ML20045G263
| ML20045G263 | |
| Person / Time | |
|---|---|
| Issue date: | 06/30/1993 |
| From: | Graves C NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| REF-GTECI-106, REF-GTECI-NI, TASK-106, TASK-OR NUREG-1364, NUDOCS 9307130111 | |
| Download: ML20045G263 (48) | |
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NU REG-1364 i
Regu a:ory Ana:ysis :?or tae Reso:tr: ion 0:? Generic Sa:?e:y Issue 106: Pi;]ing anc ile Use 0:? Higiy Connus:i1e Gases in Vita. Areas i
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U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research I
C. C. Graves
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NUREG-1364 Regulatory Ana ysis for the Reso ution of Generic Safety Issue 106: Piping and the Use of Highly Combustible Gases in Vital Areas Manuscript Completed: June 1993 I) ate Published: June 1993 C C. Graves Division of Safety issue Ilesolution Omce of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 f.a-.oy,
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AllSTRACT Ilighly combustible gases such as hydrogen, propane, and gas used for maintenance at l' Wits and llWits is small.
acetylene are used at all nuclear power plants. Ilydrogen On the basis of generie evaluations, the N1(C staff has is of particular importance because it is stored in large concluded that several possible methods to reduce risk quantities and is distributed and used continuously in could provide cost effective safety benefits at some
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buildings containing safety-related equipment. large hy-plants. Ilowever, in view of the observed large differences drogen releases at the hydrogen storage facilities or in in plant-specific characteristics affecting the risk associ-these buildmgs could lead to fires or explosions that might ated with the use of hydrogen, and the marginal generic result m loss of salety-related equipment. Ihis report safety benefit that can be achieved in a cost effective gives the regulatory analysts for the resolution of Generie Safety issue 106," Piping and the Use of liighly Combus-
- """Cf' it is recommended that this generic issue be resolved su.nply by making these results available in a tible Gases in Vital Areas." Scoping analyses showed that the risk associated with the storage and distribution of generic letter. This information may help. licensees in hydrogen for cooling electric generators at boiling water their plant evaluations recommended by Generie Letter reactors (llWils), the off-gas system at IlWils, the waste 88-20, Supp!cment 4, " Individual Plant lixamination of gas system at pressurized-water reactors (l'Wils), and I!xternal F. vents for Severe Accident Vulnerabilities /
station battery rooms and portable bottles of combustible June 28,1991.
4 iii NUltliG-1364
CONTENTS Page Abst ract.............
iii F.x ecu t iv e S u m ma ry........................................................................ vii 1 S ta t e me n t Of Probl em..............................
1 1.1 I n t rod u c t io n....................................................................
I 1.2 Existing Regulatory Requirements and Guidelines.....
1 1.2.1 General.
I 1.2.2 Regulations and Regulatory Guides.....................
1 1.2.3 Standard Format and Content (NUREG-75/094), Standard Review Plan (NUREG-0800), and Technical Specifications...............................
2 1.2.4 Electric Power Research Institute Guidelines.....................................
2 1.2.5 N R C Inspection Program....................................................
3 1.3 Scope 3
l 2 Objective 5
i 3 Al t e rnativ e R esol u t io n s.................................................................
7 j
3.1 Ilydrogen System Characteristics for Volume Control Tank and Generator Applications.........
7 3.2 Hydrogen Combustion Characteristics................................................
8 3.3 Design Features at Operating Plants Used To Reduce Risk From Hydrogen Systems for the Volume Control Tank and Generator...........................................
8 3.4 Alternatives for Volurne Control Tank and Generator Hydrogen Distribution Systems and Hydrogen Stomge Facilities..........................
10 4 Technical Findings.....
13 4.1 Introduction................
13 4.2 Review of U.S. Nuclear Power Plant Hydrogen Precursors................
13 4.3 M e t h od o f Analysis............................................................. 14 4.3.1 General................................................................
14 4.3.2 S eism ic Ev en t s............................................................ 15 4.3.3 Non Scismic Events....
16 4.3.4 Damage Mechanisms......
16 4.3.5 Hydrogen Explosion Methodology 16 4.3.6 Uncertainty Analysis 17 4.4 Analysis of Base Case PWR Plant...
17 4.4.1 Tu rbine B uilding...................................
17 4.4.2 Primary Auxiliary Building.....
