELT SMALL LOCA. NO ECCS INJECTION 1
1 1
n SIAHT ECCS COME COHE g
STOPS UNCOVEHY MELT c
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SMALL LOCA. NO ECCS REClRCULATION y
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SIAHT COHE COHE UNCOVERY MELT LOSS OF FEEDWATER. STATION BLACKOUT l
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0 1
2 3
4 TIME - HOURS Figure 22.2-1 Estimated Time Sequences for Potential Core Damage Events
situation such as the TMI-2 accident.
We are not aware of any detailed event tree / fault tree analysis of such man-machine interactions, however.
Some insight may be gained by considering some pote.tial events qualitatively.
There are essentially two primary situatiois that may lead to core damage with significant hydrogen generation; namely, a hole in the primary system with insufficient coolant makeup or insufficient feedwater flow to remove reactor heat leading to dry out of the steam generators and subsequent boil off of the primary coolant.
These events would have to be slow enough, like those in u plate potential inappropriate control Table 1, for the operator to even con m
(error of commission) of the vital functions of core cooling and/or feedwater flow.
Based on our review of small LOCAs and loss of feedwater events in Westinghouse plants (NUREG-0611), these vital functions could be interrupted for significant lengths of time (about 40 to 60 minutes) before operator corrective action would necessary to preclude significant hydrogen generation.
Thus, it is reasonable tu consider operator corrections of earlier inappro-priate operator termination of the vital functions.
Operator recognition and maintenance of these vital functions has been stressed by the past TMI-2 actions.
Emergency operating procedures and operator training for small LOCAs now emphasize the need to maintain high pressure injection flow if subcooled conditions cannot be maintained in the primary system.
In addition, reactor subcooling meters have been installed and a technical advisor added to the control room complement which would further enhance the fulfillment of the safety function of keeping the core covered.
Similarly, emphasis has been placed on the operation and control of the auxiliary feedwater system needed to provide the vital function of a heat sink under accident situations.
Although no formal quantification has been made of the human errors associated with these vital functions, the staff does not believe that human errors (uncorrected errors of commission) associated with controlling primary systems represent major contributors to the unavailabilities of the core cooling and feedwater delivery functions on a best-estimate basis because of the actions initiated since TMI-2.
Our qualitative review of potential human errors has been limited; it did not consider human control errors associated with support equipment like component cooling water, ventilation, etc., which could ulti-mately impact the vital safety functions and, in the absence of a systematic event tree / fault tree evaluation, the completeness of the sequences considered is uncertain.
The foregoing discussion of man-machine interaction and the necessity of maintaining certain vitd functions assumes a complete rectification of the human related deficiencies exhibited by the TMI-2 accident.
However, on an upper-bound basis, some undefined event may occur because of an unforeseen design deficiency and/or operator error of commission.
The severity of an upper-bound event cannot be predicted a priori because of the unknown circum-stances leading up to the event and the time window for an event to progress from core damage with significant hydroger generation to total core melt.
The occurrence of the TMI-2 event does not mean that an upper-bound event will more likely result in substantial core damage rather than core melt because of the correction of TMI-2 deficiencies.
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Since all core melt events are preceded by core damage with significant amounts of hydrogen generated, hydrogen control may have a significant impact on e
containment failure (and societal consequences) regardless of the relative a
frequency of core damage (only) events. We have not reviewed the potential benefit of hydrogen control under total core melt conditions.
Hydrogen Generation and Containment Pressures In order to perform analyses of the containment atmosphere pressure and temper-ature response due to a loss of-coolant accident, it is necessary that the releases from the reactor coolant primary system be known, including steam and hydrogen release rates.
The TVA containment ana % is was based on the reactor coolant system response and releases using results Pom the MARCH computer code as provided by Battelle Columbus Laboratory.
The MARCH code was developed by Battelle Columbus Laboratory for the NRC in order to provide the capability for the analysis of the thermal-hydraulic re-sponse of the reactor core, primary coolant system and containment to core melt accidents. With regard to hydrogen evolution the code provides a method of incorporating the Baker-Just and Cathcart metal-water reaction rate models with a history of the uncovering and overheating of the core to obtain hydrogen generation rates.
The extent of production of hydrogen is dependent on the degree of uncovering of the core and on the amount of water (steam) available for oxidation, but the reactor pressure at the time of release is a dominant factor in determining the amount of hydrogen released from the primary system according to MARCH.
