ML20059G879
| ML20059G879 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 08/29/1990 |
| From: | Chan T Office of Nuclear Reactor Regulation |
| To: | Lyster M CLEVELAND ELECTRIC ILLUMINATING CO. |
| References | |
| TAC-66121, NUDOCS 9009140004 | |
| Download: ML20059G879 (1) | |
Text
{{#Wiki_filter:. L August 29, 1990-Docket No. 50-440-Mr. Michael D.' Lyster, Vice President 4 Nuclear Group
-lho_ Cleveland Electric Illuminating-Company 10 Center Road Perry,~ Ohio -44081 -
Dear Mr. Lyster:
SUBJECT:
. PERRY NUCLEAR POWER PLANT, UNIT'1 - HYDROGEN CONTROL FINAL' ANALYSIS REQUIRED BY 10 CFR 50.44 (TAC NO. 66121) , !
The NRC Control staff.has Owners Group comp (HC0G): Topical Report HGN-112-NP.leted'its Enclosed is a' copy review of the Ma of the staff's Safety. Evaluation Report (SER) transmitted to HCOG by letter dated August 6, 1990. As discussed during an August 23, 1990-telephone conference'with'
~
Mr.. B. S. Ferrell of your staff, CEI is connitted to provide, within 6
, 1 months of SER' issuance, the final' analysis of the Perry Nuclear Power Pla'nt . ! '
combustible. gas control system required by 10 CFR 50.44. Accordingly, you' i are requested to provide the final analysis by March 1,1991, addressing- ! each of the key elements identified in the staff's SER,~Section 8.0. The reporting and/or recordkeeping requirements contained in this letter '1 affect fewer than ten respondents; therefore, OMB clearance is not required under P.L. 96-511. Sincerely, 1 Original signed by James R. Hall for/ '
-Terence L. Chan, Senior Project Manager Project Directorate III-3 Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation 4
Enclosure:
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$k.,..* August 6, 1990 Mr. 3. R. _ Langley Project Manager, Mark III' Containment i
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. Hydrogen Control Owners Group (HCOG) c/o Gulf States Utilities ,i North Access Road at Highway 61 St. Francisv111e, LA 70775
Dear Mr. Langley:
SUBJECT:
ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORT TITLED, , !
" GENERIC HYDROGEN CONTROL INFORMATION FOR BWR-6 MARK.III CONTAINMENTS", HGN-112-NP We have completed our review of the subject topical report submitted by'your
' letter dated February 23, 1987.
We find the: report acceptable for referencing in licensee analyses of hydrogen control systems for BWR Mark III containments under the limitations delineated ; in the report and its references and the associated NRC evaluation, which is enclosed. The evaluation defines the basis for acceptance of the report.. Furthemore, each licensee should provide a plant-specific analysis and an assessment of the need for'an independent power supply for the hydrogen i ignition system. The plant-specific analysis may use test data described in I the topical report to confim that the equipment necessary to establish and-maintain safe shutdown and to maintain containment integrity will be capable of perfoming their functions during and after exposure to the environmental conditions created by the hydrogen in all, credible severe accident scenarios. > 1 Recent risk studies reported in NUREG-1150 have shown that the overall core melt frequency for one Mark III plant (the Grand-Gulf-Nuclear. Station) is
'- very low, i .e. ,1E-6/ year. However, a potential vulnerability for Mark III ;
plants involves station blackout (SBO), during which the igniters would be - ' inoperable; and this condition appears to dominate the residual risk from severe accident in the Mark III plants. Under SB0 conditions, a detonable mixture of hydrogen could develop which could be ignited upon restoration of power resulting in loss of containment integrity. On the basis of a separate evaluation of this possibility in the context of the NRC staff Containment. Perfomance Improvement (CPI) program, the staff has recommended that the : vulnerability to interruption of power to the hydrogen igniters be evaluated further on a plant-specific basis as part of the Individual Plant Examinations (IPEs) of the Mark III plants. The staff has requested that the licensees consider this issue as part of the IPE in Generic Letter 88-20 Supplement 3. x n+ ! ! y' \ h(\ l
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July 27,1990
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l Mr'. J. R. Langley V We do not intend to repeat our review of the matters described in the report I and found acceptable when the report is referenced in licensee requests for approval of final analyses of the hydrogen control system, except to ensure !
. that the material presented is applicable to the specific plant involved. 1 Our acceptance applies only to the matters described in the report and its references.
In accordance with procedures established in NUREG-0390, we request that HCOG
! submit to the NRC accepted versions of this report within three months of .
receipt of this letter. The accepted versions should incorporate this letter I and the enclosed evaluation between the title page and the abstract. The I accepted versions should also incorporate as appendices those references used i as a basis for the staff's evaluation. The accepted version should include an l
-A (designating accepted) following the report identification number. Your submittal should include an application for withholding the proprietary information accompanied by an effidavit meeting the requirements of 10 CFR 2.790(b). This final report submittal should also include a non-proprietary version of the proprietary reports referenced and incorporated in the approved-topical--report and intended to be employed as a part of a licensee application. i Should our criteria or regulations change such that our conclusions as to !
the acceptability of the report are invalidated, HCOG and/or the licensees referencing the topical report will be expected to revise and resubmit their respective documentation, or submit justification for the continued effective applicability of the topical report without revision of their respective-documentation. Sincere , Ashok (OS&
.7 C. Thadani, Director Diviscon of Systems Technology Office of Nuclear Reactor Regulation l
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-THE GENERIC; SAFETY-EVALMTION REPORT RELATING 70:
THE MARK III CONTAll# LENT HYDROGEN CONTROL - E . 6
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-CONTENTS Page-1 INTRODUCTION AND' BACKGROUND.................................... 1-2 - GENERAL DESCRIPTION OF THE HYDROGEN IGNIT!0N' SYSTEM............ 3 _
3 - COMBUSTION / IGNITER-TESTING................'.................... . 6-3.1 Small-Scale Tests..........................................~6' 3.2 Quarter-Scale Test Facility............................... 7. *
- 3. 2.1 - Scal in g Methodol ogy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 ,
~ 3.2.2 Quarter-Scale Testing Approach................... 10 3.2.3- Q ua rter-Scal e - Tes t Resul ts . . . . . . . . . . . . . . . . . . . . . . . 11:
4 ' CONTAINMENT STRUCTURAL. CAPACITY- . 14 5 DEGRADED CORE EVENTS AND HYDROGEN GENERATION................... 15= 5.1 Introduction.............................................. 15 5 . 2 . E v a l u a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ 18 , 5.2.1 Acceptabit GE Sequence . . . . . . . . . . . . . . . . . . . . . . . . . . 19 - , a 5.2.1.1 . The HCOG 's : Ba se-Ca se Scena rio. . . . . . . . . . . . . . 19- t 5.2.1.2 Station Blackout and'NUREG-1150............ 20 5.2.1.3 TBU the " Acceptable Sequence". . . . . . . . . . . . . . 20 5.2.1.4 Hydrogen Generation Profiles............... _21 5.2.1.5 Non-Mechanistic Hydrogen Release Profile... 22- - 5.2.2 BWRCoreHeatupCode(BWRCHUC)...................25 5.2.2.1 I n troduc t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 5.2.2.2 Phenomenological Assumptions........r...... 26-
. 5.2.2.3 Steam Production........................... 26, j 5.2.2.4 Hydrogen Generation.................. . ..... 28. 1 5.3 Summary and Conclusions................................... 29' l
6 CONTAINMENT RESPONSE - ANALYTICAL MODELING..................... 30 L 6.1 Localized-Combustion /CLASIX-3..........................r .. 30 6.2 Containment Pressure and Tem 6 s3 Drywell Analysi s . . . . . . . . . ........................ . . . pera ture Calcula tions. . . . . . . . . 133 ' 33 6.4 Existence of Drywell D.iffusion F1ames..................... 34 . . . . . . . f 1 Mark III SER i 4 --v -w w yy-,,.-wr sw_- e e, , --- ar , . _ _ - ___ _ __m__m____m_m -__m______.____m _m__m____ - _ _ _ _ _ _ _ - _ _ _
6 c-i' 1 . CONTENTS (Cont'd)._ 7-
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SURVIVABILITY OF' ESSENTIAL EQUIPMENT........................... 38-
-7.1 Identification of Essential Equipment...................... 38J 7.2 Generic ~ Equipment. Survivability. Analysis (Localized.
C omb u s t i on ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~. . 3 9 '
. 7.3 Diffusion Flame Thermal Environment Methodology........... 411
- 7. 4 jS pray . Ava il a bil i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ; 4 2 l 7.5 Pressure Effects.......................................... 43' 7.6 Detonations............................................... 43! -!
8 CONCLUSION............................'......................... 43 APPENDICES A GENERIC HYDROGEN IGNITION SYSTEM ~ TECHNICAL SPECIFICATIONS B MARK-III COMBUSTIBLE GAS CONTROL EMERGENCY PROCEDURE GUIDELINE C . BIBLIOGRAPHY- S D ACRONYM LIST LIST OF FIGURES' 2.1_ General Hydrogen Ignitor Assembly..................: .......... 4 l 3.1 Elevation Views of'the Quarter-Scale Test Facility.............'8 3.2 Facil i ty Confi guration Tes ts S. 01-S.11. . . . . . . . . . . . . . . . . . . . . . . . . 9 4.1 Typi cal Ma rk III Conta inment Co,nfituration. . . . . . . . . . . . . . . . . . . . . 17 1 5.1 Hydrogen Genera tion Ra te (150 GPM Reflood) . . . . . . . . . . . . . . . . . . . . . 23 5.2 Hydrogen Generation Rate (5000. GPM Reflood). . ... . . ... .. . ...... . 24 6.1 6.2 Perry CLASIX-3 Mode 1........................................... SORV Wi th No Spray-Wetwel l Tempera tu re. . . . . . . . . . . . . . . . . . . . . . . . . 36 35 6.3 SORV With No Spray-Wetwell Pressure............................ 37=- LIST OF TABLES
'4 1.
Comparison of BWR Mark III Containment Characteristics.........16 ' 3 5.1 Hydrogen Rel ease Profiles . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . 22 < Mark III SER 11 1 l
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1- INTRODUCTION AND BACKGROUND 4 Following a loss-of-coolant accident in a light-water reactor (WR) plant, com-bustible gases, principally hydrogen, may accumulate inside the primary reactor
' containment as a result'of (1) metal-water reaction involving the fuel element cladding; (2) the radiolytic decomposition of the water in both the reactor core:
and the containment sump; (3) the corrosion of certain materials within the. containment by sprays; and (4) any synergistic chemical, thenna1, and radio-lytic effects of post-accident environmental conditions on protective' coatings and electric cable insulation. To provide protection against this possible h Title 10 of the Code of Federal Regulations10.(ydrogenCFR) Section as 50.44, a result 'of an accident,
" Standards for combustible gas control system in light-water-cooled power reactors," and GDC 41, " Containment atmosphere cleanup," Appendix A to 10 CFR Part 50, require )
i that~ systems be provided to control hydrogen concentrations in the containment atmosphere following a postulated accident to ensure that containment integrity is maintained. Conventional hydrogen control systems (e.g., hydrepen recombiners). historically have been installed to provide:the capability to control the rela . tively low rate of hydrogen accumulation (or oxygen accumulation in inerted containments) resulting from radiclytic decomposition of water, corrosion of metals inside containment, and environmental effects on coatings and insulation. However, the net free velame inside containment (or inerting of the containment volume) is used to control the rapid hydrogen production resulting from a metal-water reaction of the fuel cladding. That is, the containment volume is large enough so that the hydrogen generated early would not reach the lower limit of flammability (or the inerting would prevent combustible' mixtures). The rationale-for this approach is that the rate of hydrogen release resulting from cladding reaction was assumed to be too rapid (on the order of minutes) following a pos-tulated accident to allow for an active control system. Thus, hydrogen control systems (recombiners) would only be actuated later in the event to control the i slow hydrogen /or oxygen release associated with'radiolysis and-reaction of materials inside containment. To quantify the metal-water source for design-basis accidents.10 CFR 50.44
- codified in October 1978, requires that the combustible gas control system be- l t
capable of handling the hydrogen generated from five times the amount calcula- l l ted in demonstrating compliance with 10 CFR 50.46 or the amount corresponding ! l to the tot 61, reaction of the cladding to a depth of 0.23 mils, whichever amount I was greater. l- Typically, tMs would translate to a 1 to 5% metal-water reaction l of the active clad. However, the accident at Three Mile Island Unit 2 (TM1-2) on March 28, 1979, resulted in a metal-water reaction that involved approximately 451'of the fuel cladding (i.e., about 990 lbs), which resulted in hydrogen generation well in excess of the amounts specified in 10 CFR 50.44. The combustion of tu s hydrogen produced a significant pressure spike inside containment. As a result, it became apparent that additional design measures were needed to handle larger hydrogen releases, particularly for smaller volume containments and those with Mark 111 SER 1
l l o lower design pressures. The Nuclear Regulatory Commission (NRC) determined that a rulemaking proceeding should be undertaken to define the manner and extent to which hydrogen evolution and other effects of a degraded core need to be
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considered in plant design. An advance notice of the rulemaking proceeding on degraded core issues was published in the Federal Register on October 2,1980. In addition, a'new unresolved safety issue was insti;uted (A-48, " Hydrogen Con-I trol Measures and Effects of Hydrogen Burns on Safety Equipment") to evaluate this new concern. To formalize its requirements-for additional hydrogen control measures to deal with degraded core accidents affecting pressurized-water reactors (PWRs) with ice condenser containments and BWRs with Mark III containments, the NRC pub-lished an amendment to the hydrogen rule (10 CFR 50.44) on January 25, 1985 . (50 FR 3498). The amended rule required that a hydrogen control system be pro-viced and that the system be capable of accommodating, without loss of contain-ment structural integrity, the amount of hydrogen generated from a metal-water reaction of ap to 75% of the active fuel cladding. In addition, systems and components necessary to establish and maintain safe shutdown must be capable of performing their function regardless of hydrogen burning. Pursuant to the provisions of the rule, each utilit ment has installed a hydrogen ignition system.(HIS)y and with a Mark submitted III contain-- a preliminary analysis and a schedule for meeting the full requirements of the rule. The affected plants with Mark III containments are Grand Gulf Nuclear Station, River Bend Station, Perry Nuclear Power Plant, and Clinton Power Station. The staff's interim evaluations of these initial plant responses are documented in supplements to the safety evaluation reports'for each of these plants (NUREG-0831, NUREG-0989,NUREG-0887,andNUREG-0853,respectively). These responses were aided by the efforts of the Mark III Containment Hydrogen Control Owners Group (HCOG), formed in May 1981 by the utilities with Mark III containments to collectively perform testing and analyses to demonstrate the effectiveness and reliability of the hydrogen ignition systems. h addition, each licensee with a Mark III containment is required to p avide a final analysis [10 CFR 50.44(c)(3)(vii)(B)] to confirm the conclusions of the preUminary analysis and/or, if necessary, to institute modifications to ensure compliance with the rule. The scope of this analysis is specified in 10 CFR 50.44(c)(3) (vi)(B). The generic findings from the HCOG's program will be l utilized for this final analysis, supplemented by plant-specif1c design ! considerations as addressed in the licensee's IPE program. The tollowing staff evaluation focuses on the assessment of the completed gen-eric testing and analyses performed by the HCOG in support of the plant unique analysis. The HCOG activities have been summarized in a topical report trans-mitted by letter dated February 23, 1987, correspondence identification HGN-112-NP,
" Generic -Hydrogen Control Information for BWR-6 Mark III Containments." The topical report is a summary document of all of the individual generic submittals that have been sent to the staff by the HCOG. It should be noted that HCOG correspondence identification designators with a "P" suffix (HGN-XXX-P) are proprietary to HCOG. Whereas, those without a suffix or with an "NP" suffix are nonproprietary.