18 4.4.3 I lyd rogen S torage Facilities................................................. 18 4.4.4 Uncertainty Analysis Results......,...
18 4.5 Generic Analysis 18 4.5.1 M e t h od....................
18 4.5.2 Generic Analysis Results.
19 4.6 Reductions in Core Damage Frequency for improvement Alternatives 20 4.7 Dose Consequence Analysis.
20 4.8 Cost Analysis..
20 v
Page 5 Cost-lienefit Analysis..
23 5.1 Alternative 1-Take No Action 23 5.2 Alternative 2-Install low Setpoint I!xcess Flow Valves. Restricting Orifices, or Check Valves in the Ilydrogen Supply 1 ine to the Generator..............
23 5.3 Alternative 3-Provide Manual Makeup of flydrogen to Generator and Check Valve or Restricting Orifice at the Generator 23 5.4 Alternative 4-linclose Safety Related Fquipment Incated in Turbine lluilding in tilast. and I ire-Proof Structures...
24 5.5 Alternative 5-install 1 ow Setpoint lixcess Flow Valve or Restricting Orifice in liydrogen Distril ution System to Volume Control Tank and Provide Hydrogen Detectors,if Needed 24 5.6 Alternative 6-1 imit the Quantity of Ilydrogen Normally Connected to Volume Con t rol Tan k.........
24 5.7 Alternative 7-Provide Normally Isolated Supply With Daily Manual Makeup of Ilydrogen to Volume Control Tank...
24 5.8 Alternative 8-Relocate Hydrogen Storage Facility To Meet Separation Distance From Safety-Related Structures.
25 59 Alternative 9-install Blast Deflection Shield at flydrogen Storage Facility.......
25 5.10 Alternative 10-Install flydrogen Analyzer-Actuated Air Intake Louvres at Safety-Related Air intakes..
25 5.11 Other Alternatives..
25 5.12 I ife listension Considerations 25 5.13 New Reactors 25 6 Decision Rationale 27 6.1 Introduction 27 6.2 Relationship to Other Generic issues 28 6.3 llackfit Rule and Plant-Specific Considerations....
28 6.4 Conclusion.
28 7 References 29 TAllllS 1
Number of Hydrogen Events at liach Plant location 32 2
Generic Plant Configurations 32 3
Delta Core Damage Frequency per Reactor-Year for Alternatives (Calculated Mean Values)........ 33 4
Cost of Modifications Minus Onsite Averted Costs (Point Estimates) for Remaining Plant 1.ife of 20 Years 34 5
Cost of Modifications Minus Onsite Averted Costs (Point Estimates) for Remaining Plant Life of 40 Years 35 6
Cost-lienefit Uncertainty Results for Remaining Plant Life of 20 Years 36 7
Cost.llenefit Uncertainty Results for Remaining Plant Life of 40 Years 37 NURI!G-1364 vi
J EXECUTIVE SUMMAlW This report gives a cost-benefit analysis and supporting a stuck-open PORV or safety valve. This transient in-information for U.S. Nuclear Regulatory Commission's ciudes such events as station blackout and loss of compo-(N RC's) resolution of Generic Safety Issue 106 nent cooling water or service water, (GSI-106), " Piping and the Use of Ilighly Combustible Gases in Vital Areas /' The scope of GSI-106 includes
'the failures considered for the storage facility were (1)a i
hydrogen storage facilities and hattery rooms at hydrogen tank rupture resulting in a detonation at the pressurized water reactois (PWRs)and boiling water re.
facility and blast damage to a nearby safety-related struc-actors (BWRs); waste gas systems at PWRs; off-gas sys.
ture and (2) a pipe failure at the facility with a large tems at ilWRs; hydrogen distribution systems for electric release of unburned hydrogen and ingestion of a flamma-generators at PWRs and llWRs and the volume control ble hydrogen-air mixture at a safety-related air intake, tank (VUF)in the chemical and volume control system at Postulated failures of the hydrogen distribution systems PWRs: and small, portable bottles of combustible gas in the auxiliary and turbine buildings were leaks or breaks such as hydrogen, propane, and acetylene. The scope with large hydrogen releases and subsequent deflagra-does not include large amounts of liquified petroleum gas tions or detonations or smaller undetected leaks resultmg at PWRs and BWRs or the gaseous and liquid hydrogen in the buildup and subsequent detonation of large storage and distnbution systems for hydrogen water amounts of trapped hydrogen.