MARCH then models the release of hydrogen with the steam from whatever openings in the primary system may be appropriate or the scenario (PORV, small break, or large break). The release of hydrogen is assumed to accompany the steam release according to the mass average composition calculated for the steam and hydrogen.
Under small-break LOCA conditions, MARCH predicts the generation and release of hydri an prior to core melt to be spread out over a 30- to 100-minute period.
Average release rates of 8 to 30 pounds per minute cover these cases.
In some instances, a burst of hydrogen into containment when the core melts completes the release.
Another mode of hydrogen release is possible through the high point vents that are to be required on LWRs.
These are to be sized such that their discharge can be compensated for by the charging pumps.
It is estimated that vents of this size will release hydrogen at the rate of about 20 lb/ min.
These considerations have led the staff to use 20 lb/ min as a base value in scoping calculations of the release of hydrogen prior to core melt or vessel failure.
This release rate is typical of small-break LOCAs, up to 2 in. in diameter, high point venting, ind TMLB accidents modified to have small break characteristics.
The hydrogen release rates emphasized in calculations by TVA are those taken from MARCH code interpretations of a small-break LOCA.
These MARCll release rates are a time-varying function whose average is of the order of 20 lb/ min.
The staff considers these rates to be representative of a significant group of 22-10
releases that might be encountered in typical degraded core accidents short of total core melt or vessel failure, and are an acceptable upper limit basis for calculations.
The consensus interpretation of the THI-2 accident is that from 30 to 60 percent of the core zirconium cladding was oxidized generating 500-1100 pounds of hydrogen.
The applicant, TVA, has submitted analyses that pursue the course of the accident rp to the time when approximately 80 percent of the core cladding has been oxidized, and has justified the termination of the analysas at that poir.L because further oxidation would result in major melting or rearrangement of the core.
The staff notes that these analyse: do not provide for any supplementaru oxidation of ferritic materials.
The ferritic materials, however, are unlikely to be oxidized substantially until the final phases of core rearrangement or 1
melting.
The quantity of hydrogen chosen by the applicant for his principal analyses is, therefore, acceptable.
TVA then, using the releases calculated from the MARCH code, calculated the containment atmosphere transient using the CLASIX code, which was developed by Westinghouse /0PS.
The CLASIX code is a multi-volume containment code which calculates the containment pressure and temperature response in the sejarate compartments.
CLASIX has the capability to model features unique to ari ice condenser plant, including the ice bed, recirculation fans, and ice condenser doors, while tracking the distribution of the atmosphere constit: ants: oxygen, nitrogen, hydrogen, and steam.
The code also has the capability ' modeling containment sprays but presently does not include a model for stru tural heat sinks.
Mass and energy released to the containment atmosphere in the form of steam, hydrogen, and nitrogen is input to the code.
The burning of hydrogen is calcu-lated in the code with provisions to vary the conditions under which hydrogen is assumed to burn and conditions at which the burn will propagate to other compartments.
The conditions inside the containment prior to the onset of hydrogen genera-tion were determined from LOTIC analyses; LOTIC being the Westinghouse long-term ice condenser analysis code previously reviewed and approved by the staff.
The CLASIX calculations then begin at the onset of hydrogen produc-l tion, which occurs at approximately 3500 seconds following onset of the i
accident.
The base case CLASIX analy:,3 otilized the assumption that hydrogen ignition within a compartment was initiated at a 10 percent hydrogen concen-tration and that burning is assumed to propagate the other compartments with a 10 percent hydrogen concentration.
The hydrogen release to the contairiment was terminated, for the containment analysis, after approximately 1550 lbs of hydrogen were released.
This mass of hydrogen corresponds to the reaction of approximately 80 percent of the zirconium cladding (excluding plenum) in the core.
At this point in the scenario the core is dry, thus there is no steam to produce a further zirconium-steam reaction.
Extending the accident scenario to the point of reactor vessel melt-through will be the subject of generic future analysis in conjunction with TMI Action Plan Item II.B.8, i.e., the rulemaking proceeding.
22-11
The results of the CLASIX base case analysis indicate that the hydrogen will be ignited in a series of nine burns in the lower compartment with the first of the burns propagating upward into the ice condenser compartment.