l l l l Mark III SER 2
As part of the review of the HCOG program, the staff obtained technical assist-- ance from the Sandia National Laboratory (SNL),:the-principal contractor for the NRC research program on hydrogen control and combustion phenomena. _ SNL provided - . the NRC with an independent assessment of technical issues contained in selected HCOG submittals pertaining to hydrogen behavior. The staff evaluation of the generic considerations of the hydrogen control system for the Mark III containment can best be understood if categorized as follows: general description of the hydrogen ignition system combustion and igniter testing containment structural capacity " degraded core events and h drogen generation containment response-anal ical modeling survivability of essentia equipment overall conclusions Therefore, the.following_ discussion will follow this general outline; 2 GENERAL DESCRIPTION OF THE HYDROGEN IGNITION SYSTEM The regulation,10 CFR 50.44(c)(3)(iv)(A), states: Each licensee with a boiling light-water nuclear power reactor with a Mark III type of containoent..., shall provide its nuclear power reactor with a hydrogen control system justified by.a suitable pro-gram of experiment and analysis. The hydrogen control system must be capable of handling without. loss of containment structural integrity. an amount of hydrogen equivalent to that generated from a metal-water reaction involving 75% of the fuel cladding surrounding the active fuel region (excludingthecladdingsurroundingtheplenumvolume]. The concept employed by the licensees with a Mark III containment, and similarly the ice condenser licensees, is to intentionally ignite hydrogen generated inside containment. This method precludes buildup of relatively high'concen-trations of hydrogen during degraded core accident scenarios. To accomplish this early ignition a hydrogen ignition system (HIS) is installed in each of the four plants with Mark III containments. The HIS is a system which consists of approximately 100 igniter assemblies distributed throughout the drywell and containment regions. The main element of the igniter assembly is the Model 7G thermal igniter glow plug (commonly used in diesel engines) manufactured by the General Motors AC' Division. Each Mark III containment has an identical igniter assembly design. Each igniter is powered directly from a 120/12-V stepdown transformer and de-signed to provide a minimum surface temperature of 1700'F. The igniter assem-bly (see Figure 2.1) consists of a 1/8-inch-thick stainless steel box that con-tains the transformer and all electrical connections and is manufactured by the Power Systems Division of Morrison Knudsen. Igniter assemblies are Class 1E, i seismic Category I, and meet the requirements of the Institute of Electrical I and Electronics Engineers (IEEE) Std 323-1974 and NUREG-0588 for environmental l qualification. l l Mark III SER 3 l
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The: igniter assemblies are diYided into 'two groups; each group being powered from' a separate Class IE division power supply.. The intent is to have at least two l igniters located in each enclosed volume or area within'the containment that could - be subject to possible hydrogen pocketing, and each igniter be powered from a - separate power division. In open areas within the containment and drywell regions, igniter' assemblies at the same elevation are designed to have alternating power ' division sources. 11 ' safety feature (ESF)gniters are separated power divisions by about 30 are operable'or by about feet when 60 feet both when engineered one power-division is inoperable. Igniter >1acement is designed to be more widely spaced in ( the large open regions,-such as a >ove the refueling floor, and in the lower regions- , of the drywell that are subject to flooding. Requirements of placement as wel' as - l other parameters of the system are contained in the technical specificatiens for . the hydrogen ignition system, as proposed by the HCOG, and addressed in Appendix A-of this re, ort.
.Each igniter power division has a corresponding onsite emergency diesel generator.
Incorpvration of the emergency diesels-into the design addresses the question-of igniter power for many sequences,- but not for station blackout conditions. Under blackout conditions, the HIS would not be operable. On the basis of a separate evaluation of this possibility in the context of.the NRC Containment Perfonnance Improvement (CPI) program. the staff has recommended that the vulnerability to interruption of power to the hydrogen igniters be evaluated further on a plant-specific basis as part of the Individual Plant Examinations of th: W rk III plants. (See Generic Letter 88-20, Supplement 3) The HIS is designed for manual actuation from the main control room. Actuation by the operator is required by plant emergency procedures when the reactor pres-sure vessel (RPV) water level reaches the top of active fuel-(TAF). The proposed combustible gas control emergency procedures for plants with' Mark III contain-ments are further discussed in Appendix B. By letter dated March 5,1986 (HGN-073), the HCOG provided the following justi-- fication to support manual actuation. Actuation-is linked to indication of the RPV water level, which is a key safety parameter and is: closely monitored by the operators. Also, HIS actuation requires only the positioning of two hand-switches. Furthennore, operators should not hesitate to energize the system during accident scenarios in which the hydrogen threat is uncertain or. marginal, because there would be no adverse effect on the plant as a result of unnecessary igniter actuation. Time available to actuate.the HIS is the other significant parameter. Based on the hydrogen events considered, the HCOG estimated this time to be approximately no less than 25 minutes;-that is, after the water level reaches TAF to the lower hydrogen flaninability limit reached in the wetwell volume. The HCOG also noted that hydrogen would migrate to the upper portions of the containment before the wetwell reaches hydrogen combustion conditions. This effect was seen in the quarter-scale tests. Therefore, the time interval
.is expected to be somewhat greater than 25 minutes.
Because (1) manual actuation is a simple task, (2) the operator has sufficient I time to perfonn the task. (3) and there are no negative effects if the system is inadvertently or unnecessarily actuated, the staff finds manual operator actuation acceptable. l Mark III SER 5 t
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- ..- , j HIS Design Assessment The: staff finds these hydrogen igniter systems currently installed in the -
plants with a Mark III containment to be acceptable, with the coveat thatLthe: vulnerability to interruption of power to the hydrogen igniters-be further ' evaluated on a plant-specific. basis as part of IPEs of the Mark III plants. (See Supplement No. 3 to Generic Letter 88-20) - l 3 COMBUSTION / IGNITER TESTING j Numerour research programs-have been conducted since 1980 to 'better understand hydrogen combustion behavior and the performance of ignition devices. Because - these programs were varied in' scope. Sandia National Laboratory (SNL) summarized the findings of recent hydrogen combustion test programs in NUREG/CR-5079. This report also provides additional background information and insights related to hydrogen behavior.
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The specific test' programs considered necessary by HCOG to support the unique plant characteristics of a Mark III containment are discussed below. 3.1 _Small-Scale Tests 4 Small-scale hydrogen combustion tests were performed at Whiteshell' Laboratories and^ documented by the HCOG.in a letter dated lune 7.'1984 (HGN-017-NP). The-program was intended to investigate ignition and combustion behavior of mixtures predominantly composed of hydrogen and steam, i.e.. with limited available oxygen. This condition may exist for a postulated drywell break event in which air is initially swept from the drywell and then later reintroduced into the~ steam- , hydrogen environment. These tests confirmed that such hydrogen-air-steam wix-tures can be successfully ignited as long asithe oxygen concentration exceeds approximately 5-6 volume percent. . A 1/20th-scale Mark III hydrogen combustion program was conducted by Acurex ' Corporation and documented by.the HCOG in a letter dated February 9. 1984 (HGN-014-NP). The objective of this program was to provide a visual record of hydrogen combustion behavior in a 360-degree model of a Mark III - containment. Modelling included the suppression pool and major blockages in the annular region between the drywell and outer containment walls. Hydrogen , was admitted through simulated quenchers and/or vents into.the suppression pool and ignited by prototypical ignition sources. l The most important result obtained from the 1/20-scale test was the confirmation of continuous hydrogen burning in the form of steady diffusion flames above the L ' suppression pool. The significance of this mode of hydrogen burning is the observed
severe thermal loads that occur near the diffusion flames and could l
ten the integrity of the containment and equipment. Diffusion flames were wM when hydrogen flow rates of 0.4 lb/sec (full-scale equivalent) or l l l Mark III SER 6
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greater. were used. ! Combustion was initiated by the ignitors and rapidly pr agated to the pool surface and formed steady diffusion flames that werei an op-choredtheathydrogen. released the surface of the pool and located above the submerged sparge ' gen injection flow rate was increased, as evidenced by taller temperatures. ' As partsingle 1/5-scale of the 1/20-scale sparger mockup was program to determine the sensitivity constructed. ,a of scaling tests was predicted basedthatonapproximately results of 1 one-half the flame height of what would have bee these tests, it became necessary/20-scale tests was observed. On the basis of for the HCOG to pursue a larger-scale test program to containment. obtain thermal environmental data more representative of a Mark III ; Subsequently, the HCOG undertook an extensive program to better ! define the element of thiconditions that could exist during a degraded core accident . test program, s effort was the quarter-scale Mark III containment combustionA major !
- 3. 21 Quarter-Scale Test Facility t The research quarter-scale program. test program became the major element of the HCOG The primary objective of this program was the investigation and on the characterization suppression pool in a Mark of the environment that could result from diffusive III containment.
Ultimately the infoisation gathered from the ing the survivability quarter-scale of select equipment. test facility (QSTF) would be u, sed in determ 1986; HGN-115-NP, Februarycombustion test results may be found in the 10, 1987; and HGN-121-P, July 22, 1987 The test facility in West Gloucester, Rhode Island. is a quarter-linear-scale, de-model of a M pressures up to 40 psig and consists of an outside tank, 31.5 feet in d high. 49.4 feet high, containing a smaller tank, about 21 fee e and other containments. large blockages simulating the obstructions that exist in the actual Because of the unique features of the-four plants studied modularinterior,when construction needed. of the annular floors is used to modify the vessel At the bottom of the two tanks Several views of the facility are shown,in Figures 3.1 and 3.2the suppression pool is s Facility design features include: o o containment sprays simuisted
- 48) loss-of-coolant-accident (LOCA) vents (top row only, numb simulated spar totalling 24) gers (uniformly spaced every 15 degrees azimuthally and unit coolers (for the River Bend configuration)
The gas velocitiesfacilitygasis heavily concentrationsinstrumented to measure gas and surface temperature cameras are use,d for o visual recor,d. heat fluxes, pressure, and five video Mark III SER 7 l l
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3.2.1 ' Scaling Methodology The theoretical basis for modeling hydrogen flames in the test facility.is. based on Froude scaling. The modeling assumes fully developed, buoyancy dominated, turbulent flows are achieved to preserve the equivalent value for the Froude number in the model and in the full-scale plant. The technique of Froude scaling is supported by numerous experimental demonstrations in the field of fire research. Using this type of modelling full scale was scaled to quarter scale, a 4-to-1 linear scaling resulting in: 32-to-1 reduction in mass and' volume flow rates 64-to 1 reduction in total hydrogen released 2-to-1 reduction in the time scale 2-to-1 reduction in gas velocities 1-to-1 relationship for gas temperatures and gas concentrations Flame heights and global flow patterns also were detemined by Froude scaling. Generally, Froude scaling was used to reasonably and practically design the QSTF (e.g., spray flow and droplet sizes, heat sink thermal characteristics, blockages). However, the following discrepancies were noted: (1) The quarter-scale tests revealed that the insulation used in the facility became wet and its thermal properties departed from dry insulation. ; (2) The QSTF had only 30% of the mass as prescribed by Froude scaling. (3) The scaling method did not rigorously simulate convective and radiative heat losses to structural heat sinks. To assess the impact of these discrepancies, the HCOG provided a comprehensive analysis as documented by HCOG 1etter HGN-085, dated May 5, 1966. The HC0G determined that compensating effects existed in the treatment of heat sinks. Thus, the data obtained from the QSTF does provide a reasonably accurate description of the full-scale thermal environment when extrapolated by Freude scaling. To acsist the staff in the review of this complex matter, SNL 5tudied the subject analysis and submitted various comments. On October 7, 1987, a meeting between the HCOG, SNL, and the staff took place to resolve the SNL's comments. Subse- l i quently, HCOG documented its responses in a letter (HGP-128) dated November 6, 1987. On the basis of the additional information, SNL concluded and the staff concurs that the application of quarter-scale experimental data directly to full-scale equipment survivability can be done conservatively in spite of the above discrepancies. SNL's assessment is documented in~ correspeiidence dated December 23, 1987. 3.2.2 Quarter Scale Testing Approach ._, Tests were performed in the QSTF for the four different plants with a Mark III containment; i.e., the QSTF was customized to reasonably represent the plant-specific characteristics of each Mark III configuration. During the tests, Mark III SER 10
l hydrogen was released through spargers used to simulate the automatic depres-surization system (ADS) and a stuck-open relief valve (50RV) or through the simulated LOCA vents. Two h
- l a low-reflood case (150 gpe)ydrogen release eau and a high-reflood proff(5000
, t were usedA in gps). the facility, discussion of the development of these profiles Is contained in S.:: tion 5.
However before the plant unique or production tests were conducted, a series
- of scoping tests werQ p rformed to assess data repeatability and the significance .
1 of various parameters. Results from these tests forined the basis for developing the final test matrix that were used in the production test program. Also, in i the early development of postulated degraded core events spray availability wasuncertain,thereforetestswereperformedwithandwIthoutspraysactivated. ' In the production tests, each slant had its own specific array of tests focus - on the 50RV locations, the comsination of ADS spargers and LOCA vent releases,ing , and the effect of sprays / coolers. The data obtained from these production tests formed a basis for determining the full-scale thermal environment and became a central element of each licensee's final analysis. This information was used as input to analytical evaluations of equipment therinal response for assessing sur-vivability of critical equipment. Further discussions on the use of this data are contained in Section 7. 3.2.3 Quarter-Scale Test Results
! The scoping' test and partial production test results are summarized and presented
- in the HCOG s correspondence (HGN-098-P andHGN-121-P). These results demor stratedthatthedistributedglowplugIgnitersystemcanprovideaneffective means for limiting accumulation of hydrogen in plants with Mark III containments.
Hydrogen concentrations throughout the facility were maintained near or below 5 i volume percent (dry basis) for all tests and steam concentrations were deter-mined to be about 10 u 5 volume percent for selected tests. Although low hydro- ' gen concentrations were maintr.ined, different types of combustion behavior were observed during the tests, depending on the synergistic conditions. The various
, observed combustion modes are described below. ,
Diffusion Flames When hydrogen was released into selected spargers, hydrogen combustion would initiate as a mild deflagratica or lightoff (pressure rise about I psi) in the wetwell region between the hydr;ulic control unit (HCU) floor and the suppres-sion pool surface and would persist in the form of standing diffusion flames anchored to the pool surface. This was the dominant .tode of combustion and occurred for bulk oxygen concentrations of 8 volume percent (dry) and hydrogen injection rat,es greater than 0.15 lb/sec. It should be noted that the hydrogen flow rates are full-scale equivalent values (i.e. In this j regimeofsteadyflames,combustionwasessentiall a 32:1 incrAase). Foraninjection rate of 1 lb/sec, a flame height of about 8 feet (y full complete. scale) was reached. As the hydrogen injection rate was decreased to about 0.1F lb/secrcombustion became less complete and the flames less stable. As the rate was further de-creased, diffusion flames on the pool surface could no'. be maintained. This point is known as the flame extinguishment limit. Moreover, it was observed that this limit was strongly influenced by background gas concentrations. - l l l Mark III SER 11 l
l \ '... , . . l .. l' To illu'trate s the various relationships, the following was extracted from.the QSTF scoping test report: l l (1) The flame extinguishment limit ranges from ~ 0.07 to - 0.15 -lb/ m see for. ambient hydrogen concentrations below ~ 4.1 volume per-cent dry (high oxygen conditions). (2) The flame extinguishment limit decreases with increasi hydro n. concentration to a minimum between - 0.025 and - 0.03 /sec a a " hydroiten concentration of - 4.5 volume percent (high oxygen condi41ons.) (3) For comparable ambient hydrogen concentrations,.the flame extin-guishment limit is slightly higher at low oxygen conditions.
- Another effect that accompanied hydrogen burning was the formation of bulk t air currents. Horizontal air flow was created above the pool surface allowing i diffusion burning to continue by providing a source of s ygen. Another pattern i j of circulation was the creation of chimneys, which provide for flow to and from the region of burning and exchange flow with the upper containment, that is, ;
hot (upward flow) and cold (downward flow) chimneys. localized Combustion , l Below the flame extinguishment limit, flames on the pool were not observed. : The prevalent burning mode at very low hydrogen release rates has been termed , c ' localized combustion. This phenomenon is characterized as weak flames or volume burning through a marginally combustible hydrogen / air / steam mixture. -This type i of combustion was detected only in regions at or above the HCU floor and concen-trated mostly in chimney areas. This was evident by temperature. measurements; i localized combustion was not observed b Localized combus- , tion appeared to be relatively benign (y video recordings.i.e., less than 250'F at instrume , locations). Burning was more widespread a'nd somewhat more intense at low oxygen conditions and was accorr. , concentrations (i.e., near 5%).panied by slightly higher background hydrogen - 4 The QSTF was oriented to investigate the burning phenomena in the area immedi-1 ately above the suppression pool. As a result, tw instrumentation layout above the HCU floor was not sufficient to provide a detailed mapping of the conditions in that region. Consequently, a rigorous inve tigation of localized combustion was net possible; however, the instrumentation that was present, along with . provided a reasonable characterization HCOG's of analytical the phenomena. effort (see Localized Section combustion 6)Is discussed further in Section 6 as it relates to the analytical methods used by the HC0G. i Secondary Burnino During the quarter-scale testing, an additiora1 combustion phenomenon was ob-served late in one of the testt When the bulk oxygen concentration dropped below 8 volume percent (dry), flames extinguished on the surface of the pool but , fomed at the HCU floor elevation. This type of burning has been termed secon-j dary burning, i .