chemistry installatiena at IlWRs covered under Licensing
.The alternatives considered to reduce a vulnerability m Issue 136.
the storage area included (1) relocation of the storage area,(2) installation of a blast shield to prevent unaccept-Idaho Natior.at lingineering I.aboratog (INiiL) provided able blast damage, and (3) installation of shutters actu-technical atsistance for resolvmg this isse. Scopmg and ated by hydrogen detectors to prevent ingestion of flam-screening analyses by INEL mdicated that the nsk associ mat Ic hydrogen-air mixtures at safety-related air intakes.
ated with the use of hydrogen for electric generators at For the turbine building, the alternatives were (1) use of IlWRs. bittery rooms, PWR waste gas systems, llWR excess flow valves and check valves or restricting orifices off-gas ss st ems, and portable bottles of combustible gases to limit flow from the corage facility and generator to a was small therefore, the more detailed risk and cost" break in the hydrogen supply line, (2) use of a normally benefit analyses by INEL were limited to the hydrogen solated hydrogen supply vith periodic manual makeup storage facilities and to the hydrogen distnbution systems and a check valve or restric ing orifice to limit back flow to for the VCI,and generator at PWRs.These facilitics and the break from the gen:rator, and (3) modifications to distribution systems are not categori/ed as safety related.
protect safety-related equipment from the consequences However, because of the use of large quantities of hydro' of hydrogen deflagrations or detonations. For the auxil-gen for these applications, there is a potential for damage lary building, the alternatives considered were (1) use of to safety-related equipment because of hydrogen defla' excess flow valves or restricting orifices to limit the flow grations or detonations. 'Ihe basic regulatory require-rate from larr;e hydrogen supplies to the break, supple-ment pertment to GSI-106 is General Design t riterion 3 mented by hydrogen detectors, if needed; (2) use of lim-m Appendtx A to Part 50 of'lille 10 of the Code offederal ited hydrogen supplies; and (3) use of normally isolated Repdarions.
hydrogen sapplies. These alternatives include adminis-trative controls and design features to prevent inadver-
'ihe hydrogen storage facilities and distnbution lines to tent bypass of the flow limiting devices and limited sup-the VCrand generator are not near the pnmary coolant plies, to moaitor for hydrogen leaks, and to isolate the system or reactor pressure vessel. Ilence, hydrogen hydrogen supply following loss of normal ventilation in deflagrations or detonations would not lead to pipe-break the auxilia*y building.
loss-of-coolant accidents (LOCAs), steam generator tube ruptures,oranticipated transient-without-scram se-A number of hydroun events have occurred and continue quences.The remaining core damage events in the proba-to occur in the turbine buildings at U.S nuclear power bdishe risk analysis for this issue were divided into tran-plants. In additior, several large fires involving hydrogen sients with failure of decay heat removal (DHR) systems and oil have occurred in turbine buildings at foreign (T/DilR) and transient-induced I OCAs (T/LOCAs).
plants. Hydrogen events in the turbine building are not The T/DHR transients involve loss of all forms of core expected to be significant sources of risk for most U.S.
coohng and release of reactor coolant at high pressure plants because (1) vital equipment is not located in the from pressunier power-operated relief valves (PORVs) building or (2) recovery operations for T/DHR transients or safety valves. The T/LOCA transients involve failure can prevent core damage (e.g., feed-and-bleed operations of reactor coolant system makeup or recirculation follow-and recovery of main feedwater). 'these plants could ing a consequential reactor coolant pump scal failure (as a suffer significant economic losses because of damage to result of loss of seal cooling)or a PORV l.OCA caused by plant equipment and replacement power costs, but not vn NUREG-1364
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i core damage. Ilowever, some plants are considered Although the adequacy of fire protection measures is susceptible to core damage resulting from hydrogen outside the scope of GSI-106, the staff notes that these esents in the turbine building that lead to T/Dillt tran-measures can include accident-mitigation strategies for sients or T/IDCA transients (predominantly seal turbine building fires such as (1) early climination of the l
l OCAs).