The total interval over which the series of burns occurs is approximately 3300 seconds.
For the first burn, a peak pressure of 26.5 psi was calculated for the lower compartment and 28.5 psia for the ice condenser and upper compartment.
The pressure in the containment before the first burn was approximately 22.5 psia.
Subsequent burns resulted in successively lower pressure peaks.
Peak gas temperatures of 2200*F, 1200 F and 150 F were calculated in the lower compart-ment, ice condenser, and upper compartment, respectively.
As a result of the action of engineered safety featuces, such as the ice condenser, air return fans, and upper compartment spray, the pressure and temperature spikes were rapidly attenuated between burns.
The pressure was decreased to its preburn value roughly 2 minutes after the burn occurred.
After the last ignition of hydrogen, whicn occurs approximately 6800 seconds after onset of the accident, there were roughly 300,000 goundsoficeleftin the ice condenser section (representing at least 40 x 10 BTVs in renaining heat removal capacity).
TVA has requested a technical specification change regarding the amount of ice required in the ice condenser.
The staff has this matter under review.
In summary, the results of the TVA base case analysis show only a modest increase in containment pressure due to hydrogen burns on the order of 4-6 psi, with the containment remaining well below the estimated failure pressures.
The burning criterion used in the analysis caused virtually all of the burning to occur in the lower compartment, thereby gaining the advantage of neat removal by the ice bed.
It should also be noted that each burning cycle involved the combustion of only 100 pounds of hydrogen, or roughly 6 x 10 6 BTUs of energy addition.
By burning at a given icenL. ion in the lower compartment (where one might naturally assume hyucogen concentrations to be higher since this is the area of hydrogen release), there is also the advantage of burning less total hydrogen at a time because the lower compartment volame is only around one-fourth of the total containment volume, which allows for expansion of the hot gases to the rest of the containment free volume.
TVA has also performed preliminary sensitivity studies to determine the effects of ignition criteria and safeguards performance on the containment response.
Results of several of these studies are shown in Table 22.2-2.
The sensitivity analyses performed to date demonstrate thut (1) the ignition criterion, within the bounds chosen, has little effect on the containment pressure; (2) partial vs full operation of the air return fans makes little difference on the results; (3) ice condenser heat removal is effective in reducing pressure; and (4) without any fal operation to assure mixing, the containment pressures due to burning rise dramatically to the point where the containment can be expected to lose structural integrity.
It should be noted that the case (refer to Table 22.2-2) which considered only enough ice exists to reduce the pressure spike for two burns (out of seven) is non-mechanistic; j
i.e., it is not representative of the actual S2D scenario.
However, it does 1
importantly demonstrate that even without ice, the containment pressure, with the assumed igniter operation, remains below the estimated containment failure pressure.
This serves to indicate some insensitivity to whatever accident scenario is chosen.
l h
)
22-12
l l
TABLE 22.2-2 CONTAINMENT ANALYSIS SENSITIVITY STUDIES Calculated Peak Pressure (psia)
Lower Upper Comp.
Comp.
1.
Base Case 26.5 28.5 2.
H Ignition and 2
Propagation 8%
28.5 30.5 3.
1 Air Fan 26.5 29.5 4.
No Ice
- 41 41 5.
No Air Fans 46.4 92.4
^ Ice exists only for the first two of seven burning cycles.
22-13
The calculations performed to date by TVA are the first efforts to analytically define the value of the distributed ignition system.
TVA plans to refine the analytical models in the CLASIX code, perform additional parametric analyse:,
and evaluate other accident sequences in assessing the effectiveness of a hydrogen ignition system.
Theu additional analyses will be discussed in a future supplement to the SP TVA has initiated effort? o verify the CLASIX code which was used to perform the preliminary containw nt transient analysis of hydrogen distribution and deflagration.
CLASIX, which was developed by Offshore Power Systems (OPS)/
Westinghouse, has been described as a code under development.
Therefore, in order to increase confidence in the code's calculations, OPS has begun efforts to verify the code by comparison with the results of other Westinghouse con-tainment codes, namely the TMD and COC0 codes.
The C0C0 code, which is the Westinghouse dry containment code, has been used for several years and most recently was used to perform containment pressure calculations with hydrogen burning in the Zion / Indian Point (Z/IP) studies.