! Mark III SER 12 j -
I' _-___ ._. _ _ _____ __ _ _. _._ _ _ _ _
i l . During a June 1986 meeting with the staff, HCOG revealed the presence of secondary burning in one of the Perry production tests. This phenomenon was not observed in previous production tests or in the scoping test phase. Until this particu-lar test, only in a scoping test did the containment oxygen concentration drop below 8%. Oxygen concentrations were generally maintained above 8 percent due to a unique need associated with the video coverage. Each of the five video cameras used in the QSTF required a continuous air purge for the camera lenses to prevent a condensation on the lens. This resulted in a continuous inflow of oxygen in the facility, thus precluding the atmospheric oxygen concentration to fall below 8%. However, the camera air purges were not run continuously in the subject test; subsequently, late in this test the oxygen concentration fell below 8%. Additional information is provided by HCOG submittal HGN-106-P dated September 29, 1986, and also discussed in detail in the quarter-scale combustion test report, To present its overall assessment of the significance of secondary burning, HCpG began by addressing the limitations of the QSTF. The QSTF has various physical and practical limitations associated with the investigation of secondary burning. The instrumentation in the facility was geared to define the thenaal environment produced by diffusion flames anchored to the surface of the suppres-sion pool, which is the dominant combustion mode. Therefore, more instrumenta-tion would be needed to investigate burning above the HCU floor. Since all plants with Mark III containments hsve different containment volumes, simulation of the expected oxygen depletion profile for each plant would be difficult. Therefore, HCOG evaluated the need to further consider the secondary burning phenomenon. The following identifies the various factors considered and their
- relationship to the four plants with Mark III containments
(1) Secondary burning is expected to occur over a narrow range of oxygen concentrations, approximately from 6 to 8%. Based on the hydrogen generated by a 75% metal-water reaction, a Mark III containment would experience this oxygen concentration. interval late in the transient or not at all. Assuming the drywell air is not added to the containment inven-tory, a metal-water reaction of 55 to 67% would be reached before the oxygen concentration is expected to fall below 8%. This range applies to three of the four plants. Because of the larger containment volume-to-power ratio, Clinton is not expected to fall below 10% oxygen; thus, secondary burning is not anticipated. When the drywell air inventory is included in the containment region, a metal-water reaction of 67% would be reached for Perry before the oxygen concentration is expected to fall-below 8%; River Bend and Grand Gulf would have already consumed the equivalent hydrogen required by the rule (i.e., 75% of the fuel cladding surrounding the active fuel region). (2) Considerable uncertainty is inherent in predicting the long-term hydrogen profile, especially in the latter phase of the profile (refer toSection5). For example, an alternate accident sequence such as a drywell break sequence, different hydrogen release rates, the.use of the drywell mixing system, or not even reaching a 75% metal-water-reaction value, could reduce or possibly eliminate secondary burning. f Mark III SER 13
( (3)' Also, not all conditions expected to exist in'an actual plant existed in the test when secondary burning was observed. For example, sprays were not activated and the burning on tw HCU floor was in a sector where the hydrogen - release was most concentrated. This sector was in the 45-dt ,ree chimney. in which the SORY was located and the steam tunnel structure would reduce the upward cross-sectional flow area. Therefore. it is expected that these factors' contributed to the locally high concentration of hydrogen that is required for secondary burning. It should also be noted that the overall 7 shape of the flames occupied a relatively small area, forming a flame zone ' near.the corner of the steam tunnel and drywell wall. On the basis of these difforences, the following significant mitigating factors can be inferred to reduce the consequence of secondary burning: (a) Increased turbulence inside containment through spray operation or unit cooling could potentially deley or preclude secondary burning. This was evident to some degree in one of the scoping tests.during which conditions were similar to those during the Perry test where a secondary burning was present. This scoping test had sprays fune-Secondary burning did not occur.tioning Also sprays and the oxygen unit coolers wouldconcentration fell to ap provide cooling to mitigate the consequ,ences of secondary burning if it were to occur. (b) Secondary burning appears to.be extremely localized.. It occurred in 1 the region above the location where three adjacent safety relief valves ($RVs) spargers released hydrogen. Further, secondary burning occupied only a small zone. Because of equipment redundancy and separation, secondary burning is expected to affect only one train of equipment. On the basis of its findings, the HCOG determined furthe.r experimental investi-gation of secondary burning was not necessary. The staff's review of the evidence indicated that secondary burning is not ex-pected to present a significant additional threat and -if this combustion mode were to occur, it is expected that the thermal zone of influence would be lim-ited. Therefore burning is unwarr, anted.the staff agrees However, since thethat further detail redundancy study of secondary of equi i spatiel separation of equipment performing the same function) is pment (i.e., the most l important elemert the staff requests that the Perry River Send, and Grand Gulf licensees (ex,cluding Clinton) confirm that sufficient separation (i.e., i at least a 90' azimuthal displacement) exists between the redundant equipment expected to b,e affected by secondary burning. 4 CONTAINMENT STRUCTURAL CAPACITY The burning cf hydrogen inside containment has the potential to induce pressure excursions in' excess of the containment /drywell design values. To determine the pressure capability of the containment structures, required by 10 CFR 50.44 (c)(3)(iv)(B), each licensee provided its plant-specific analysis for staff review. The details of the staff's evaluations regarding the containment and drywell ultimate capacities are documented in each of the plant's respective SER supplements. Rather than repeating these evaluations,-a brief description of the Mark III containment'will be provided. Mark III SER 14
t . In the Mark !!! containment design, the containment completely surrounds the drywell. At the bottom of the containment, a 360-degree annular suppression pool is located between the containment wall and drywell wall. Below the pool surface, horizontal vents are constructed in the drywell wall. The principal difference between the four plants is in the characteristics of the containment shell, as illustrated in Table 4.1. For Grand Gulf and Clinton, the primary containment is a steel-lined, reinforced concrete structure consisting of a i vertical cylinder and a hemispherical dome top. For River Bend and Perry, the primary containment is a free-standing steel vessel consisting of a vertical cylinder and a torus-spherical dome surrounded by a concrete shield building. The internal containment design pressure of 15 psig is the same for each plant. The ultimate pressure capacity was determined to be about three times design (i.e. , approximately 50-60. psig) for each plant. Since the drywell structure ' is designed to greater pressure values than the containment vessel, the drywall i ultimate capacities also are greater and are not limiting in the forward or reverse direction. The containment pressure capacity taking into consideration limiting containment penetrations, is used as the limlting parameter when evaluating the consequences of hydrogen deflagrations inside containment. Figure 4.1 is an illustration of a Mark III containment configuration. 5 DEGRADED CORE EVENTS AND HYDROGEN GENERATION 5.1 Introduction i To determine the consequences of hydrogen burning, the hydrogen generation ' release must be addressed to establish a representative hydrogen-event (HGE) and define representative hydrogen release profiles. generation The regulation 10 CFR 50.44(c)(3)(vi)(B), specifically requires that the fol-lowing be considered in the analysis: ' (1) large amounts of hydrogen generated a'fter the start of an accident (hydrogen resulting from the metal-water reaction of up to and including 75% of the fuel cladding surrounding the active fuel region, excluding the cladding surrounding the plenum volume); (2) the period of recovery from the degraded condition; (3) accident scenarios that are accepted by the NRC staff and that are accom-panied by sufficient supporting justification to show that they describe the behavior of the reactor system during and following an accident result-ing in a degraded core. The HCOG analyzed two degraded core accident sequences (HCOG transmittals. HGN-003, -006, -018-P, -031, -052, -055, -072, -104-P, -112-NP, -129-P and 4132). The base-case scenario begins with a loss of offsite power, followed by reactor scram, is'olation of both the containment and MSIVs, and power conversion system i unavailability. One diesel generator fails to start and the relief valves cycle l on high reactor pressure as a result of MSIV isolation. Relief valve cycling ! results in one stuck-open relief valve (50RV). The second scenario models a small break in the drywell by using the same total hydrogen and steam release ! histories as the previous case but the predicted hydrogen and steam release is mechanistically split between the drywell and the containment. Mark III SER 15
Table 4.1 Comparison of BWR Mark III containment Characteristics Characteristic Grand Gulf Perry River Bend Clinton Rated thermal output, MWt 3,833 3,579 2.894 2,8 94 Number of fuel bundles 800 748 624 624 Drywell struct r e: Design pressure, psig 30 30 25 30 External design pressure, 21 21 20 17-psid Air volume, ft8 270,000- 277,685 236,196 246,500 Suppression pool volume (includes vents), ft: 1.3E4 1.12E4 1.3E4 1.1E4 Suppression pool surface area, fts 553 482 522 455 Holdup volume, fta 50,000- 40,564 20,353 33,804 Holdup surface area, ft: 3,145 2,617 2,564 2,490 Containment vessel: Design pressure, psig 15 15 15 15 Ult *icate pressure capacity, psig 56 50 53 63 External design pressure, psid 3 0.8 0.6 3 Total air volume, ft8 1.4E6 1.141E6 1.192E6 1.551E6 Air volume below hydraulic control unit floor, ft8 151,644 . 181,626 153,792 173,000 Suppression pool volume, ft8 1.24E5 1.06ES 1.28E5 1,35E5 Suppression pool surface area ft2 6,667 5,900 6,408 7,175 Upper p,ool makeup volume, 5 ft8 36,380 32,830 0 14,655 Containment spray flow rate (1 train), gpm 5,650 5,250 0 3,800 Number of loss-of-coolant-accident vents 135 120 129 102 1 l Hark III SER 16
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_ _ ~ __. _ .. _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ i i The transients resulting in an 50RV were selected (1) to ensure a rapid foss of inventory and (2) O account for and create a limiting local thermal containment . environment for analysis and testing. Small-break loss-of-coolant accidents i (SBLOCAs) were selected as an alternate sequence to address the potential and consequences for hydrogen combustion in the drywell. Otherwise the SBLOCA sequence is identical to the 50RV sequence. For the base-case scenario, all ac-powered reactor makeu systems are assumed l to initially fail. According to emergency procedures, t e operator will depres-surize the reactor when water level decreases to the top of the active fuel or when conditions requiring steam cooling are met. Following vessel depressuriza-tion, low pressure system injection is assumed to fail. The scenario continues with the core becoming uncovered and core heatup beginning at about 35 minutes.
- into the transient. Limited hydrogen is produced during core heatup. At about !
65 minutes into the event, the core is reflooded before it becomes nonrecover- 1 l able (exceeding a 50% zirconium [Zr) melt fraction). During reflooding of the ! core a significant amount of hydrogen is generated. This hydrogen is transported ! to the suppression pool through the safety relief valve spargers and into con- ; tainment where it is ignited and burned, l The selection of the 50RV sequence was based on the reactor safety study methodo1-ogy applications program (RSSMAP) stud In 1986, the staff questioned the , ' absence of the station blackout (580) y. sequence (letter dated February 21.1986). l HCOG relativelyheld the melt low core viewfrequency that 580(HGN-055 is not a .likely)HGE This conclusion contributor is based on based the on its results of the RSSMAP study (NUREG/CR-1659), which assumed Grand Gulf to be representative of the four plants with Mark III containments. The results of the GESSAR-II PRA (NUREG-0979) also found that SB0 is a dominant contributor - to thelow. quite probability of core damage, although the core damage probability is The PRA results were reinforced by the staff findings reported in i an NRC report (NUREG-1150). In view of these studies, the HCOG revised its submittal to account for 580. These revised results are contained in two reports l transmitted by letters dated January 8 and September 9,1987. The staff's review focused previous on those revised results and also drew on information from submittals. Additionally, the review effort focused on the hydroge production profiles that were derived using the BWR core heatup code (BWRCHUC) described in Science Application Inc. and International Technical Services ITS) reports to the staff; HGN-020. -031, -032. -034. -089. -096, and -132; and HCOG/NRC meeting August 28, 1984). The objective of this review was to ascertain the capabilities and the . acceptability of the BWRCHUC for use in generating the hydrogen generation pro-files. Particular BWRCHUC concerns were (1) the tircaloy oxidation model, (2) the transient simulation capabilities, (3) the ability to predict the maximum expected hydrogen production rate, and (4) the ability to predict the total amount of hydrogen produced in each transient. 5.2 Evaluation - The evaluation was divided into two parts: (1) the establishment of an accept- ' able HGE scenario production and (2) the acceptability of the BWRCHUC to estimate hydrogen histories. Mark III SER 18 ____.___-___________._______-.m, ______,,..,,,m..,
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- ~ - 5.2.1 Acceptable HGE Sequence ! The analysis required by 50.44 is to be based upon an accident sequence which l is " acceptable to the staff" and is at the same time limited to
- recoverable *
- events. The rule, however, does not ;
I
- acceptability" or
- recoverability." provide criteria for the determination of .
The staff position with regard to recoverability is that there should be a reasonable expectation that the original core geometry is generally maintained. 1 However, a quantitative _ definition of a degraded core state that is recoverable is not required. The degraded core condition is a condition in which the 1 4 reactor core has experienced or is at the onset of experiencing damage from I excessivetemperature(includingpermanentdeformationorlocalizedmelting). Inherent.ly in the degraded core condition is an extended loss of coolant injection without a chance of immediate recovery. The purpose here is not j to associate core recoverability with detailed phenomena of cladding or fuel meiting and relocation, but rather to provide a reasonable cut-off as far as the deterministic calculation of hydrogen production is concerned. The total amount of hydrogen production which must be considered is specified in the rule itself. It is in tiis limited sense that the term " recoverable" is used in ] this evaluation. j i HCOG proposed a definition of recovuability in terms of the fraction of Zircaloy cladding which has reached or exceeded the Zircaloy melting tempera-ture of 2170 degrees K. The staff accepted a 50% Zircaloy clad melt fraction as the cut-off point for " recoverability" based on HCOG's report that analyses indicate that at this point significant fuel melting is in progress. , It is the staff's judgement that the maintenance of the original core geometry I after damage to this extent is unlikely. Therefore, for the purposes of hydrogen rule considerations and hydrogen generation rate estimates, the 50%
- Zircaloy melt fraction criterion is acceptable.
With regard to the " acceptability" of sequences, the staff considered two cri-teria: (1)thelikelihoodofagivensequenceand(2)thecontributionto risk from a given sequence. Based upon NUREG-1150, the staff concluded that , (1) the most likely HGEs would occur with the reactor vessel depressurized, (2)thepotentialforgreaterconsequencesisassociatedwithHGEsathigh pressure, and (3) the risk from all HGEs is estimated to be extremely low. In assessing which sequences should be considered by HC0G, the staff also con-sidered the uncertainties associated with low- and high-pressure events. For low-pressure events the requirements of the rule force conditions which are physically unrealistic (e.g. that the core be recoverable yet 75% of the Zircaloy is oxidized). This results in sequences which are somewhat artificial and therefore considerably uncertain. For high-pressure events these uncer-tainties are further complicated by a further lack of experimental data. The staff therefore judged that it is sufficient to consider only low-pressure sequences because (1) the overall risk from HGEs is believed to be low 2)from a risk perspective the reduced likelihood of a high-pressure HGE is like(ly to offsetthepotentiallyhigherconsequencesofsuchanevent,and(3)the additional uncertainties associtted with high-pressure event progression. 5.2.1.1 The HCOG's Base-Case Scenario The base case scenario proposed by HCOG results from a transient caused.by loss Mark III SER 19
- g. .
l I 4 of offsite power, subsequent reactor scram. MSIY rnd containment isolation, and l ene 50RV. All ac-powered reactor makeup systems are assumed to fail initially. ! However, the reactor core isolation cooling (RCIC) system which is de powerad and/or the fire truck diesel supply are available at Grand Galf. Emergency operating procedures require depressurization when the water level reaches the l top of the active fuel region and some low pressure injection system is avail- l able. It is assumed that following depressurization, the low-pressure systems ! fail to injee The core becomes uncovered and core temperature begins to rise l at about 35 minutes into the transient. As the core temper +ture continues to ! rise some hydrogen is produced. At about 65 minutes into the transient, the core is assumed to be ref* 7ded at a high flow rate. During the reflooding of the core, large amounts et hydrogen are produced and transported to the con-tainment through the safety relief valves. At this point in time, the core has } reached the recoverability criterion (i.e., the Zircaloy melt fraction is at I about 50%). A variation to the 50RV sequence is the SBLOCA scenario resulting from a hypo-thetical drywell break.. The drywell break is essentially the same as the 50RV sequence except that the hydrogen is discharged into the drywell as well as throughthesafetyreliefvalves(SRVs). The staff considered this conservative 1 i sequence in order to evaluate the effect of hydrogen burning in the drywell l where essential control equipment cabling is found.