hydrogen sources by isolating the storage facility and venting and purging the generator with carbon dioxide There have been a number of leaks but no fires, explo-and (2) controlling the pumping of lubricating oil that sions, or large hydrogen releases involving the hydrogen could spread the fire, system supplying the VCI'in the auxiliary building.The hydrogen events considered potentially significant are l'or the hydrogen distribution systems in the auxiliary and those causing loss of vital equipment and resulting in a turbine buildings, the generic estimates of the reduction seat i OCA. l_icensees of a number of plants have pro.
in core damage frequency obtained with the alternatives vided corrective measures (e.g., normally isolated or lim.
for the distribution systems ranged up to 0.511-5/ reactor-ited supplies, flow-limiting devices, and leak detection year (l(Y) to m-5/ItY. Ilowever, risk reductions at indi-equipment and procedures) to reduce the risk from this vidual plants may be significantly larger than the generic j
source. Ilowever, some plants are considered to be sus.
values because of the proximity of the hydrogen distribu-ctptible to core damage because of a large storage facility tion system to vital equipment. The estimated costs for and the lack of protective features to prevent large hydro-several of the proposed alternatives are small. Ilence, gen releases.
when considered individually, several of the alternatives analy7ed for reducing the risk for this issue would be cost Since industry will continue to use appreciable amounts effective in meeting the $1000/ person rem guideline.
of hydrogen to cool the generators and for water chemis-
'Ihc analyses for these alternatives indicate cost savings try control, the risk from hydrogen fires and explosions w hen onsite averted costs are included. Ilowever, in view cannot be completely eliminated. Ilowever, a blend of of the obsetted large differences in plant specific charac-accident prevention and -mitigation capabilities may re-teristics affecting the risk associated with the use of hy-duce the risk from these sources to acceptably low levels, drogen, and the marginal generic safety benefit that can
!!xamples of preventive measures iriclude the use of ex.
be achieved in a cost-effective manner,it is recommended cess flow valves or restricting orifices to reduce the possi.
that this generic issue be resolved simply by making these bilityof the release oflarge quantities of hydrogen follow-results available in a generic letter. This information may ing a piping leak or rupture and admimstrative controls or help licensees in their plant evaluations recommended by hydrogen detectors to detect leaks. In regard to mitiga-Generic 1.etter 88-20, Supplement 4, " Individual Plant tion, standard fire protection measures can go far to re.
lixamination of!!xternallivents forSevere Accident Vut.
doce the consequences of a hydrogen fire or explosion.
nerabilities " June 28,1991.
1 h
N dlW G-1364 viii
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1 STATEMENT OF PROllLEM l.1 IlltI*O(Illet.iOlt requires an evaluation of the risk associated with the use
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of hydrogen in nuclear plants.
Combustible gases such as hydrogen, acetylene, and pro-panc are used at all nuclear power plants. Of these gases, 1.2 Existing Regulatory llequirements hydrogen is of principal interest because it is stored in
- gg3() gg3jgp]jg)p3 large quantities and is distributed and used m some safety-related buildmgs durmg normal plant operation. It jy gg is provided as a coolant for the main electne generators at both pressurized-water reactors (PWRs) and boiling-llecause the structures, systems, and components in-water reactors (ilWils). It is also fed to the volume con-volved in the use of hydrogen at nuclear plants are not trol tank (VCT) in the chemical and volume control sys-classified as safety related, they do not have to be scismie i
tem of PWRs to reduce oxygen in the reactor coolant Categon I, environmentally qualified, or redundant.
system.The hydrogen for these purposes is usually stored However, hydrogen fires or explosions or the release of as a high-pressure gas [e.g.,1500 to 2400 pounds per unburned hydrogen could result in damage to nearby square inch gage (psig)] in large storage facilities and is safety related equipment and should be considered in distributed to the auxiliary and turbine buildings through setting design and operational requirements.
small-diameter field run piping. Failure of the piping or bottles / cylinders at the storage area could result in flam' l.2.2 Regtilations aiul Regulatory Guides mable hydrogen-air mixtures at safety-related air intakes or fires or detonations that could damage safety-related The basic regulatory requirement dealing with the stor-structures. Ilence, the storage area should be located at a age, distribution, and use of combustible gases at nuclear safe separation distance from these structures and in-power plants is General Design Criterion (G DC) 3, " Fire takes. The distribution piping to the electrie generators in Protection,"in Appendix A to Part 50 of Title 10 of the the turbine building generally would not be near safety-Code of Federal Regulations (10 CFR Part 50). This crite-related equipment. Ilowever, some plants have safety-rion states, in part, that " structures, systems, and compo-related equipment such as motor control centers, cables, nents important to safety shall be designed and located to switchgear, and diesel generators in, or adjacent to, the minimize, consistent with other safety requirements, the turbine building. In PWRs the auxiliary building, which probability and effect of fires and explosions."