A comparison of results has also been made for selected cases using the TMD code. The TMD code is the Westinghouse subcompartment and short-term transieat ice condenser code, which has been reviewed and approved by the staff.
For both two phase and superheated mass and energy releases, the CLASIX and TMD codes predict pressure transients in close agreement.
In summary, the efforts to date to verify CLASIX using familiar licensing codes has demonstrated that the CLASIX code adequately predicts the containment transient.
To independently assess the role of the IDIS in mitigating the consequences of a degraded core accident, the staff has obtained technical assistance from the Battelle Columbus Laboratory to analyze the containment atmosphere response to the combustion of hydrogen.
The calculations were performed using the MARCH code with a 2-volume model of the Sequoyah containment and assuming a small LOCA consistent with the TVA analysis (S20).
The MARCH code model consisted of a lower and upper compartment, with the ice bed modeled as a junction and not as a separate volume.
The MARCH code features include models for ice bed heat removal, structural heat sinks, return air fans, and containment sprays.
The sprays in the Sequoyah model, however, were presently assumed, due to code constraints, to have heat removal capacity only after the ice is completely melted.
The results of analyses performed using the MARCH code were similar to those calculated by TVA using the CLASIX code in that hydrogen combustion was cal-culated to originate in the lower compartment in a series of burns.
Following each burn and concomitant pressure spike, the containment pressure was rapidly reduced until the next burn was calculated to occur.
l The majority of cases analyzed, assuming various ignition setpoints, indicated a peak containment pressure of 23 psia.
With an initial containment pressure of approximately 20 psia prior to burning, the pressure rise following a hydrogen burn is approximately 3 psi.
Similar to previously discussed CLASIX analysis, a MARCH analysis was performed for a S20 transient with the arbitrary assumption that the ice bed had completely melted before the onset of hydrogen ourving.
Again, this assumption conserva-tively neglects calculations which demonstrate a large portion of the ice bed l
22-14
would be remaining for this accident.
Nevertheless, calculations show that without ice, the pressure rises to approximately 50 psia which is sufficiently low that containment structural integrity should not be seriously threatened.
The control of hydrogen generated during a severely degraded core accident through the use of a deliberate ignition technique such as the IDIS proposed by TVA requires the cc,nsideration of the ef fects of the environment on struc-tures and equipment.
In an environment where hydrogen deflagration is taking place two pronounced effects on the containment atmosphere, namely the con-comitant pressure and temperature increase, may adversely affect containment structures and internai equipra u.
We have discussed the analysis which has been performed to demonstrate that the advantage of deliberate ignition is to limit the amount of hydrogen burned such that relatively low pressure increases are experienced.
The burning of hydrogen, however, was calculated to result in extremely high temperatures for the compartment where deflagration occurs.
High temperatures would naturally follow the burning of hydrogen due to the relatively low thermal capacity of the atmosphere.
However, it is TVA's view, which is shared by the staff, that the CLASIX analysis which neglects of effects of structural heat sinks is too conservative in the calculation of the atmosphere temperature transient.
- TVA, therefore, has an intensive effort underway to modify the code to incorporate heat transfer to structures.
In the interim, however, TVA has provided infor-mation to demonstrate that vital equipment (i.e., equipment needed to safely shut down the plant and maintain shutdown status) will survive the hydrogen burn environment and perform its intended function.
Survivability of Essential Equipment In response to staff requests relating to the survivability of equipment required for a cold shutdown after a cegraded core event and for the prevention of breach of containment during a hydrogen burn event, the licensee provided submittals dated December 11, 15, 17, and 24, 1980, and January 22, 1981.
The submittals provided information relating to TVA's program for controlling hydrogen burns subsequent to a LOCA event, the results of testing conducted, and a list of equipment necessary to assure both cold shutdown and containment integrity. We have reviewed this list, and, as discussed in more detail below, have concluded that with two additions it contains all equipment in containment needed for containment integrity and to bring the reactor to a safe stable shutdown condition.
Further, we have concluded that this equipment and the two additions will survive the postulated hydrogen burn events.
As noted below, one additional matter is still undergoing review.
The return air fan system and containment spray system, designed as safety grade systems, operate to ensure mixing of the containment atmosphere.
The natural flow path, resulting from pressurization effects, is from t.he lower compartment through the ice condenser into the upper compartment.
Following depressurization, reverse flow along this path will also occur.