)
5.2.1.2 Station Blackout and NUREG-1150 l The results of the GESSAR-Il PRA (NUREG-0979) also found that $80 is a dominant contributor to the probability of core damage, although the core damage probability is quite low. These results were reinforced by Grand Gulf findings documented in NUREG-1150. Subsequently, the HCOG submitted infonnation to account for SB0 to the staff by letters dated January 8 and September 9,1987. The results of analyses of Grand Gulf documented in NUREG-1150 indicate that the most probable HGEs result from 580. The most likely of these sequences (designated as TBU sequences) consists of loss of offsite power followed by the ' failure of onsite ac power in divisions 1 and 2 the failure of high-pressure core spray (HPCS) and reactor core isolation cooling (RCIC), and depressuriza.- ( tion of the reactor vessel. TBU represents more than 90% of all HGE sequences i and more than g3% of all low-pressure HGE sequences. The develo> ment of the loss-of-offsite-power (depressurized vessel) sequence begins by > oiling off the entire reactor vessel coolant inventory. With the core dry but at pressure, the operator depressurizes the reactor to increase the length of time available to support core recovery before the initiation of core damage. Following vessel depressurization the core begins to heatup causing oxidation of incore Zircaloy and core damage. Before core damage progresses to a point where a nonrecover-able core geometry could develop, a reactor vessel reflood system is assumed to be recovered. This reflood system then covers the fuel region with water termi-nating the event with a degraded, but recoverable, core geometry. _ 5.2.1.3 TBU the " Acceptable Sequence" The HCOG had considered the applicability of the various significant sequences identified in the draft version of NUREG-1150 to the HCOG program. The scenarios were divided into three categories; the short-tenn (about I hr) damage states TBU,TBUX,TCUX,theintermediate(4-6 hrs),andthelong-term (8-10 hrs)se-quences TB, TBUI, and TQUX (HGN-123). Differences, however became apparent l Mark III SER 20 l
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i whan the results of the revised NUREG were reviewed. The revised version of I NUREG-1150 estimates TBU to be the dominant HGE sequence which accounts for 93% of core damage frequency. ThephenomenologyoftheYBUsequenceis similar to tnat of the HCOG base case with regard to the 50RV. However, the HCOG experimental testing and analyses, which encompasses the TBU sequence, assumed the ignitors were continuously powered i of the transient when oc power was not available.ncluding during the portion In addition, the rule requires that the containment structural integrity and a safe shutdown be established and maintained. The ability to satisfy these requirements depend on both the total amount and the rate of hydrogen production. To estimate the maximum hydrogen production rate a d the total amount of hydro-gen produced, the rate of water. supply in the recomy phase of the HGE is critical. For purposes of the hydrogen control rule, the TBU sequence as described in NUREG-1150 (which encompasses the $0RV as described by the HC0G) is an acceptable sequence leading up to core recovery. In summary, the TBU is ; acceptable for the time sequence of events and for the hydrogen production rate , and total amount. J i (In probabilistic and TQUX denoterisktheassessment following: TB notation, theBlackout.
- Station tems TB TBU, fBU -Loss of offsiteTSUX, TBUI, TCU power (LOSP) with failure of all high pressure functions. The SRVs are opera- l tional and the vessel is depressurized. TBUX - LOSP with loss of all AC
- divisions and failure to depressurize. TBU1 - LOSP with failure of AC divisions 1 and 2 and of the high pressure core spray (HPCS). The reactor core isolation cooling (RCIC). operates for 6-8 hours before failing due to high pressure in the tubine exhaust TCUX - ATWS with LOSP loss of AC and HPCS and RCIC fail-ures. TQUX - Failure of all ECCS functions except power.)
5.2.1.4 Hydrogen Generation Profiles Recovery of cooling water flow is effectively bounded between 150 gpm from a single control rod drive cooling pump to 5000 gpm from the emergency core cooling ' system (ECCS) low-pressure high-flow-rate core recovery system. The hydrogen generation profiles for these extremes are qualitatively and quantitatively different. The probability of a high-flow-rate recovery is expected to be higher than that of a low-flow-rate system, because there are more high-flow systems (or combinations of systems) to inject water into a depressurized vessel; hence, it is reasonable to assume that the operator will attempt more often to recover one of the high-flow-rate systems. A high-flow reflood rate is associated with a high, narrow spike of hydrogen release, while the low-flow reflood rate will yield lower hydrogen production rates but for longer times (see Figures .1.1 and 5.2)(HGN-132). These profiles have been estimated by the HCOG using the BWRCHUC code, which is discussed in Section 5.2.2. The total mechanistic estimated amount of hydrogen released in the low-rate reflood case is higher than 1. hat released in the high-rate reflood case. The hydrogen peak release of th high-of the high-reflood case is about 35 seconds win at half maximum, while for the low-reflood case significant hydrogen release lasts about 8.5 minutes. ~ Table 5.1 shows a summary of the main features of both cases. l 1 l l Mark III SER 21
ji' Table 5.1 Hydrogen Release Profiles l i.ength of Total Peak Width at Reflood - Transient Hydrogen Rate Half Max ; Rate (gpe) (min.) (1bs) '(1bs/sec). (sec) 150 80.0 903.4 0.95 510 5,000: 25.8 604.8. 8.00 35 . 5.2.1.5 Non-Hechanistic Hydrogen Release Profile 4 The hydrogen rule requires consideration of metal-water reaction (WR) for 75% I of the Zircaloy cladding surrounding the active fuel region. -However the esti- y mated amount of metal-to-water reaction in either reflood rate scenario is- far !
; less than the required 75%. For Grand Gulf,;the active core region cladding is~ i 79,100 lbs. For the oxidation to proceed as: Zr + 2He0+2r0s + 2He, the amounts of Zr that correspond to the hydrogen released in Figures 5.1 and 5.2 are 20,450 lbs and 13 700 lbs, which represent about 26.0% WR and 17.0% WR rem
.t i spectively." (ThIs WR incorporates channel box and stainless steel oxidation.)
- A mechanism was needed to increase the release profiles to 75% WR of the active ]
1 i core region cladding, as required by the rule. The 75% WR of the Zr in Grand ) j Gulf is 59,300 lbs, which when oxidized will create'about 2,600 lbs of hydrogen. l It must be pointed out that mechanistic models that account for 75% WR of clad-l ding-oxidation result in a severely damaged core exceeding the recoverability j criterion. There are many i 75% WR cladding oxidation;possible however, noscenarios attempt isthat can made be hypothesized to estimate the phenom- to yield ena associated with such an oxidation level because it would require an . unreasonable recovery criterion. , As discussed previously, after core quenching in the low-rate reflood case, the calculated maximum amount of metal-water reaction is limited to about 26%. To meet the rule requirement of 75% WR, the HCOG submitted a nonsechanistic model used to predict hydrogen production based on an energy balance in a severely damaged core. It assumed that such a core has energy losses at least adequate to remove decay energy in the core, the energy produced by continued oxidation : of Zircaloy, and excess stored energy in the core. It also assumed that terai-nation of oxidation at 75% WR takes place by quenching of the core and removal of all excess energy (HGN-034). Considering the above, the oxidation rate will support a constant hydrogen release of about 0.10 lbs/sec. The staff finds i that this release rate is acceptable for hydrogen release to 75% WR, as required by the rule. . Therefore, for the scenarios shown in Figures 5.1 and 5.2 the " tails" correspond to 1700 lbs and 2000 lbs of hydrogen, i.e., an extension of about 17,000 seconds ; (4.7 hours) and 20,000 seconds (5.6 hours), respectively. The staff concludes that (1) mechanistic models can not predict the requifed 1 75% WR of cladding oxidation in the active fuel region without core damage
*Zircaloy is assumed to consist of 100% Zr. The actual composition of Zircaloy-2 in weight percent includes, Sn: 1.2-1.7, Fe: 0.02-0.07, and Ni: 0.05-0.15 '
l Mark III SER 22 l
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- 1 a beyond the recoverability criterion and (2) the use of a non-mechanistic release '
model based on heat balance is h asonable and acceptable. This leads to an oxi-dation rate producing 0.10 lbs of hydrogen per second, requiring an extension of ' about 4.7 and 5.6 hours for the scenarios of Figures 5.1 and 5.2, respectively. 1 5.2.2 BWR Core Heatup Code (BWRCHUC) 5.2.2.1 Introduction i The BWRCHUC has been used by the HCOG to calculate hydrogen rate release profiles for the hydrogen generation sequences described previously. The areas the staff ! considered in the review of the BWRCHUC is discussed below. The BWRCHUC was not validated or benchmarked to global core experimental data, rather i.t relies on collective engineering judgment and understanding of the phenomena taking place in a core disruptive accident. The lack of benchmarking or This lack of ben, validationprevents chmarking is due to the the results absence of of thesuitable experimental code from being used data. directly without 3 appropriate consideration of selected input parameters. The results of the BWRCHUC should be seen as an engineering estimate of the anticipated phenomena. Accordingly, t~e n code review was aimed at the reasonableness of the modeling, 3 the physical significance of the assumptions, and possible conservatisms in the ; l estimate. Reasonableness was assessed in terms of models and hy>otheses that have been advanced by other researchers in this field and any otier evidence !
- that could be gleaned from whatever limited and partial experimental informa-tion was available. For code modeling, the TBU sequence for an HGE was consid-ered equivalent to the 50RV sequence (paragraph 5.2.1.4) with respect to the t1us similar as far as hydrogen depressurization generation and This is concerned. core uncovery time,is the simplest and most sequence straight-forward, thus having the highest probability of being modeled correctly.
The BWRCHUC is a well-written computer code in that (1) it faithfully represents i the BWR geometric core design and (2) the models included in the code are ade-l quate to cover the specific HGEs selected for analysis by the HCOG and repre-1 sented by the TBU sequence. Modular architecture has been used extensively i where each module (subroutine) in the code treats a different phenomenon or,as- i pect of the problem. The code is built by connecting the various modules with exer dve routines. The numerical solution technique applied in the BWRCHUC is l apparently as good as any employed in severe accident codes. Numerical stability, l as reported by HCOG, is evidenced by the graphs of code output and the fact
- that reflooding calculations can be run.
l l The BWR core geometry is very complex. Some subtleties,of the i the potential to affed the prediction of hydrogen generation. geometry Therefore,haveit is ap3ropriate that a best-estimate code contain a re try titt is as complete as is reasonably achievable. presentation'of This has beenthe geome-done in the BWRCHUC. Considerable attention has been drawn to the fact that the BWRCHUC allows for a different two phase water level to be predicted in each fuel assen-bly represented. A separate level for the core bypass level varies according to the bundle power since the water in an assembly is assumed to be saturated at i the system pressure. Water levels may also vary because the void fraction of Mark III SER 25 4
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l . l the water in an assembly is a function of assembly power. The bypass level cal-culation further assumes that water in the bypass is subcooled and thus corres-ponds to the collapsed water level in the core. There is a hydraulic connec-tion between the assemblies and the bypass so that the water level in the bypass is reduced as the core water inventory is boiled away by the decay heat generated within the assemblies. This representstion closely corresponds to a partially covered BWR/6 core at low pressure before any structures in the core reach ten-peratures significantly greater than saturation. l 5.2.2.2 Phenomenological Assumptions A model for channel blockage was included in the code, but has not been employed i in the calculations since experimental results did not support total flow blockage. The blockage model assned that the fuel rod cladding melts while the channel box remains intact, holten cladding is then assumed to slump and refreeze within the channel feming a complete blockage, which prevents steam from reaching the Zircaloy surfaces within the assembly. In addition, steam i generation below the blockage pressurizes that portion of the assembly forcing the two-phase level in that assembly below the core plate. Since no steam enters the channel, all oxidation would stop. Experimental results from the PBF tests (HCOG presentation to NRC January 14,1985) indicated that a reduction in the flow area as a result of Zircaloy slumping did occur, but that complete blockage did not form. Without the channel blockage model hydrogen production is maximized all other conditions being the same. The lack of clad motion or channel blockage is a very conservative assumption with regard to hydrogen production. ' It is assumed that the control rods will remain intact since it is consistent with the recovery criterion. Under certain conditions, experimental evidence (R. O. Gaunt) suggests that BWR control rod blades could melt early in the core i i heatup phase of a transient. This would lead to the possibility of local loss of control; thus, when the core is reflooded, local criticality could result in intense heat productior, and core damage beyond the limits of recoverability. Therefore, control rod melt would be beyond the scope of this program. 5.2.2.3 Steam Production The modeling of steam generation within the reactor pressure vessel (RPV) can l have a significant impact on the quantity of hydrogen generated. While some aspects of steam generation are accurately modeled in the BWRCHUC, other sources of steam are not modeled at all. The steam generation modeling generally is incomplete; however, for the most likely HGE considered, the steam sources not represented do not significantly impact production of hydrogen. Within the BWRCHUC, the following five sources of steam generation are modeled: deposition of the decay power from that portion of the fuel assembly below the' two-phase level into the saturated water within an assembly heat transfer (by nucleate boiling) from portions of fuel rods, channels, control blades, and the core shroud that are at temperatures greater than saturation when they are covered by the two phase level l Mark III SER 26 i
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i L - radiative heat transfer from portions of control blades that have tempera-L tures higher than saturation to surrounding channel walls when the two phase level within the channel is at or above the portion (at elevated tempera-tures) of the control blade flashing of water in the downconer and lower plenum as a result of reduc-tions in the RPV pressure (Pressure-time history is provided by user input.) ' evaporation of core spray droplets entering the top of- a fuel assembly during reflooding of the core Steam generation resulting from flashing of the water inventory within the fuel- i assemblies and in the bypass region is not modeled. If systen pressure would decrease, flashing would occur. However the selected HGE sequence does not involve changes in pressure vessel pressu,re after hydrogen generation has begun. It is assumed that the RPV pres >ure is constant for at.least 10 minutes, which is the time required to rntes the bulk of the heat in the lower plenum struc-tures. Therefore, the lat M a flashing model is not a factor. Downward relocation of molten Zircaloy can have a large effect on steam genera- ; tion. If the two phase levt1 is above the core support plate, molten Zircaloy ! can run into water. Quenching or relocating Zircaloy in water would enhance steam and hydrogen generation. This phenomenon is not modeled in the BWRCHUC. i However, there is no water above the core support plate when Zircaloy melting- : occurs. Modeling of the melt relocation into the water would not increase the ' quantity of hydrogen produced compared to that which will be produced in the core reflood because of the more favorable surface-to-volume ratio. An oxidation cutoff temperature is used in the BWRCHUC as a surrogate for the effect of cladding and channel box relocation and subsequent quenching thereby removing the Zircaloy from the oxidizing environment (R.0. Gaunt and HC0G pre-sentation to NRC January 14,1985). The HCOG estimated Zircaloy oxidation vs. Zircaloy temperature and concluded that 2400'K is a conservative representation to account for this effect (HGN-032, item 4). Based on the evaluation performed by ITS, the staff has accepted the 2400'K as the irreversible oxidation cut-off temperature (letter to HCOG June 4, 1985). Ir. the reflond stage, quenching of Zircaloy that is at temperatures high'r e than the saturation temperature is nonmechanistically estimated. This can lead to overprediction of the steam generation rate during the reflood phase. For nodes ; i that are more than 100'K above saturation, quenching is assumed to take place ; in a single time step thus accelerating the heat transfer process and steaa production. , i Steam flow in the bypass reghn is underpredicted. However, the effect~of this underprediction of the bypass steam flow rate on the overall prediction of hydrogen' release is small. Overall steam generation rates in an overheated core could be underpredicted for transients in which the two phase. level is above the core support plate. In the staff's judgment, the extent of this underprediction is small compared with the uncertainties associahd with predictions of this nature. , Mark III SER 27 I