contains the VCT, also contains most of the components of the safety-related systems at the plant. llence, leaks or Section 50.48 of 10 CFR requires that every plant have a breaks in the piping or components of the hydrogen distri-fire protection plan that satisfies G DC 3.The plan should bution system in the auxiliary building at PWRs or in the include descriptions of special features needed to limit turbine buildings at some PWRs and llWRs could result damage to structures, systems, and components impor-in fires or explosions that represent a threat to plant tant to safety so that the capability to safely shut down the safety because of potential damage to safety-related plant is ensured.
equipment. Capacities of individual hydrogen bottles or cylinders at the storage facilitics range from about 200 to Appendix R to 10 CFR Part 50 gives the fire protection 10.000 standard cubic feet (sef) of hydrogen, while total program requirements to meet GDC 3. It includes new amounts of hydrogen in the storage facility may be more requirements and guidance dealing with fire protection than 100.000 scf. This represents a significant potential measur es to limit the damage to systems important to safe shutdown and the use of alternative or dedicated capabil-energy release.1Iydrogen gas also reptesents, to a lesser degree, a potential threat to safety-related equipment ny for areas where fire protection features cannot ensure -
safe shutdown, should a fire occur in that area. Appen-because of its presence in PWR waste gas systems, ilWR off-gas systems, station battery rooms, and small bottled dix R applies to all plants licensed to operate before supplies in plant buildings at PWRs and ilWRs.
Januay 1,1979, except to the extent described in 10 CFR 50.48. Revisior 2 of liranch Technical Position CMill!
9.5-1 in Section 9.5.1, " Fire Protection Program," of A number of events involving hydrogen have occurred at N U RilG-0800 contains revised guidelines, which include U.S. and foreign nuclear plants.These have ranged from the acceptance criteria in Appendix R and 10 CFR 50.48, detcetion of concentrations above the lower flammability for implementing GDC 3 for later plants.
limit in the waste gas system and leaks in the auxiliary building to explosions and very large fires in the turbine In accordance with 10 CFR 50.49,"linvironmental Quali.
building. Although no safety-related equipment appar-fication of I!!ectric liquipment important to Safety for ently has been lost to date in the United States because of N uclear Power Plan ts," non-safety-related electric equip-a hydrogen event, the occurrence of these precursors ment should be environmentally qualified if its failure 1
NURl! Gal 364
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4 under postulated environmental conditions could pre-July 1981, the staff did not use this guidance in its review vent safety-related equipment from accomplishing its of some plants.
safety function.
SitP Section 11.3 (Itevismn 2) provides guidance on de-NltC Itegulatory Guide 1.29. "Scismic Design Classifica.
sign features in gaseous waste management systems for tion," states that parts of non.scismic Category I strue.
both PWits and llWits to satisfy GDC 3 with respect to tures, systems, or components whose continued function hydrogen explosions. l'or systems designed to withstand is not required, but whose failure could reduce the fune.
the effects of a hydrogen explosion, SitP Section 11.3 tioning of seismic Category I structures, systems, or com.
specifies analysis of the process gas stream for potentially ponents to an unacceptable safety level or cause incapaci.
explosive conditions and annunciation both locally and in lating injury to control room occupants, should be the control room. l'or a system not designed to withstand designed and constructed so that a safe shutdown carth.
the effects of the explosion, SitP Section 11.3 specifies quake would not cause such a failure.
two independent gas analy/crs operating continuously to provide two independent measurements.The gas analy/-
ers should annunciate alarms both locally and in the con.
l.2J Simulan! Format and Content trol room. Guidance also is given for systems with recom-(NUltEG-75/094), Situulard Iteview biners and for equipment testing intervals.