As a supplement to this natural flow path, the return air fan system at a total flow rate of 80,000 cubic feet per minute draws suction from various regions of the containment, principally the upper compartment, and discharges into the lower compartment.
This flow rate is sufficient to recirculate the entire lower compartment volume atmosphere (one air change) in the order of 5 minutes, which is a comparatively short time period in relation to the calculated interval of hydrogen release of 22-15
approximately 45 minutes.
Furthermore, the analysis performed with the CLASIX code verifies that the large volumes of a detonable mixture will not exist for the postulated accidents.
TVA has concluded, and the staff concurs, that large detonable mixtures of gas will not be formed inside containment.
Another aspect related to use of deliberate ignition as a mitigation technique for control of excessive hydrogen release is the possible adverse impact of a local detonation on structures'and equipment.
The staff believes that there is potential for small volumes of a detonable mixture to be formed.
We have, therefore, required the applicant to consider the effects of local detonations on structures and equipment.
TVA is in the process of evaluating the con-sequences of detonations and has stated that investigation of this phenomenon will continue.
Nevertheless, TVA has provided an estimate of a pressure loading resulting from a local detonation with a peak pressure increase of 180 psi.
The staff has evaluated this and find it to be a reasonable estimate of the pressure loading which may be associated with local detonation.
Although l
the pressure from this detonation is high, the time duration for the pressure is brief (on the order of 0.5 milliseconds).
As discussed above, this very short duration loading will not affect containment structural integrity.
The impact of the pressure for this very short duration on the relatively small surface area of effected equipment will similarly not result in serious equipment damage.
We believe that needed equipment will be able to functionally survive.
The equipment necessary to assure functionability rf the IDIS and to validate the analytical predictions regarding precluding breach of containment are specifically the igniters, the air return fans, arid containment sprays, and related power and control cables.
The function of these systems in assuring containment protection is discussed above.
The first two items are contained on TVA's list.
The third item, the containment sprays, contains only piping and check valves in containment.
TVA also includes on its list of key equip-ment the hydrogen analyzers used to deal with the conseqance of hydrogen generated as a result of a LOCA not a severely damaged core.
During our review we have considered the need to ensure the functionality of the hydrogen recombiners for long term hydrogen control.
TVA is evaluating its survivability.
The staff will continue to evaluate this matter and resolve it to its satisfaction.
The following minimum systems and components are required to maintain degraded core in a stable safe shutdown condition following H burn:
2 1.
RCS instruments for pressure and temperature 3.
ECCS 4.
PORVs or PORV block valves (including operators) 5.
Wiring and power cabling associated with these systems and components The essential equipment inside containment associated with these systems needed to assure safe shutdown is:
22-16 l
i steam generator, pressurizer, and sump level transmitters air return fan motors hydrogen analyzers hot leg RTD's gasket and seals for flanges, electrical boxes, air locks, and the equipment hatch l
l hydrogen igniters electrical penetrations containment isolation valves including hydrogen sample valves FCV-43-201, 202, 207, and 208 wrapped cable exposed cable - core exist thermocouples exposed cable cold leg RTDs junction boxes reactor coolant system pressure indicator pressurizer power operated relief valve block valves, and related power and control cables.
In addition, there are pioing runs in containment associated with other safe shutdown systems, but no equipment other than the types included in the foregoing list.
TVA's list of key equipment to assure safe chutdown includes all items on ti.e staff's list except the Reactor Coolant System Pressure Indicator and Pressurizer Power Operated Relief Valve Block Valves.
Based upon discussions with TVA, we understand that the pressure indi.ators will survive the effects of hydrogen ignition.
It is our positior that either the relief or the block valve is required and therefore must survise the effects of ignition.
TVA has indicated that this valve is not required; however, they are evaluating its survivability.
The staff will continue to evaluate this matter and resolve it to its satisfaction.
On January 21 and 22, 1981, the staff completed a visit to the plant site and reviewed all data relative to the equipment discussed above, and in addition has performed an onsite review of selected equipment.
The staff has also reviewed information obtained from the testing conducted at FENWAL with regard to the hydrogen control system and the expected containment environment.
As a result of the information submitted by the licensee and the onsite review, the staff conclusions are as follows:
22-17
1.