In summary, steam generation before reflood is' reasonably well predicted provided the RPV pressure has been constant for approximately 10 minutes. In the sequence considered, the RPV is depressurized and steam and hydrogen production take . place under conditions of constant pressure. 5.2.2.4 Hydrogen Generation As with the modeling of steam generation, the approach to modeling hydrogen generation is reasonable considering the difficulty of representing the phenom- 1 i ena to modelling techniques. The lack of models for a few relevant phenomena combined with some of the assumptions made fur phenomena that are modeled, leads to some uncertainty with regard to the predictions of the hydrogen generation rate during the dominant HGE. This uncertainty is expected to be negligible I (i.e., possess compensating effec'.s) in the present context. However, consider-ing the conservative assumption of no clad motion, we concluded that the overall ' hydrogen generation estimate is conservative. The considerations /pheromena that are related to hydrogen generation and are - not modeled or are underpredicted are listed below. . Oxidation below the location at which melting occurs is not modeled. I Because of the underprediction of steam in the bypass channel, oxidation of stainless steel and the outside of the channel is probably underpredicted. 1 Ballooning of the cladding and localized failure resulting in simultaneous interior and exterior oxidation is not modeled, thus limited hydrogen nderprediction may result. 1 Film boiling in a quenching mode is not modeled. This leads to higher { rate of hydrogen production for shorter time periods. It is not clear that an overall underprediction will. result. In the reflooding stage, vaporization of droplets that enter the top of fuel assemblies by radiant heat transfer does not remove heat from fuel i rods. This results in a conservative hydrogen production if the maximum temperature is below the cut-off and possibly not conservative if it is above the cut-off. It is not clear if the overall effect is nonconservative. Reaction rates of Zircaloy and stainless steel with steam are calculated using I the Arrhenius relationship. The reaction rate constants used in these expres-sions were derived by others from experimental results. This modeling of reac-tion rates and the associated heat generation is appropriate and consistent with what is used'in other severe accident modeling codes. A hydrogen blanketing fac- i tor is included in the formulation of the Arrhenius reaction rate expression. Hydrogen blanketing refers to the possible limitation of the oxidation rate from the diffusion rate of steam through the hydrogen emitted from the oxidizing sur-face. While the process represented by the hydrogen blanketing factor is real, a reduction of the extdation rate is almost certainly not realized under the conditions expected durir.g core damage in BWRs. Diffusion of steam through the oxide layer is the rate-1 miting process. Therefore, the hydrogen blanketing effect was not considered in the HCOG calculations, which represents a slight conservatism. , Mark III SER 28
l
- Because the oxidation rate varies exponentially with temperature, the represen-tation of. intact Zircaloy nodes reaching temperatures significantly higher
- than the melting temperature leads to higher oxidation than would be predicted ,
if melting were explicitly treated. Therefore, this is a conservative assump-tion. However, because the oxide layer is generally thick at these times, the actual quantity of additional oxidation is considered to be small. 'One could view this enhanced oxidation es a r.onr4chanistic approach to representing the initial enhancement in oxidation that probably accompanies slumping molten 2ircaloy. Heating of the cladding reduces the tensile strength and increases ductility. Simultaneous heating of the fuel and gases within the cladding leads to pres-surization of the rod from within. Ballooning of the cladding and localized failure may occur before melting. Failure of the 31 adding would allow the in-terior surface to be exposed to steam. It is therefore entirely possible that both the interior and exterior surfaces of the cladding will undergo oxidation. Since this possibility is not modeled in the BWRCHUC, hydrogen generation rates and total hydrogen generation could be underpredicted. However, in the staff's Judgment, the conservatism in the assumption that there is no clad slumping will adequately compensate for this potential underprediction. Overall, it is the staff's judgment that the modeling of the hydrogen genera-tion rate in BWRCHUC is reasonable and the total and peak hydrogen production estimates are expected to be conservative. 5.3 Sumary and Conclusions 1 1 There was a twofold objective in this portion of the evaluation corresponding to two requirements of 10 CFR 50.44: (1) "Use accident scenarios that are acctpted by the NRC staff." Paragraph (c)(3)(vi)(B)(3). , (2) " Provide an evaluation of the consequences of large amounts of hydrogen generated after the start of an accident...up to and including 75% of the fuelcladding(1). (c)(3)(vi)(B) surrounding the active fuel region...."Paragraph The first requirement corresponds to the dominant accident sequence that leads to an HGE. The information for such a sequence was derived from HCOG submit-tals and conforms to the revised (final) version of NUREG-1150. The TBU se-quence was found to represent 93% of all HGE sequences and consists of loss of l offsite power followed by failure of onsite ac power in divisions 1 and 2 and o failure of the HPCS and RCIC. The TBU sequence was found to be similar to the L SORV, which was initially proposed by HCOG. Thus, for hydrogen generation l purposes the TBU sequence satisfies the requirements of.10 CFR 50.44. The second requirement is limited to the acceptability of the BWRCHUC to estimate hydrogen generation profiles. The essential characteristics of such profiles are the peak rate, its duration, and the total amount of hydrogen produced. Mark 111 SER 29
j The staff reviewed the 8WRCHUC code on the basis of these requirements and ! information submitted by the HC0G. The staff finds the BWRCHUC code acceptable for use in calculating hydrogen production profiles. Therefore, the staff .; finds that the profiles estimated by HC0G using the BWRCHUC are acceptable for use in demonstrating compliance with 10 CFR 50.44. 1 4 6 CONTAINMENT RESPONSE - ANALYTICAL MODELING i In view of the quarter-scale test program, the emphasis on analytical methods ! for predicting containment response to hydrogen burning has significantly di-minished. This conclusion is based on the broad range of hydrogen release rates in which diffusive combustion is expected to occur. The staff believes ) 1 the evaluation of the survivability of essential equipment should be based on the i QSTF data. Therefore, there is limited value in pursuing such analytical meth-ods as the CLASIX-3 code; thus, the following evaluation addresses HC0G's of-fort to resolve the CLASIX-3 code analysis generically. As such, this effort is only relevant at low hydrogen flow rates that are near the flame extinguish-ment limit. As documented by various staff evaluations performed before the completion of the quarter-setle test program, the CLASIX code has been the principal analyti- l cal tool in predicting the containment response as a consequence of burning l hydrogen for plants with an ice condenser or Mark III containment. The CLASIX ' code or the CLASIX-3 code (which is the latest modified version that includes Mark III containment features) deals with deflagration (discrete-type) hydrogen , burning. The code is a multivolume containment code that is used to calculate i the containment ph sure and temperature response in separate compartments. Moreover, the code has the capability to model characteristics that are unique to Mark III containments while tracking the distribution of the atmospheric constituents (i.e., oxygen, nitrogen, hydrogen, steam). The staff's desire to demonstrate verifica' tion / validation of the CLASIX code has been an extensive effort. CLASIX results have compared well with results of other NRC-accepted analytical codes and hydrogen burning experiments. Fur-thermore, the HCOG has performed additional code validation by conoaring the " more recent Nevada Test Site large-scale hydrogen experiments to C.ASIX-3 code predictions. This is documentet. in HCOG's letter (HGN-113) dated January 8, 1987. With regard to hydrogen burning, the focus on code validation has been on pressure predictions because temperature comparisons are more difficult to predict due to their time and spatially dependent fluctuations. As discussed earlier, the major element of the HC0G's program is the quarter-scale test program. The data obtained from tests were used to perform equip-ment survivability analysis (see Section 7). These tests revealed that diffusion flames on the suppression pool surface can exist at a hydrogen injec- ' tion rate as low as 0.02 lb/sec under certain background conditions. As such, it is expected for a significant portion of postulated degraded core hydrogen profiles that diffusion flames will be the dominant combustion mode. Since CLASIX-3 does not model diffusion flames, these results have a significant bearing on the extent to which the CLASIX-3 code can be relied upon to predict . containment temperature environments, which further emphasizes the importance of the test program. Mark III SER 30 L.,. .-_i_...__ ._,. -.- . . .-_-,. .- , , , . _ . - . ,,,-_...,..,____.,.__-._,_-_m
4 i 6.1 Localized Combustion /CLASIX-3 ! l As noted in Section 3.1, testing performed in the QSTF revealed that combustion ' which occurred below the flame extinguishment limit was not global deflagration-type combustion; but " localized combustion." Localized combustion is character-ized as weak flames or weak volume burning through a marginally combustible hydrogen-air-steam mixture. By letters dated June 10,1986(HGN-092-P),and l December 15, 1987 (HGN-111-P), the HCOG provided various analyses to demonstrate that the CLASIX-3 model provided a boundary calculation for the combustion occurring below the diffusion flame threshold. This model employed a combustion mechanism that produces a more severe global thermal environment than has been measured locally in the QSTF for localized combustion, l The staff requested Sandia to review this approach. In its HGN-092-P submit-tal, the HCOG compared CLASIX-3 predictions for a quarter-scale model to exper - ) imental results of the corresponding test. The CLAS1X-3 predictions of the ' wetwell volume showed that the temperature prc, file exceeded the volume-weighted average of experimental data. On the basis of this analysis, the HCOG conclud-ed that CLASIX-3 yields conservative predictions of thermal environments inside containment for very low hydrogen flow rates. The staff questioned the appli- ~ cability of using experimental data pertaining to a local phenomena to demon-strate the capability of a lumped volume code. SNL shared the staff's concerns and recommended that the local combustion phenomenon observed in the QSTF warranted further evaluation. In its HGN-111-P submittal, the HCOG provided a comprehensive assessment of localized combustion seen in several quarter-scale tests. In thase tests com-bustion activity, as evident b periods of low hydrogen flow. Testydata thermocouple responses, do not indicate is widespread durIng that concentrated flame energy deposition occurs at fixed locations. Energy deposition appears to be rapid and diffuse and is dominated by convective mixing, combustion-induced turbulence, plume influence, and background gas flows. Typically, peak temperatures recorded during localized combustion are relatively low and persist for short durations. The tem to relatively low background levels. perature responses are cyclic and return The HCOG analyzed five quarter-scale tests conducted during the scoping test phase of the program in an effort to better understand localized combustion. < At the low hydrogen flow rates, these tests demonstrated certain repeatable l trends and the thermocouple activity observed recurrent and generally predict- ! able. Some of the findings resulting from the evaluation of localized combus-tion are briefly described below. (1) In tests'without sprays, combustion was generally widespread. Whereas when sprays were activated, combustion appeared to be suppressed in ope,n chimneys (i.e. , annular quadrants) as a result of cooling effects and shifts in global flow patterns. Also, enhanced mixing resulting from I i sprays induced slightly higher temperatures in some areas, but not appre- l ciably higher than those recorded when sprays were off.
- (2) Comparing the scoping test results, it appeared that the location of the l SORVs did not have as significant an effect as other parameters, such as variation in hydrogen flow rates.
Mark III SER 31
(3) Pr'obably the most inr.ortant finding pertained to combustion activity in the vicinity of the hydrogen igniters. The closest themocouple located to a nearby igniter was about 15 inches laterally or about 6 inches later-ally and 15 inches above the igniter. Data indicated that toAperatures close to igniters were generally no more severe than recordi'ags several feet away. Also whe to open flow regions,na comparing the effects significant increase of blockageswas in temperature above not ignitors ob-served. Combustion energy dispersion was prevalent. To further support its findings of turbulent mixing induced by combustion, the HCOG included a discussion of a test in which pool flames were observed.- A thermocouple was placed about 1 foot directly above the pool surface over an active sparger. Readings indicated that'at low hydrogen flows the flames at this location appeared to be intermittent and unttable. The temperature re-sponse did not exceed 425'F as a result of these unsteady pool flames. The HCOG contends that,.because of the efficient mixing, one should expect local hydrogen concentrations elsewhere in the facility to be less than at the sup-pression pool surface. Moreover, this fact coupled with temperature readings (discussed above) and the absence of visual indications regarding flame forma-tions above the HCU floor, is not strongly supportive of a hypothesis that sus-tained high-temperature localized combustion zones will be established at very low hydrogen flow rates. In addition, HCOG indicates that the temperatures generated from pool burning at low hydrogen flow rates (i.e., about 0.15 lb/sec) from resultant hot than localized combustion. plumes represent a more severe thermal environment As part of this assessment, HCOG provided additional information with regard to the role of the CLASIX-3 code in its analyses and the conservatisms used for i containment modeling. While global or large volume deflagration as modeled by CLASIX-3,didnotoccurintheQSTF,theHCOGcontendsthatthedLASIX-3model-ling would conservatively bond the observed localized combustion environment. To assess the severity of the environment from an equipment survivability per-spective, the HCOG compared the thermal lo' ads created by the most severe local-ized combustion measurements at the QSTF to the corresponding CLASIX-3 temperature profile. The results of this comparison show the CLASIX-3 profile generates a significantly more severe environment than that produced by localized combustion. l The staff requested that SNL review this issue along with the consideration of scaling aspects. SNL determined that the HCOG adequately addressed the likely locations for localized combustion and identified reasonable bounds for the most threatening thermal environment for equipment lofated near regions of lo-calized combustion or in the resultant hot plumes. Moreover the thermal local comparisonincombinationwiththeHCOG'sdiscussionoflocalizedcombustion provide adequate justification that the CLASIX-3 thermal load would be more severe than that experienced in the QSTF for low hydrogen injection rates. In conjunction with discussions contained in the QSTF test re '$NL assertsthatthereisreasonableassurancethatthetherma$ response at ort fu (HGN-121-P)ll scale will be no more threatening than that experienced in the QSTE. The staff has also evaluated this issue and concurs with SNL's assessment. Based on the modeling methodology used in the referenced submittals (e.g., low hydrogen flow rates and the focus on the wetwell profile), the staff finds that the CLASIX-3 predictions would be acceptable in determining the containment environmental conditions as a consequence of localized hydrogen burning. Accordingly, these profiles could be used to evaluate the survivability of equipment. Mark III SER 32 i
i 6.2 Containment Pressure and Temperature Calculations Letters HGN-092-P and HGN-109-P documented the HC0G's calculations of the con-tainment pressure and temperature response based on postulated degraded core scenarios that are discussed in Section 5 using the CLASIX-3 code. To deter-mine the adequacy of the hydrogen ignition syste'. (HIS), the HC0G considered
- two types of accidents in its generic analysis
- a stuck-open relief valve (SORV) transient and a small-break loss-of-coolant accident (SBLOCA) in the drywell. Thecomponentofthehydrogenreleasehistorythatisofinterestin this analysis is referred to as the tail" portion and represents a nonmechanistically defined constant hydrogen production rate. As discussed above, the CLASIX-3 results bound the thermal environment that may be produced for low hydrogen release rates that are below the diffusion flame extinguish-ment limit.
HCOG provided a generic sensitivity study using the Perry Nuclear Power Plant containment characteristics for the CLASIX-3 model. In this sensitivity study, different parameters were varied to assess the effects on the calculated re-suits. The staff focused on the most important parameter considered, which was the assumed availability of the containment sprays. HCOG had chosen to use the CLASIX-3 code predictions without sprays (i.e., for the 30RV case) in its ge-neric survivability study (discussed in Section 7). For the equipment surviv-ability analysis to be generic, it became necessary to consider the no-spray case because fan coolers rather than sprays are part of the River Bend con-tainment design. For the 50RV case, all mass and energy releases were directed into the suppres-sion pool. The CLASIX-3 model used in the generic analysis simulated four com-partments of the Perry containment: the drywell volume, the wetwell volume (bounded by the HCU floor and the surface of the suppression pool), the inter-mediate volume (bounded by the HCU floor and the refueling floor), and the con-tainment volume (above the refueling floor). Figure 6.1 presents a schematic representation of the model. Ignition of hydrogen combustion was assumed to occur at a 6% hydrogen concentration with 65% combustion completeness. The CLASIX-3 50RV base-case model produce a transient in which the hydrogen was ignited in a series of burns in the wetwell volume. Fi l the computed wetwell temperature and pressure profiles.gures The early 6.2 and 6.3 of portion show the transient resulted in the highest wetwell temperature. This is symptomatic of the hydrogen spiked release in the early phase of the release profile. Dif-fusion flames would be prevalent in this interval and would be beyond the range of use for the CLASIX-3 methodology. For the major portion of the temperature j profile, the wetwell burns produce a peak wetwell temperature of above 800'F. At the end of the hydrogen release period, the calculated hydrogen concentration in the containment volume did not reach the ignition criterion of 6%. In the CLASIX-3 calculation, the HCOG assumed a containment burn to occur at this lower concentration which resulted in the most severe pressure excursion, to approxi-mately 23 psig.