I'lan (N UI(EG-0800), and Technical Specincations 1.2.3.3 Technical Specincations The technical specifications for most ilWRs and PWRs 1.2.3.1 Standard l'ormat and Content (NUiti:G-75/094) are expected to include coverage of the explosive gas for gaseous radioactive monitoring mstrumentation NURiiG-75/094 does not have a separate section de.
waste, NitC Generie i.etter 89-01, which addresses the scribmg the hydrogen storage facilitics and distribution relocation of portions of the Radiological liffluent Tech-systems. Ilence, most final safety analysis reports nical Specifications to the Offsite Dose Calculation Man.
(1;S Alts) do not give this information. Ilowever, the staff ual, states that existing requirements for explosive gas has prmided some guidance on acceptable approaches monitoring instrumentation for waste gas systems will be for meetmg GDC 3 with respect to the scope of Generic retained in the technical specifications. It also provides Safety Issue (GSI) 106 in Standard Review Plan (SRP) model specifications for retaining existing requirements (NUR1!G--0800) Section 9.5.1, " Fire Protection Pro-for explosive gas monitoring instrumentation that apply gram," and in SRP Section 11.3, " Gaseous Waste Man.
on a plant specific basis.
agement Systems."
1.2A Electric Power Itesearch Institute Guidelines 1.2.3.2 Standar d Iteview Plan (NUlti;G-0800) liranch Technical Position (llTP) CMl!Il 9.5-1, Revi-W recent hydngen wata demistry (IlWC) installa.
tions for il% Rs that were considered under Ucensing sion 2, in SRP Section 9.5.1, Revision 2, contains some guidance concerning b".e 136, " Storage and Use of large Quantities of C go-geme Lombustibles on Site, involve the storage oflarger quantities of hydrogen and higher average consumption the use of excess flow valves and other protective rates than those encountered in typical applications for e
f eatures for the distribution of hydrogen m safety-hydrogen supplied to the VCT in I!WRs and the main related buddmgs generator in all plants. In 1987 (letter from J. Richardson e
alarms and annunciation for fire detection and loss dated July 13,1987), the staff approved new guidehnes by of ventilation in safety-related battery rooms the lilectric Power ltesearch Institute (IIPRI,1987) for these iIWC installations. The guidelines describe several l
barrier design features and possible additional de-system design features and procedures for the prevention fense in depth in the turbine building to protect or mitigation of the consequences of fires, explosions,or against fires in the turbine od system or the genera-unburned leaks of hydrogen that are in addition to, or tor hydrogen cooling system more restrictive than, those given in SRP Section 9.5.1, Revision 3, and IITP CMiill 9.5-1, Revision 2. They in, The guidance on the hydrogen distribution systems in-ciude cludes the use of scistnic Category I piping, sleeved piping new relations for separation distances between the with the outer pipe sented directly to the atmosphere.or excess flow check valves si/cd so that the hydrogen con.
hydrogen storage location and safety-related strue-centration in alfected areas does not exceed ? volume tures and air intakes that often gise much larger percent. llecause these SitP resisions were published in separation distances than the values from the NURlhl364 2
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National Fire Protection Association cited in SRP the staff included the new concerns in Licensing issue Section 9.5.1 136," Storage and Use of Iarge Quantities of Cryogenic Combustibles on Site" (NURl!G-0933). Licensing Issue e
color-coded piping and warning signs (American Na-
.136 addressed the use of large quantitics of tional Standards Institute Standards A13.1 and Z35.1) hydrogen and oxygen in permanent hydrogen water e
e excess flow check valves, system trips, and other de-chemistry (IlWC) installations being installed at sign features (e.g., hydrogen detectors) to mitigate llWRs to reduce oxygen in BWR piping to control the consequences of leaks or breaks in hydrogen intergranular stress corrosion cracking lines and to perform the intended design function
! quified propane in, for example, proposed systems.
with or without normal ventilation and, as a mmi' for the incineration of radioactive waste mum, a system t rouble alarm and/or annunciator in the main control room In 1987, the staff accepted licensing topical report EPRI periodic testing of excess flow valves used to protect
. NP-5283-SR-A (EPRI,1987).which provided guidelines e
against breaks in hydrogen lines and components for these llWC installations. With the issuance of staff safety evaluation reports on both concerns, Licensing in the letter of July 13, 1987, from J. Richardson, trans-Issue 136 was closed in 1988 (see NURIIG-0933).