Because the current version of the CLASIX code neglects heat removal to the structural heat sinks and therefore, overpredicts the containment atmosphere temperature, TVA has performed revised calculations to estimate the temperature response of equipment.
In this regard TVA has estimated both the long-term and short-term temperature effects of hydrogen burning on equipment required for cold shutdown after a degraded core event and the prevention of breach of containment.
The staff discussed the calcula-tions at length with TVA and concludes that the approach used by TVA as well as the assumptions used are reasonable, and provide reasonable results.
As a result of these calculations, the temperature response of this equipment was seen to be comparable to that experienced during a LOCA or MSLB.
Based on our extensive reviews of equipment qualification for LOCA and MSLB, we conclude that there is reasonable assurance that all needed equipment will survive in the degraded core accident environ-ment assuming the occurrence of hydrogen burns.
2.
Regarding the operability of the air return fans and the related power and control cables which are necessary to prevent breach of containment in the event of a hydrogen burn, TVA has provided information to demonstrate that the short lived effects of a hydrogen burn are minimal.
The staff has examined this information and agrees with TVA's conclusions.
Inspection of the fan motor reveals that the motors are massive (1300 pounds) fin cooled casings and are sealed.
3.
Regarding the igniters, their ability to function in a burning hydrogen atmosphere is discussed in other sections of this supplement.
4.
Regarding the containment sprays, only piping and check valves are located in containment.
All other EquiDGent of this system is outside Containment.
In summary, based on the information i,upplied by TVA, the staff's review of this information, and the staff's on-site visit to the facility, the staff concludes that there is reasonable assurance that this equipment will survive repeated hydrogen burns and function properly to mitigate the consequences of a degraded core event and prevent breach of containment.
We will report on the block valve upon completion of our review with TVA.
The licensee plans to meet with the staff to finaliz aquirements for further investigation to confirm the above conclusions.
The 3sults of these investiga-tions will be provided to the staff for review in May 1981.
This will include improved calculational methods for containment temperatures, confirmatory tests on selected equipment exposed to hydrogen burns and new calculations to predict differences between expected equipment temperature environments and containment temperatures.
Sequoyah Containment and Structural Capacity The use of the IDIS has raised two questions which are addressed herein.
First, will a local hydrogen burn within the ice condenser area damage the air return duct. system so that the insulation between the ducts and the containment shell would be exposed to hydrogen burning? Second, will local hydrogen burning produced by activation of the IDIS pose any threat to the integrity of-the primary containment structure?
22-18 i
l The Sequoyah containment structure is a free standing steel containment shell which is composed of a right circular cylinder with a hemispherical dome.
The cylinder is 115 feet in diameter and 105 feet tall and is comprised of eight different thicknesses of steel plates,1-1/4 inches at the base and stepping down to 1/2 inch at the top.
The dome is comprised of four thicknesses, 7/16 inch at the intersection with the cylinder and stepping up to 15/16 inches at the top.
The dome and the cylinder are reinforced with horizontal stiffener rings of steel plates placed about 10 feet apart vertically.
Also there are vertical stringers spaced around the cylinder and spanning from the base slab up to the top stiffener plate on the dome.
The air return duct system, which provides refrigeration for the ice bed, runs vertically along the containment shell.
Between the duct and the containment shell is a layer cf foam insulation which is encapsulated by the containment
{
shell and the duct system.
Figure 22.2-2A shows the ice condenser air ducts l
and insulation and Figure 22.2-2B shows a plan view of the duct system at the I
containment shell.
TVA estimates that there is approximately 30,000 pounds of polyurethane form in the sealed panels adjacent to the duct.
TVA reports that the foam insulation can withstand a compressive pressure of 30 psi.
The maximum pressure on the outside of the duct work from a hydrogen burn in the area is estimated by TVA to be between 13 to 15 psig.
The duct system was originally designed to withstand an external pressure of 15 psig and was actually tested in compression of 19 psig.
Therefore, it can be concluded that the duct system will not be damaged by a hydrogen burn in the ice condenser compartment and that the foam insulation will not be exposed.
Independent analyses were conducted by several organizations and concentrated on predicting the ultimate static pressure for the steel containment shell.
The containment static pressure capacity is based on the a.'tual material properties for the shell as shown in Table 22.2-3.
Based on these analyses and the statistical variations of the material properties, the staff has con-cluded that the lower bound of the containment static pressure-retaining capacity is 36 psig due to an overall pressurization of the whole containment.