- .6. 3 Drywell Analysis For the base-case analysis of a small pipe break in the drywell (DWB) the CLASIX-3 containment model was similar to the SORY case except the hydrogen / steam source Mark III SER
- s3
I terms are directed to both the drywell volume and the suppression pool. The DWB scenario was chosen because of the potential and consequences for hydrogen . combustion in the drywell. Of the cases studied, only the 2-inc.h DWB case had conditions where a hydrogen burn was predicted to occur. Ignition in the dry-well was limited by lack of oxygen (i.e., below 55) because air is forced from the drywell by vessel blowdown. The only burn predicted by CLASIX-3 for the 2-inch DWB case resulted in a peak drywell temperature of about 1050'F and a peak drywell pressure of about 13 psig. For the events considered, where high steem flows are directed into the drywell along with the diversion of most of the hydrogen to the suppression pool, the hydrogen threat to the drywell appears ) to be relatively small. As indicated in Section 7. HCOG performed thermal re-sponse analysis of selected drywell equipment. The results demonstrated that the equipment would survive the drywell burn. l 6.4 Existence of Drywell Diffusion Flames In the DWB case, air would be reintroduced in the drywell through vacuum breaker actuation or operation of the drywell mixing system. The drywell environment is predicted to be a hydrogen-rich /exygen-lean mixture. When oxygen is re. introduced in the presence of an ignition source, a diffusion flame may result in the vicinity of the oxygen source. This possible combustion phenomenon is referred to as an inverted diffusion flame. This is a concern since the poten-tial to establish a continuous inverted diffusion flame at the oxygen source may result in locally severe thermal loads.
)
By letters datec June 25,1986(HGN-091),andJune 10,1987(HGN-119),HCOGeval- l usted the potential impact of inverted diffusion flames in plants with a Mark 111 containment. In the HGN-091 submittal, the HCOG discussed the criteria for establishing the existence of inverted diffusion flames. The HCOG indicated that flames will not occur in the drywell when conditions are outside the flam-mability curve. In the HGN-119 submittal,.the HCOG further discusses the low likelihood of achieving the necessary combustible conditions in the drywell based on the CLASIX-3 predictions. SNL reviewed the initial submittal and determined that the HC0G did not provide sufficient justification to preclude drywell burning. Specifict11y, SNL com-mented that the flammability limit merely establishes the limit.; that will allow I flame propagation; burns that do not propagate into the mixture are not precluded by being outside the flammability limits. Furthermore, it was not obvious that , the burning mixture should De expected to follow the path predicted by the HCOG. Generally, there is a lack of experimental data to support the HCOG's position. However, recent risk studies do not support the DWB case as a dominant core-l melt / degraded-core event for plants with a Mark III containment; therefore, fur-ther phenomenological investigation may not be warranted. Drywell break events are further discussed in Section 5. In addition, the expected redundancy (i.e., l spatialseparationofequipmentperformingthesamefunction)ofthecritical equipment should compensate for possible locally severe thermal loads. The staff believes there is a reasonable level of assurance that the consequences of a drywell break event would not pose a significant threat to containment integrity and would not preclude safe shutdown of the plant. However, the staff believes, as part of IPE process that each licensee of a plant with a Mark III Mark Ill SER 34
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- mixing system to support the above conclusion 1for each plant.
h( 7 SURVIVABILITY OF-ESSENTIAL EQUIPMENT s '4 As part of the analysis,10 CFR 50.44(c)(3)(vi)(B)(5)(ii) states: , N i Systems and components necessary to establishtand maintainisafe shut - down and to maintain: containment integrity will be capable of perform - ] ing their functions during and after exposure to the environmental j conditions created by the-burning of hydrogen, including the effect of . local detonations, unless such detonations can be shown unlikely to occur.- 1 Accordingly, each licensee with a Mark'll! containment is' required to demonstrate that the essential equipment located inside the containment will survive the
' hydrogen burn environment. To support this objective, the HCOG conducted two-programs to define the environment that would result from hydrogen combustion.
As discussed earlier in this evaluation, the quarter-scale test data will be used to define the environment that would be produced by diffusive combustion: ; on the suppression pool- surface. In addition, the CLASIX-3 code analysis will ; be used to define a bounding environment for. localized combustion below the diffusion flame extinguishment limit. 7.1 IdentificationofEssentialEquipment i The equipment that has to survive hydrogen burning was selected on the basis of function during and after a postulated degraded core accident. Generally, all the equipment located in the containment that was considered to be in one of
-the five categories listed below was considered to be essential for the safe' shutdown of'the plant. ,
(1) -systems and components that mitigate the consequences of the accident i (2) systems and components needed for maintaining the integrity of the containment boundary (3) systems and-components needed for maintaining the core in a coolable geometry. (4) systems and prov,and components needed for monitoring the course of the accident iding guidence to the operator for initiating action in accordance< ' with emergency procedure guidelines (5) components whose failure could preclude the ability of the above systems to fu' fill their intended function '"~ Using these criteria, the HCOG identified the equipment that would be needed to be evaluated for survivability. In its letter (HGN-084) dated May 16, 1986, the
>HCOG transmitted to the staff the list identifying the following system / components:
Mark III SER 38
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l (1) containment vent ~ valves,and penetrations: vacuum breakers-air lock, hatch seals, electrical penetrations, ! i (2) drywell components: breakers air locks, hatch : seals, and post-accident vacuum (3) hydrogen igniter sysses (4) combustible gas control system: hydrogen recombiners system, and post-accident atmosphere sampling valves , drywell mixing (5) containment cooling: unit coolers sprayisolationvalves,LPCIinjectionvalves,and . (6) automatic depressurization. system 1 (7) containment and reactor monitoring: . h
- containment and drywell temperature instruments reactor pressure vessel wide-range pressure instruments reactor pressure vessel wide range and fuel-zone level instruments (8) junctionblocks. associated instruments, controls, cables, interlocks, termina equipment conditions. essential for the mitigation of postulate As part of the final analysis containment should provide plant-specific each licensee with a Mark corresponding Information III to t generic list the selection criteria. in conjunction with unique design features that are relevant to l
l 7.2 _ Generic Eauipment Survivability Analysis (Localized Combustion) As discussed in Section 6, by using the CLASIX-3 code. pressure and temperature predictions were obtained the containment environment or boundary conditions .nec ment response analyses for hydrogen release rates below the diffusion fl extinguishment limit where burning is limited to localized combustion. believes that equipment survivability can be established The genericall The HCOG su porting analyses was presented in its letter dated A . nature of used,in the combustion the generic phenomena modeled by the CLASIX-3 c conditions equipment survivability analysis. has listed some of the more important items below.The HCO l (1) Theconstanthydrogenreleaserateof0.1lb/sec,whichistheionmechan-isticrecovery. core " tail" portion of the release profile, is unlikely to occur after surface of the pool may exist as low as 0.02 lb/sec indic Mark III SER 39
4 . -4 combustion would not' occur. It is expected that the presence of diffusion flames would probably be the dominant combustion mode, pessib1 in combi-nation with localized combustion phenomena when the hydrogen f ow rate is- - p below the flame extinguishment limit. The significance of these two dif- ~ ferent combustion modes is the spatial shifting of themal-loads; as such 3 a single piece of equipment would not continually be exposed to hydrogen , .{ burning resulting in a lower temperature profile. ' 1 (2) The CLASIX-3 wetwell temperature profile was:used as the boundary condition ! for the equipment response analysis although the most sensitive equipment i is located outside of the.wetwell volume. The wetwell has the severest '
- environment of the three containment volumes. The staff finds that based on limitations of the CLASIX-3 methodology used in the generic analysis,- ~
s there is no choice but to use the wetwell volume. However,-the staff does i recognize-the selected profile is limiting i 1 (3) In the selected CLASIX-3 case there are no active containment cooling mech-anisms(i.e.,thelackofavaIlabilityofspraysorunitcoolers). Because .p of the type of event considered, a recoverable degraded core, the HC0G expects that sometime during these relatively long transient eventsi spray / - 1 unit coolers would become available. A set of equipment common to each plant with a Mark III containment was compiled - H from the list of generic equipment. Subsequently, the most thermally sensitive? equipment, such as cables pressure transmitter i ADS solenoid valve, were Included in-the generic, hydrogen igniter survivability assembly, analysis. Based and j on the results of the generic equipment survivability analysis, the drywell break l equipment response analysis showed favorable results; whereas, in the 50RV case ; the thermal analysis response for the pressure transmitter indicated a 27'F ex , coedance above its qualification temperature. The significance of this result is assessed below. , The equipment response analysis for the SORY case used the wetwell CLASIX-3 tem-perature profiles, presented in'Section.6, with some modifications. These mod-ified profiles exclude the few initial burns in which diffusion flames would exist and the last induced global burn. As a result, the modified wetwell pro-file contains about 90 serial hydrogen burns. The calculated critical component ', first of the pressure hydrogen transmitters burn. However, the HCOG exceeded itsthe indicated that qualification after the pressure transmitter seventy is ! expected to survive the hydrogen event because of various conservatisms in the analysis. )' The staff acknowledges conservatisms, as 44 cussed earlier, are contained in these analyses which could compensate for the temperature exceedance over the qualification of the pressure transmitter. Nonetheless, the staff requested the-HCOG to provide additional data on the qualification of the pressure transmitter. By letter dated April 5, 1988 (HGN-131-P), the HCOG indicated that during.quali-fication-testing the transmitter had operated without failure at Aurface temper-. atures approaching 380*F for several minutes (as compared to the qualification of 320'F). As part of tb HCOG's response, an equipment response analysis of i the pressure transmitter was re-evaluated assuming containment sprays to be available. This response analysis indicated that the equipment surface tempera-ture was about 70'F less than qualification temperature of 320'F. These results Mark III SER 40
' demonstrated the impact of sprays to cool the containment environment thus maintaining the function of essential equipment. RiverBendStationIsthe {
only plant without containment sprays, but unit coolers are part of its design. - While specific analyses have not been performed to support quantification of the cooling effeet provided by unit coolers versus sprays, the HCOG concludes ) that a reduction in background temperature would be adequate to reduce the thermal loads on the pressure transmitter. In summary, HCOG contends that further analysis is unwarranted because, with the potential for active containment cooling and the conservatisms inherent in the ! analyses, the prcssure transmitter will function as designed during recoverable degraded core events that progress to 75% metal-water reaction. With regard to i L thehe analyses at low hydrogen flowrates, the staff agrees with HCOG's position - that further effort in this area is not warranted. . Moreover, the. staff finds , that the determination of equipment survivability based on the data obtained a from the QSTF for diffusion flames is more appropriate than an assessment based on localized combustion conditions. 7.3 Diffusion Flame Thermal Environment Methodoloay In its letter dated July 30 1986 tobeusedbyeachmemberIIcensee(HGN-103),inethefull-scaleplant-specificthe to determ HCO containment thermal environments from the QSTF data. The full-scale environmen-tal conditions would be used as boundary conditions in the HEATING-6 computer code to analyze the response of containment equipment during postulated diffu-sive combustion events. As a result of these analyses, the survivability of essential equipment would be determined. From the production test series conducted for each Mark III containment, the testthatproducesthemostlimitingenvironmentatthecorresponding(tofull scale) equipment location is used. Thermal profiles are constructed by spatial mapping of the test facility data. Specific plant profiles are develope 0 from average temperatures for time intervals of maximum hydrogen flow and low con-stant hydrogen flow from the production tests. This allows determination of l the plume locations and the effects of blockages and spargers. Full-scale velocities are computed from the quarter-scale measured velocities using Froude scaling; test temperatures are used directly .The i convective heat transfer and radiative heat fluxes are comp uted using (scaling is 1:1). the scaled velocities and temperatures. Since this approach establishes an environmental map, the heat transfer modes that should be considered are dependent on the i location of the affected equipment. To validate a'nd assess the heat transfer methodology, a complex (three-dimensional ! geometry) calorimeter assembly was used in several quarter-scale tests to subject I the calorimeter to different locations and different thermal environments. A HEATING-6 model of the com)1ex calo!'imeter was constructed and the calculated response was compared to tie measured response to validate the methodology. This effort is presented in the HCOG's letter (HGN-105-P) dated August 29, 1986. SNL assessed the submittals and determined that most of the computed results were conservative from the standpoint of the equipment survivability. Therefore, the correlations and results presented are reasonable. However, SNL recommended Mark III SER 41 l
that when the generic methodology is used for plant-specific equipment evaluations, a review should be conducted to assure conservative specification of the boundary conditions. n In summary, the staff finds that the heat transfer methodology dealing with diffusive co6ustion as presented in HGN-IO3, provides an acceptable fcundation to perform p ant-specific equipment response analyses. Accordingly, each . licensee with a Mark III containment intends to use this methodology as part of its final analysis, required by the hydrogen rule. The staff agrees with SNL's. , recommendation, that sufficient detail of input data should be provided by. each licensee to ensure its analysis is conducted in an appropriate manner. 7.4 Syray Availability In tl 3reliminary evaluations of hydro Gulf ! ?R Ho. 3, n' UREG-0831 July 1982) the gen staffigniter systems allowed credit(e.g., see Grand for operation of conti. W sprays in the analyses of the consequences of hydrogen combus-tion during o% 'ded core accidents. The validity of the assumption of contain-ment spray operab111ty was premised on several considerations. First, in the preliminary evaluations of igniter systems the staff and HCOG focused en the SORV transient and drywell pipe break accident sequences. These accident sequences do not necessarily imply loss of the containment spray function of the RHR pumps, RHR pumps may be operable but the LPCI injection path may be interrupted or lost. Further, at the time of the preliminar overall tone of the BWR emergency procedure guidelines (EPG)y was teevaluations focus on the containment integrity rather than adequacy of core cooling at &n earlier point in a degraded core accident sequence. Since the preliminary evaluations were conducted, additional information has been developed which raises questions regarding the validity of assumptions concerning availability of the RHR pumps 1.n the containment spray mode. In contrast to the earlier focus on the 50RV transient and drywell pipe break recent riskevents. generation analysis indicates that SB0 is a significant contributor to hydrogen For the SBO, the loss of reactor. makeup is tied to the loss-of pumps, including RHR pumps, in either the LPCI or containment spray mode. Thus for the SB0 sequence, the RHR spray function cannot be reasonably assumed to be available until ac power is restored. Finally, the earlier emphasis in the EPG'shas operation on been containment. reversed.integrity vs core cooling for containment spray InRev.4totheBWREPG's,(March 1987)the sequence of steps has been modified. Use of RHR r, umps in the containment spray mode, irrespective of adequate core cooling, is r.ow directed as the last step, to control pressure rather than befcre the deci:, ion to vent. For the above reasons the staff concludes the BWR Mark III owners should ev the containment and essential equipment response to hydrogen generation events assuming and the EPGs. containment sprays are unavailable, consistent with SB0 assumptions I Spray operability can be modelled but should be trg,ated in the context of establishing margins for a variety of possible plant conditions. Similarly, assumptions regarding availability of containment coolers should be < consistent with the basic premise of the SB0 accident sequence. Mark III.SER 42
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. s.
7.~5 Pressure Effects In HGN-118-P, the.HCOG indicated equipment located inside containment is quali- ,, fled to a pressure loading.of at least 30 psig applied externally. The CLASIX-3 predictions produced the most severe pressure rise of about 23 psig in the' Mark
- III containment. The staff concludes that pressure is not a concern pending confirmation by each licensee of the.30 psig capability. When the hydrogen ig-nition system is functionir,g, various containment.subvolumes will be randomly -
affected by hydrogen burnirig, however, a large pressure spike is not expected to occur. 7.6 Detonations The HCOG believes that a detonation is not a" credible phenomenon in'the Mark , III containment because (1) no rich hydrogen concentrations will accumulate. t inside co9tainment since the distributed igniters will initiate combustion as the mixture reaches the lower flamability limit and effective mixing will occur and (2) there are no regions of the containment with sufficient geometrical con-finement to allow for the flame acceleration necessary to yield a transition to detonation. The staff agrees with the HCOG position. As confirmed by the quarter-scale test results, the atmospheric conditions inside the test facility was well mixed and burning at low hyarogen concentrations was prevalent. Thus, the potential I for localized accumulation of significant concentrations of hydrogen is concluded to be unlikely, j 8 CONCLUSION l On the basis of the above evaluation, the staff finds the HCOG topical report, t
" Generic Hydrogen Control Information for BWR-6 Mark III Containments,"
l (HGN-112-NP)datedFebruary 23, 1987, provides an acceptable basis for techni-cal resolution of the Mark III containment' degraded core hydrogen issue. Each licensee should rovide a plant-specific final analysis, as required by 10 CFR 50.44(c)(3)p(vii)(B), which will address the elements specified in 10CFR50.44(c)(3)(vi)(B). The HCOG topical report, or portions thereof, ray be referenced where appro)riate, taking-into consideration the staff l recommendation as stated in t11s' report. The plant-specific analysis will use l I test data described in the topical report to confirm that~ the equipment necessary to establish and maintain safe shutdown and to maintain containment integrity will be capable of performing their functions during and after exposure to the environmental conditions created by the nydrogen in all credible severe accident scenarios. l An elt nt the staff's assessment for determining the adequacy of the HIS was the : termination of whether or not an alternate power supply was appropriate. An im)ortant factor in this decision process is the level of risk associated witi SB0 events leading to core damage. Recent risk studies reported in NUREG-1150 have shown that the overall core melt frequWncy for one Mark III plant (Grand Gulf Nuclear Station) is very low, i.e.,1E-6/ year. However, a potential vulnerability for Mark III plants involves station blackout (SBO), during which the igniters would be inoperable; and this i Mark III SER 43
cendition appears' to dominate the residual risk from severe accident in the Mark III plants. Under $80 conditions, a_ detonable mixture of hydrogen could develop which could be ignited upon restoration of power resulting in loss of containment integrity. On the basis of a separate evaluation of this ! possibility in the context of the 'NRC staff Containment Perfomance - ' Improvement (CPI) program, the staff has recomended that the vulnerability ' to interruption of power to the hydrogen igniters be evaluated further on a plant-specific basis as part of the Individual Plant Examination (IPEs) of the ' i Mark III plants. The staff has requested that the licensees consider this 1 l issue as part of the IPE in. Generic Letter 88-20, Supplement 3. With the caveat that the vulnerability to interruption of power to the hydrogen igniters should.be further evaluated on a plant-specific basis as part of-the IPEs of the Mark III plants, the staff finds that there is reasonable assurance that~ the HIS installed in the plants with Mark III containments will act to control the burning of h adequate protection against containment failure. ydrogen so that there is , In summary, the staff concludes that the following key elements should be ad-dressed in each licensee's plant-specific final analysis to resolve the degraded core hydrogen control issue: ' (1) hydrogen ignition system design. (the vulnerability to interruption of - power to the hydrogen igniters should be further evaluated on a plant-specific basis as part of the Mark'III plants IPEs) (2) confimation of applicability to the generic effort (3) quarter-scale plant specific production testing results i (4) primary containment structural. survivability l quarter-scale testing pressure capacity analyses for drywell and containment, for example.- confirm previous plant-specific analyses (5) survivability of essential equipment ' i - identify plant-specific essential equipment - 1 define themal environment from quarter-scale testing perfom equipment response analysis
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confirm that redundancy exists for that equipment affected by second-ary burning and drywell inverted diffusion flames confirm pressure capability of equipment l l Mark III SER 44
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.N r APPENDIX A l GENERICMhDROGEN1GN]T]DN'SYSTEMTECHN]CALSPIC1FICAT10 Generally, technical specifications (TS) of a particular system consist of two-L distinct sections: Surveillance Requirements to. ensure system operability and ;
i Limiting Condition for Operation (LCO) to define the allowable operability- d range in conjunction with various plant actions when neededt Each of the fourl plants with Mark lll-containments have similar TS for the hydrogen ignition I
- system (HIS). The following discussion on this-subject'is focused on the ;
l proposed generic His TS and its deviations-from the current TS. -i Currently, TS on igniter systems in plants with Mark !!! containments prescribe
.two types of-surveillance practice.. At'184-day intervals, all the igniter i assemblies are energized and current / voltage measurements are performed and' !
compared with similar measurements taken previously. If more than_three igniter. ossemblies on either subsystem hre determined to be-inoperable, there is-an l increase of the surveitlance frequency to a 92-day -interval. A second'part of- i this first surveillance . requirement is the verification that inoperable igniters-i are not adjacent to each other, if more than one igniter:on each subsystem is determined to be inoperable. This requirement is based on the staff's view
)
regarcing potential hydrogen pocketing in enclosed areas. .The second type of . surveillance is conducted at 18-month intervals to verify a surface temperature - l of at least 1700'F for each accessible igniter and verify by measurement-sufficient current / voltage to develop 1700*F, surface temperature for'those igniter assemblies in inaccessible areas. Accordingly, the bases section of the TS indicates that inaccessible areas are defined as areas that have high-radiation levels during the entire refueling outagf; iuch enclosures' include the heat exchanger, filter domineralizer, and the pomp room for the reactor water cleanup (RWCU) system. The current LCO allows no more thar.10% of the igniter assemblies inoperable per subsystem. And if one subsystem is inoperable, the action statement ie-quires restoratier, to operable status, or be in the required operational condition, within 30 days (similar to the hydrogen recombiner 75). By letter dated April 16, 1986 (HGN-070), the HCOG proposed to revise selected portions of the existing plant specific TS as outlined above. Principally, there are two significant proposed changes: (1) an increase of the allowable inoperable . igniters per subsystem to about 40%, as compared to the current value of 10%, and (2) removal of the surveillance requirement in determining the location of the inoperable igniters after the requisite number of failed igniters have been attained. Also, as discussed above, the current action statement system to operable status. requires an allowable period of 30 days to restore the igniter sub-HCOG proposed to change this interval to 60 days because these events in which the HIS is required to be operable are less probable than design-basis accidents. The staff finds the' proposed-inoperable period increase is not based on sound engineering judgment since it relies Mark III SER 1 Appendix A
, s on being beyone the DBA but does not provide a rationale for 60 days. Thus the 3D-day interval is appropriate and should be maintained.
=1n its letter of April 16, 1986, the HCOG provided a justification for the other two significant proposed changes. Essentially, the HCOG cited the con-clusions observed regarding the QSTF testing (see Section 3.1) of a particular -
scoping test in which about 40% of an igniter subsystem was inoperable in con-junction with the other subsystem not functioning. The staff has concluded relative to the hydrogen aspects of the TS justifica-tions that the HCOG had not provided sufficient justifica*. ion to relax the TS to_such a degree. The staff made its determination beta se of the inherent i i uncertainties such as extrapolation of the quarter-scale results to full scale, various injection rates, different safety relief valves actuating, and differ-ent combinations as to where the 40% of the inoperable igniters could be located. Therefore, the allowable value for inoperable igniters should be as low as practical; a 10% value appears to be a reasonable limit. , The second proposed change is to remove the surveillance that ensures that inoperable igniters are not adjacent. Principally, this surveillance ensures at least one operable igniter in each enclosed area and coverage of the azimuthal positioneo igniters in the cpen regions. Essentially, the HIS TS are intended to prevent buildup of hydrogen in subvolumes of the Mark III contain-ment, and thereby prelude the occurrence of large volume burns. The following considerations are highlighted as part of the HCOG's justification for proposing are not adjacent: to remove the TS provision to determine that inoperable igniters The HCOG evaluated the potential flow paths that could transport hydrogen l in or near enclosed regions of the containment and determined that no potential hydrogen source exists. It is expected that igniters in open areas will function to preclude local hydrogen pocketing. t Observations of the quarter-scale tests indicated that the released ' hydrogen will tend to mix with the surrounding atmosphere and thus reduce the potential of locally high nydrogen concentrations. The likelihood is low for inope'able r ignitors being located in such a fashion as to create a large containment subvolume that would be without
- igniter coverage. Igniters would tend to fail in a random manner.
Currently, whenever at least one igniter is inoperable in each sub-l system, containment entry is normally required to find the location of the failed igniter. This would subject plant personnel to various l occupational safety hazards such as radiation ex associated with the construction of scaffolding.posures and tha. risks - On the basis of these considerations provided by HCOG, the staff has determined that there is reasonable assurance that the proposed TS without the adjacent igniter system. igniter provision would not have an adverse effect on the effectiveness of the Mark III SER 2 1 Appendix A
l 'r '. l~ t
. The staf f finds the generic HI~ TS as documented in th.) HCOG 1etter dated- .
April 16, 1956, to be acceptable contingent on the following changes: . the 40% l value of allowable inoperable igniters should be 10% in the appropriate lacations j of the text and the 60-day interval-to restore a subsystem in the action stats-l cent should be 30 days. Each Mark III evner that intends to adopt the generic .. HIS technical specification must confirm that the HCOG assumptions used in the > development of the TS are valid for their plant-specific configuration. ! l l 1 l l l . l l l l I Mark 313 SER 3 Appendix A
*4: .
APPENDIX.B MARK III COMBUSTIBLE GAS CONTROL: EMERGENCY PROCEDURE GUIDELINE As part of the generic program, the HCOG has developed Combustible Gas Control . Emergency Procedure Guideline (EPG) for plants with Mark III containments.- The latest version of the guidelines with supporting appendices were sent to the , NRC by letter (HGN-122-P) dated July 8, 1988. This procedure includes operator. i action for the hydrogen igniter system as well as other combustible gas control systems designed in the N"k III containment, such as hydrogen recombiners and the drywell= mixing system. In addition, the proposed procedure provides guid- ' ance for, spray actuation ano containment venting.' This effort is to supplement the.overall BWR Owner Group's EPG program.. F16ure B.I . highlights the ' operator actions in dealing with hydrogen in an emer-gency situation. These actions for controlling hydrogen depend on a determina-tion of hydrogen concentration in the containment and drywell as indicated by hydrogen monitors and/or analyzers that obtain gas samples from the containment and drywell. The significant trigger limit used in the EPG is when the drywell or containment hydrogen level reaches a concentration at which a: global deflagra-tion could threaten containment or drywell . integrity from overpressurization, referredtoasthehydrogendeflagrationoverpressurelimit(HDOL). At this limit or when the containment hydrogen concentration cannot be determined to be below the HDOL and it cannot be determined that the igniters have been con- ' tinuously operating, the HIS should not be used. The containment HDOL is a curve of hydrogen concentration versus containment pressure, whereas the dry-well HDOL is a single value representing a peak hydrogen concentration. The containment HDOL is more limiting than the drywell HDOL. The hydrogen ignition system, the hydrogen' recombiners and the drywell mixing-system are the key hydrogen mitigating systems. As indicated in Figure B.1 these systems are activated at appropriate trigger points'to deal with a pro- , gressing hydrogen build-up. With the addition of an independent power source
- l. to the HIS, it is anticipated that for most severe accident / degraded core situations the resulting large amounts of hydrogen can be acconnodated.
As part of the subject letter, HCOG had addressed staff connents which were discussed in a October 22, 1986 HCOG/NRC meeting. In this latest version of theMarkIIICombustibleGasControlEPGs(Revision 3),HCOGhasaddressed . l staff concerns or provided sufficient justification for their position as discussed below.
- As one 9 the initial steps in the EPGs, the operator is instructed to vent the suppression chamber or drywell, whenever either of the respective regions reaches the minimum detectable hydrogen concentration (0.5%), provided the offsite radioactivity release rate is expected to remain below thFoffsite release rate limiting condition for operation (LCO). It should be noted that this step is similar to the BWR EPGs for Mark I and Mark II combustible gas l
l Mark III SER 1 i
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i o . l control. .The staff previously commented that venting may not be necessary . l solely upon hydrogen concentration above the minimum detectable level and below flamability levels; the use of recombiners is valuable and should.be ' utilized where appropriate, g . In response, since dissolved hydrogt,n is present in the reactor coolant' system l during normal operation and the EPGs are based on a symptomatic approach, it l is the intent of this step to remedy a hydrogen problem during normal operation, and within the constraints of technical specification limits. HCOG believes there is sufficient guidance to preclude this action from being implemented ! during a genuine emergency situation. Also. HCOG had comitted to modify %e generic Hark III procedure at later date, if necessary, to be consistent with the BWROG's approved combustible gas control procedure. The staff finds that the subject procedure and the HCOG approach to be acceptable. As one of the last steps to control hydrogen accumulation during a progressively worsening situation, containment venting is directed. Venting the containment ' irrespective of.the offsite radioactivity release rate would only be considered to restore and maintain the containment hydrogen concentration below excessive limits. Containment failure may follow if a large deflagration were to occur. ' Venting the containment may be the only mechanism which remains to prevent an-uncontrolled and unpredictable breach of the containment. The controlled re-lease of radioactivity to the environment is prt.ferable.to containment failure whereby, adequate core cooling might also be lost and radioactivity released with no control. This concept of venting is-similar to the emergency procedures for pressure control. Regarding the second issue, HCOG provided additional information responding to NRC comment dealing with the limited use of the drywell mixing systems. The staff views containment venting as a last resort to deal-with extraordinary conditions. The use of the drywell hydrogen mixing system may delay contain-ment venting by diluting the containment volume (at a higher concentration of hydrogen) with the drywell volume (at a lower concentration of hydrogen). HCOG , cited various factors to demonstrate the drywell mixing system is not beneficial i for hydrogen control inside the containment volume, which includes: dilution ' effects are marginal, since the containment is significantly larger than the drywell; the mixing system would re-initiate a LOCA signal and potentially interfere with event recovery; and implementing a modified procedure may induce conflicting direction. In addition, the design intent of the mixing system is to der.1 with hydrogen in the drywell. The staff believes some of HCOG concerns are valid. In addition HCOG had modified its procedures to assure that the HIS would remain opera,tional above the HDOL if it can be determined that the igniters have been continuously operating. The addition of an independent power supply to the MIS would further enhance the reliability of the system. Consequently, the added reliability would reduce the potential for containment venting to' control hydrogen inside containment. Therefore, the staff agrees with HC0G that the inclusion of the drywell mixing system would not provide significant benefits (in delaying venting) as compared to its disadvantages. Overall,thestafffindstheproposed(Revision 3)MarkIIIcontainmentEPGs are based on sound technical judgment, and are acceptable. Accordingly each Mark III licensee should address its combustible gas control emergency p,roce-dure in the plant specific final analysis. Hark III SER 2
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._iz______________._ 'E._.________._....__.. ,~_e--
w. y . i p APPENDIX C H BIBLIOGRAPHY <
]
H Gaunt, R. D. et al. ,-"The DF-4-BWR Control Blade / Channel Box Fuel Damage: ! Experiment," Sandia National Laboratories, to,be publishedi 1 Hydrogen Control owners Group (HC0G), HGN-003,; letter from J. D. Richardson-(HC0G) i to;H. R.E Denton (NRC), " Report on Hydrogen Control Accident. Scenarios, Hydrogen -i Generation Rates and Equipment Requirements," . April 8,1982. '
~
-- , HGN-006, letter from J. D. Richardson (HCOG) to H. R. Denton-(NRC), " Report' l on Hydrogen Control Accident Scenarios ~. Hydrogen Generation _ Rates and Equipment !
Requirements," Rev. 1,-September 9,.1982.
-- , HGN-009-P, letter from J. D. Richardson (HC0G) to H. Denton (NRC), "CLASIX-3 Report," March 16, 19B3.
-- , 01.1-NP, letter from~J. D.. Rich'ardson (HCOG) to R. W. Houston'(NRC), "Re-t sponses to NRC Requests for Additional Information," May 11, 1983. '
r -- , 012-NP, letter from S. H.- Hobbs (HCOG) to R. Bernero (NRC), "1/4 Scale Hydrogen Test Program," August 12, 1983.