With respect to potential hydrogen detonations, TVA has initiated its study by analyzing the proposed response to a pressure profile for a local detonation assuming a 6-foot-diameter gas cloud which results in peak overpressure in a local area of 180 psig with a rise to this peak pressure in 0.10 milliseconds and a dissipation back to zero pressure within 0.5 milliseconds.
An analysis has also been performed by the staff, approximating the response of the con-tainment to the aforementioned local detonation by modeling the containment as responding in a breathing mode.
Damping was conservatively neglected.
The results of this analysis indicate that the effective equivalent static pressure on the steel containment is approximately 14 psig.
This pressure compares favorably with the 36 psig static capa'oility of containment.
The above-described approximate analysis shows that the structural integrity of the primary containment would be maintained in the event of a local hydro-gen detonation of the type postulated.
However, it is noted that this pulse shape used in this evaluation is very general in nature.
The actual pulse shape will depend on individual events.
Further studies are needed to bound the variation in pulse shapes in order to confirm the finding noted above.
22-19
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I 22-21
TABLE 22.2-3 MEAN AND STANDARD DEVIATION OF MATERIAL PROPERTIES PROPERTY MEAN STANDARD DEVIATION Modialus of elasticity (normal) 29,000 ksi 174 ksi Poisson's Ratio (normal) 0.3 0.009 Yield Stress (lognormal) 47.2 ksi 2.50 ksi Ultimate (lognormal) 66.2 ksi 1.80 ksi Fracture Stress (lognormal) 197 ksi 102 ksi Bolt Yield Stress (lognormal) 105 ksi 2.50 ksi 22-22 l
l l
1
Conclusions On the basis of the analyses and testing performed to date, subject to satis-factory solution of the survivability of the PORV block valves (or the PORVs themselves) the staff concludes that the IDIS w eld serve to reduce the con-sequences of a severely degraded core accident.
In the report dated January 13, 1981, the ACRS concurs with the conclusion that the IDIS will reduce the consequences of a severely degraded core accident.
The operation of the igniter system acting in concert with the heat removal mechanisms in the plant, i.e., the ice bed, sprays and passive heat sinks, would sufficiently reduce, for certain accident scenarios, the increase in containment atmosphere pressure resulting from the burning of hydrogen.
The staff has previously determined that without a reliable ignition system, ice condenser plants could tolerate the burning of only that amount of hydrogen that would be released by approximately a 25 percent zirconium cladding-water reaction.
The analyses presented herein which have credited operation of the IDIS have been performed for an accident which was allowed to proceed to the point where 80 percent of the core cladding reacted.
This increased containment capability to accommodate hydrogen releases represents a substantial improvement over the design basis l
capability.
The increased capacity of the containment to tolerate the energy addition due to hydrogen burning is a rate dependent phenomenon such that a distributed ignition system allows for a more controlled burning of lean hydrogen mixtures in the containment.
Moreover, the actuation of igniters precludes the formation of large volumes of hydrogen gas at a detonable con-centration in the event of large hydrogen releases.
Accordingly, we conclude subject to the reservation on the PORV block valves, (or the PORVs themselves) that TVA has shown by testing and analysis that the IDIS will provide with reasonable assurance protection against breach of containment in the event that a substantial quantity of hydrogen is generated.
TVA has instituted a degraded core task force program in order to systematically study the benefits and options related to mitigation of the consequences of severe accidents which are beyond the current design bases.
The first quarterly report on this program was filed on December 11, 1980.
The NRC staff has also begun evaluating future design requirements for accident control as part of a process which is anticipated to continue for several years.
Because many of the proposed mitigation concepts represent state-of-the-art technology, the staff expects that information needs will evolve throughout this process.
The staff therefore believes that it is prudent to continue the detailed review of the distributed ignition system for the Sequoyah plant.
As part of this ongoing effort, the staff will continue to consider issues related to controlled ignition and evaluation of this technique.
The staff believes that iseres such as small local detonations should be studied further before a long-term commitment to deliberate ignition is endorsed even though our current evaluation leads us to conclude that these issues do not represent an apparent deterrent to use of such a system.
Consideration of various accident scenarios should also be a part of the staff and licensee efforts over the next year.