-- , NGN-014-NP, letter from S. H. Hobbs (HCOG) to H. Denton (NRC), " Final.
1/20th Scale Test Report," February 9,1984.
-- , HGN-017-NP, letter from S. 'H. Hobbs (HC0G)-to H.' Denton (NRC), " Mark III
' Hydrogen Control Owners Group Final Whiteshell Ignition Test Report," June 7, 1984.
-- , HGN-020, letter from S. H. Hobbs (HCOG) to H. R. Denton (NRC), " Transmittal of BWR Core Heatup Code Manual," September 5, 1984.
j
-- , HGN-027-NP, letter from S. H. Hobbs (HCOG) to R. Bernero (NRC), "1/4 Scale Test Facility 3D-Complex Calorimeter," February 13, 1985' .
-- , HGN-031, ' letter from S. H. Hobbs (HC00) to R. -Bernero (NRC), " Hydrogen l Release Histories and Test Matrix for 1/4 Scale Test Program," March 13, 1985, t
' -- , HGN-032, letter from S. H. Hobbs (HCOG) to R. Bernero (NRC), " Submittal of Information on BWR Core Heatup Code,' ' April 16, 1985..
-- , HGN-034, letter from S. H. Hobbs (HOOG) to R. Bernero (NRC), "Model for Hydrogen Production Equivalent to 75% MWR," May 17, 1985.
-- , HGN-051, letter from S. H. Hobbs (HCOG) to R. Bernero (NRC), " Availability of Containment Spray System," July 26, 1985. .
l l l 1 Appendix C
- n. , - - - - . . - . _ . , . ._, . - _ - .-- . .
' J'
.. 8 l
l . p 1
-- , HGN-052, letter from S. H. Hobbs (HCOG) to R. Bernero (NRC), " Hydrogen- {
Release Time Histories, August 1,1985. j
~
-- , HGN-055, letter from S. H. Hobbs (HC00) to R. Bernero (NRC)..* Evaluation l cf SBD and ATW5 Contributions to Hydrogen Generation Events," September 27,
.. 1985. ,
l
-- , HGN-070, letter from J. R. Langley (HCOG) to R. Bernero (NRC), " Mark III: 1 1
l Hydrogen Ignition System Technical Specifications," April 16, 1986. L --- , HGN-072, letter from S. H. Hobbs (HC0G) to R. Bernero (NRC), " Event Sce- ~ narios Considered for Evaluation of Drywell Response to Degraded Core Accidents,"
- March 5, 1986.
1
-- , HGN-073, letter from J. R. Langley (HCOG) to R. Bernero (NRC), "Justifica-tion for Manual Actuation of Mark III Hydrogen Ignition Systems," March 5,1986.
-- , HGN-084, -letter from J. ' R. Langley (HC0G) to R. Bernero (NRC), " Generic Equipment Survivability List," May 16, 1986..
-- , HGN-085, letter from J. R. Langley (HCOG) to 2. Bernero (NRC), " Final Re-port May 5, Assessing 1986. Adequacy of Heat Sink Hodeling in the 1/4 Scale Test Facility,"-
-- , HGN 089, letter from J. R. Langley (HCOG) to R. .Bernero (NRC), "8WR Core Heatup Code Responses," June 9,1986.
-- , HGN-091, letter from J. R. Langley (HCOG) to R. Bernero (NRC), " Criteria
- for Existence of Inverted Diffusion Flames in the Drywell," June 25, 1986.
-- , HGN-092-P, letter from J. R. Langley (HCOG) to R. Bernero (NRC), " Report of CLASIX-3 Generic Analyses.and Validation of CLASIX-3 Against 1/4 Scale Test i Facility Data," June 10, 1986.
l l
-- , HGN-096-P, letter from J. R. Langley (HCOG) to R. Bernero (NRC), "8WR Core Heatup Code Report," July 30, 1986. l l
-- , HGN-098-P, letter from J. R. Langley (HCOG) to R. Bernero (NRC), " Scoping '
l Test Report," July 18, 1986. l
-- , HGN-095-P, letter from J. R. Langley'(HCOG) to R. Bernero (NRC), " Transmit-tal of Handouts from June 19, 1986 HCOG-NRC Meeting," July 11, 1986.
l -- , HGN-100 Tetter from J.- R. Langley (HCOG) to R. Barnero (NRC), " Nevada Test i Site Data Evaluation," July 31, 1986.
-- , HGN-101, letter from J. R. Langley (HCOG) to H. Denton (NRC), " River Bend Staion Unit Coolers," July 30, 1986. .
-- , HGN-103, letter from J. R. Langley (HC0G) to R. Bernero (NRC), " Diffusive '
Combustion Thermal Environment Methodology Definition Report " July 30, 1986. 2 Appendix C
'of ' . ' ;
l: .
-- , H3N 104-P letter from J. R. Langley (HCOG) to R. Bernero (NRC), "Evalua-l tion of Emergency Procedure Guidelines Operator Actions Against HCOG Assumptions L for Analysis of a Hydrogen Generation Event," August 18, 1986. .
I l
-- , HGN 105-P, letter from J. R. Langley (HC00) to R. S-wro (NRC) " Diffusive .
i combustion Heat Transfer Methodology validation for Equip...it Survivability ! j in Mark III containments," August 29, 1986. l -- , HGN 106-P, letter from J. R. Langle l cental Information on Secondary Surning,g September (HCOG) 29, 1986. to R. Bernero (NRC), " Supple-l
-- , HGN-109-P, letter f rom J. R. Langley (HCOG) to R. Bernero (NRC), *CLASIX-3 .;
Generic Sensitivity Analyses," December 9, 1986.
-- , HGN-110-P, letter from J. R. Langley (HCOG) to R. Bernero (NRC), " Combust-ible Gas Control Emergency Procedure Guideline and Supporting Appendices," <
December 1, 1986.
-- , HGN-111-P, letter f rom J. R. Langley (HCOG) to NRC, *CLASIX-3 Summary Re-port," December 15, 1987. .
-- , HGN 112-NP, letter from J. R. Langley (HCOG) to R. Bernero (NRC), " Generic
- Hydrogen Control Information for SWR-6 Hark III Containments," February 23, 1987.
4 -- , HGN-113, letter from J. R. Langley (HCOG) to R. Bernero (NRC), " Comparison i of CLASIX-3 Predictions to Nevada Test Site Data " January 8,-1987.
-- , NGN-114, letter from J. R. Langley (HCOG) to R. Bernero (NRC), " Response
- to NRC Concerns Regarding Potential' Impact of SB0 Events on Operation and Per-formance of the Hydrogen Ignition System," January 8,1987.- . .
-- , HGN-115-NP, letter from J. R. Lang1,ey (HCOG) to R. Bernero (NRC), *1/4
- Scale Test Facility Final Design Report,' February 10, 1987, i
j -- , HGN-118-P, letter from J. R. Langley (HCOG) to NRC, " Generic Equipment
- Survivability Analysis," August 7, 1987.
1 l -- , HGN-119, letter from J. R. Langley (HC0G) to NRC, " Final Inverted Diffusion-Flame Report," June 10, 1988. j -- , HGN-121-P, letter from Langley (HCOG) to NRC, " Report of Hydrogen Combustion Experiments in'a 1/4 Scale Model of a Mark III Nuclear Reactor Containment," July 22,1987. 1' -- , HGN-122-P letter from J. R. Langley (HCOG) to USNRC, " Revision 3 to Mark III Combustible Gas Control Emergency Procedure Guideline," July 8,1988. . -- , HGN-123, letter from J. R. Langley (HCOG) to USNRC,
- Response to NRC Ques-tions on Station Blackout and ATWS Sequences," September 9,1987.
-- , HGN-128, letter from J. R. Langley (HCOG) to NRC, " Responses to NRC/SNL Comments on Heat Loss Report," November 6, 1987.
i -- , NGN-129-P, letter, J. R. Langley (HCOG) to NRC, " Revision 4 EPGs vs HC0G
; Scenarios," April 8, 1988.
3 Appendix C
~
__,_..m._.__.,_.
,. . 3
.. - . _ l
-- , HGN-231-P letter, J. R. Langley (HCOG) to NRC, " Supplemental Discussion of Pressure Transmitter Survivability at Low Hydrogen Release Rates," April 5,
-1988.
---, HGN 132, letter, J. R. Langley (HCOG) to NRC, " Final HCOG Hydrogen Release
-Histories," April 4, 1988.
~
- r, HCOG/NRC meeting presentation, "BWR Core Heatup. Code Results," August 28, 1984.
-- , HCOG/NRC meeting presertation, G. R. Thomas -(HC0G), " Review of Basis fo'r 2ircaloy Oxidation Cutoff itothe BWR Core Heatup. Code," January 14.-1985. !
International Technical Services (ITS) ITS/ LWR /8NL 85-1, " Validity of the Use , of a Temperature Cutoff for 2ircaloy Oxidation," by H. Komoriya and P. Abramson, l September 1985.
, l
-- , letter from H. Komoriya (ITS) to L. Lois (NRC), " Review of Oxidation Model-ing in 8WR Core Heatup Code," October 30, 1987.
Institute of Electrical and Electronics Engineers IEEE Std 323 1974, "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations." ! Mississippi Power and Light Company, letter from L..F. Dale (Mississippi Power and Light Company) to H. Denton (NRC), "Hr Igniter Environmental Qualifications Test Results," February 14, 1983. , Sandia National Laboratory, letters from S. E. Dingman to A. Notafrancesco (NRC), June 26, August 10, September 3 September 4, and December 23 (two letters), 1987. > Science Application Inc., SAIC/87/3114 " Hydrogen Generating Events for Boiling Water Reactors with Mark III Containments," by Atefi et al., April.15, 1988' , l U.S. Nuclear Regulatory Commission, Title 10, Code of Federal Reculations, ) 10 CFR 50.44, " Hydrogen Control Requirements."
-- , letter from R. Bernero (NRC) to S. H. Hobbs (HC0G) dated June 4,1985.
l
-- , letter from R. W. Houston (NRC) to J. R. Langley (HCOG) February 21, 1986.
-- , NUREG-0011. " Safety Evaluation Report Related to the Operation of Sequoyah Nuclear Plant, Units I and 2," Supplement 6 December 1982.
-- , NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," Rev.1, July 1981. )
-- , NUREG-0831, " Safety Evaluation Report Related to the Operation of the Grand Gulf Nuclear Station, Units I and 2," Supplement 3. July 1982, and l Supplement 5, August 1984. '
-- , NUREG-0853, " Safety Evaluation Report Related to the Operation of Clinton t
Power Station, Unit No.1," Supplement 6, July 1986. 4 Appendix C l
4
.. t
-- , NUREG-0887, " Safety Evaluation Report Related to the Operation'of Perry ;
Nuclear Power Plant,- Units It and 2." Supplement.6,' Ap?il.1985.
-- , NUREG-0979, " Safety Evaluation Report' Re' lated to the Final Design Approval' of the GESSAR II BWR/6 Nuclear Island Design," Supplement No. 2. November 1984.
i
-- , NUREG-0989, " Safety Evaluation Report Related to the Operation of- River -
Bend Station," Supplement 4. September 1985.
-- , NUREG-1150, " Severe Accident Risks: an Assessment for Five U.S. Nuclear.
Power Plants " Second Draft, June, 1989.
-- , NUREG/CR-1659. Vol. 4 of 4. " Reactor Safety Study Methodology Applications Program-(RSSMAP) " Grand Gulf Power Plant Unit 1, November 1981.
-- , NUREG/CR-4866; "An Assessment of Hydrogen Generation for-the PBF Severe.
Fuel Damage Scoping and 1-1 Tests." A. W. Cronenberg et al... EG&G April-1987. i
-- , NUREG/CR-5079, " Experimental Results Pertaining to the Perfonnance of Ther-mal Igniters." Sandia National Laboratory, to be published.
Vinjamuri, K., et al., " Severe Fuel Damage Test 1-4 Data Report," EG&G, Septem-ber-1987. : f l i. l' l 0 l S APPENDIX C
} ;$
'& W -l r ,
s ( 'i 3 t 4 g p 4 APPENDIX D
~
, ACRONYM LIST i
AD51 automatic. depressurization ~ system , , g - L BWR - . boiling-water reactor . L: ' SWRCHUCL boiling-water-reactor, core heatup code
- DWB : Lsmall pipe break'in the drywell-EPG' emergency procedure guideline
- ESF- engineered safety feature FMRC; Factory Mutual Research Corporation HCOG Hydrogen Control' Owners Group (Mark Ill Containment)-
HCU- ' hydraulic control unit HDDL ' hydrogen deflagration overpressure limitL HGE hydrogen generation events H15 hydrogen ignition system ' HPCS high pressure core spray 3EEE- -1 Institute of Electrical-and Electronics Engineers- , LCD limiting. condition of operation '
-LOCA r loss-of-coolant accident- '
LWR light-water reactor '
- MiR metal-water reaction ,
NRC Nuclear Regulatory Comission PWR pressurized-water reactor. QSTF quarter-scale test facility ', i RCIC reactor core isolation cooling
.RPV.- reactor pressure vessel .
- RSSMAP reactor safety study methodology applications program RWCU reactor water cleanup i
~
SAIC Science Applications Inc..(report designation) .
- SBLOCA- small-break loss-of-coolant accident
$B0 station blackout SER- safety. evaluation report SNL Sandia National Laboratory -
50RV stuck-open relief valve SRV safety relief valve
- Mark 111 SER 1 Appendix D .
,t, s .y;J. o _ y.. a _ o . * ' t -. , i , }e
- L ~
- .- 3 ,
, g. . . , s-
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, 'i 1
'n A
'; TAF E 10'CFR; top of active fuel'. . .. .
< -TSt
. Title 10 of the Code of Federal Reculations -
. technic'al-specifications
.i
. : : 2r; -' aireonius. 3l '
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Mark III SER 2 Appendix D
. }$.
a -. - s q i)) L j
. LMr.: Michael D. Lyster. Perry Nuclear Power Plant
.The Cleveland Electric Unit ~1 Illuminating Company r
;cc: Jay E. Silberg, Esq. . . Mr. James W. Harris, Director _ ,
Shaw, Pittman, Potts & Trowbridge- Division'of, Power Generation-
' 2300 N Street,' N.W. ,
Ohio Department of. Industrial , Washington, D.C. 20037' Relations L P. O. Box 825- , David E. Burke ~
- Columbus,' Ohio ,43216 ;
The Cleveland Electric Illuminating Company The Honorable. Lawrence Logan l LP.O. Sox 5000. Mayor, Village of Perry
-Cleveland, Ohio 44101 4203 Har)er Street _ >
l . _ Resident Inspector's Office Perry, 0110 44081 j l U.S. Nuclear Regulatory Commission- ' The Honorable Robert V..Orosz Parmly at Center Road: Mayor, Village of. North Perry & Perry, Ohio E44081- North Perry Village.Ha'il 4778-Lockwood Road Regional Administrator, Region III- North Perry Village, Ohio 144081 U.S. Nuclear Regulatory Commission 799 Roosevelt Road Attorney General Glen Ellyn, Illinois 60137 Department of Attorney General. 30 East Broad Street = , Frank P. Weiss, Esq. Columbus, Ohio 43216 Assistant Prosecuting Attorney 105 Main Street - Radiological Health-Program Lake County Administration Center Ohio Department of Health
- Painesville, Ohio 44077 1224 Kinnear Road
; Columbus,:0hio 43212 Ms. Sue Hiatt OCRE Interim Representative Ohio Environmental Protection 8275 Munson Agency. ._
!. Mentor, Ohio 44060 DERR--Complianca Unit , PO Box.1049 Terry J. Lodge, Esq. 1800 Watermark Drive 518 N. Michigan Street ATTN: Zack A. Clayton Suite 105 Columbus, Ohio 43266-0149 Toledo, Ohio 43624 John G. Cardinal, Esq. Perry Township Board of Trustees
. Prosecuting Attorney Box 65
- 1. Ashtabula County Courthouse 4171 Main Street Jefferson, Ohio 44047 e Perry, Ohio 44081 Robert A. Newkirk State of Ohio ,I Cleveland Electric Public Utilities Commission l Illuminating Company 180 East' Broad Street 4
Perry Nuclear Power Plant Columbus, Ohio 43266-0573 P. O. Box 97 E-210 Perry, Ohio 44081' i i 4 4 l}}