Furthermore, combustion phenomena in a post-accident containment environment will be studied by both the industry and the staff over the ensuing year, and this future study demands our consideration before final approval of the Sequoyah distributed ignition system can be reached.
22-23
me APPENDIX D
[o w%
UNITED STATES ks.h.f[,i 8h NUCLEAR REGU'.ATORY COMMISSION
^
{ ;'Qh/ f
- /
ACVisORY COM.*.1ITTEE ON REACTOR SAFEGUARDS e,,
.usmucron. o. c. zesss January 13, 1981 l
l Honorable John F. Ahearne Chairman U.S. Nuclear Regulatory Commission l
Washington, D. C. 20555
SUBJECT:
REPORT CN THE SEQUOYAH NUCLEAR PCWER PLANT, UNITS 1 Atl0 2
Dear Dr. Ahearne:
Ouring the 249th meeting of the ACRS, January 3-10, 1981, we discussed the NRC Staff's review of the interim hydrogen control system proposed for use in the Sequoyah Nuclear Power Plant, Unit 1.
Tnis matter was also discussed at a Subcoccittee meeting on January 6,1981. We have previously cccrented on this subject in our report dated July 15, 1980 and in two reports cated September 3, 1980.
In this previous correspondence we indicated that distributed ignition systems of the type being considered for use in the Sequoyah plant could pro-vide an improved capability for controlling the burning of a large amount of hydrogen and that the use of such a system would probably reduce risk.
We now believe that the results of analyses and tests which we have discussed with the NRC Staff and the Tennessee Valley Authority (TVA) support these conclusions. The NRC Staff and TVA are continuing to work together to resolve the issue of the survivability of the equipment within containment which is important to safety. Although much further study is needed to determine the ability of the many essential components to continue to operate after exposure to the conditions imposed by possible hydrogen burning, the conditions imposed will not be aggravated by the operation of the ignition system, and in all probability will be less severe. We wish to be kept informed of the NRC Staff's and TVA's progress in this work.
We concur with the NRC Staff recommendation to allow the operation of the Interim Distributed Ignition System and believe that this systen will provide improved protection against creach of containment in the event that a substan-tial quantity of hydrogen is generated. We recommend that the NRC Staff and TVA continue their efforts to describe the performance characteristics of the system over a broader range of conditions.
Sinc erely,
J. Carson Mark Chairman D-1
NU$[G-00Y1
'U""
U.S. NUCLE AR REGULATORY COMMIS$10N
,7 BIBLIOGRAPHIC DATA SHEET Supplement No. 4 4 TITLE AND SUBTSTLE (Add L'o/ume No,. rIwormriate)
- 2. (Leave b/whi Safety Evaluation Report related to operation of Sequoyah Nuclear Plant, Units 1 and 2, Docket Nos.
3 RECIPIENT'S ACCESSION NO.
50-327 and 50-328 1
- 7. AUTHOR (S)
- 5. DATE REPORT COMPLETED l YEAR MONTH January 1981
- 9. PERFORMING ORGANIZATION N AME AND MAILING ADORESS (/nclude Isp Code /
DATE REPORT ISSUED lYE^a U.S. Nuclear Regulatory Commission MONTH Office of Nuclear Reactor Regulation January 1981 Washington, D. C.
20555 6 Iteeve uma*>
i 8 (Leave danki
- 12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include lip Codel g
Same as 9 above li. CONTRACT NO.
l i
- 13. TYPE OF REPORT PE RIOD COVE RED //nclusive dates) i Safety Evaluation Report, Supplement 4
- 15. SUPPLEMENTARY NOTES 14- (Leave D/d"*/
Pertains to Docket Nos. 50-327 and 50-328
- 16. ABSTRACT (200 words or less/
o Supplement No. 4 to the Safety Evaluation Report of Tennessee Valley Authority's application for licenses to operate its Sequoyah Nuclear Plant, Units 1 and 2, located in Hamilton County, Tennessee, has been 2 prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
This supplement provides further information on the hydrogen control measures for Unit 1.
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17 KE Y WOHOS AND DOCUMENT AN ALYSIS 17a DESCRIPTORS 17b IDENTIFIE RS OPEN ENDE D TERMS
21 NO C' PAGE S Unclassified Unlimited SE qu Ri T Y.cpSS (ra4 o,.,1 22 Price nClaSs1Iled s
! NHC FORV 33$ (? 77)
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