mitting the safety evaluation report on the EPRI guide-lines, the staff recommended that the guidelines be ex-A significant number of events involving combustible tended to include hydrogen systems supplying hydrogen gases have occurred at rmicar plants and ranged from-to the VCTin PWRs and for cooling the main electric detection of flammable mixtures and unburned leaks to generators in PWRs and HWRs. Although EPRI has not explosions and fires in turbine buildings. In April 1987, done so, the staff used the separation distance criteria in the staff issued Information Notice 87-20 to the licensees EPRI's guidelines as an initial screening mcchtmism dur-of all plants as the result of a reported leak in the hydro.
ing plant surveys made in this study.
gen piping in the' auxiliary building at the Vogtle nuclear plant; this leak was caused because a conventional globe 1.2.5 NRC Inspection Program valve was used instead of a valve designed specifically for hydrogen. Such notices are not requirements, but cach The NRC Light Water Reactor inspection Program has licensee is expected to review them for applicability to its several inspection requirements pertinent to GSI-106, facility and for consideration of applicable actions.
Inspection Procedure 64704," Fire Protection Program,"
which specifies an inspection frequency of once every In another instance, during a visit to the Trojan nuc! car other systematic assessment of licensee performance cy*
plant in April 1989, N RCinspectors noted that the hydro-cle, is required for all operating plants. This procedure gen storage facility was located on the roof of the control includes icviews of control of combustible material, re-building near air intakes. This increased concerns about duction of fire hazards, and fire control capabilities. Sites similar hazards at other nuclear plants. As a result, the for storing combustible gas and hydrogen lines in safety-staff issued Information Notice 89-44. In addition, it related areas are specifically identified in this inspection asked cach NRC regional office to supply information on procedure. In addition to the plant and NRC inspections, the size of hydrogen tank farms and the separation dis-this procedure covers three audits required under the tance from safety related structures and air intakes at all technical specifications: (1) an annual audit by the offsite plants in that region.
fire protection specialist, (2) a 2-year audit by the licen-sec's quality assu rance organization, and (3) a 3-year audit As a result of these and ather events involving combusti-by a consulting fire protection finn.
ble gases, the staff expanded the scope of GSI-106 in 1989 to include both PWRs and BWRs. The expanded 1.3 Scope scope included risk from the storage and distribution of hydrogen for the
' in 1981, the staff identified GSI-106 in NUREG-0705, e
" Identification of New Unresolved Issucs in U.S.
VCTin PWRs and the main generators in PWRs and Commercial Nuclear Power Plants." The work on this BWRs issue was directed initially at the risk associated with the other sources of hydrogen such as battery rooms, the e
distribution of bydrogen to the VCI,in the safety-related waste gas system in PWRs, and the off-gas system in i
t auxiliary building of I WRs. In 1986, the staff considered gwg3 j
expanding the scope to include the new concerns associ-
)
small quantitics of hydrogen and other combustible 1
ated with the storage of large quantitics of liquid propane e
and the cryogenic storage of hydrogen and oxygen at gases such as propane and acetylene that would be i
reactor sites. Instead of expanding the scope of GSI-106, used for maintenance, testing, and calibration 3
~
d The scope did not include risk from the hydrogen storage petroleum gas that were considercd under I.icensirig (gaseous or liquid) and distnhution systems for flWC issue 136.
Installations at ilWits or the larger i uantities ofliquified l
i 4
i d
1 k
NUlt!!G-1364 4
L
d 2 OlljECTIVE The purpose of the (icneric Safety Issue 106 program is to 30 years would he about 0.03 based on a population of evaluate the risk associated with the use of combustible about 110 plants. A similar objective (l.Oli-5/ItY) was gases for certain applications at nuclear power plants and noted in USI A-44,"livaluation of Station Illackout Acci-to examine the cost effectiveness of alternative measures dents at Nuclear l'ower Plants." and in GSI-130,"lissen-for reducing this risk.
tial Service Water System l'ailures at Multi-Unit Sites."
The application of such objectives to GSI-106 was limited Probabihstic risk analysis techniques were used to esti-to using these in. sights as general guidelines for the deci-mate the reduction in core damage frequency (CDF)and sion process described in Section 6. Itipid application of the cost effectiveness of the alternative actions. l'or Un-such a quantitative objective to define an absolute re-resolved Safety I3 sue (USI) A-45 " Shutdown Decay lleat quirement is discouraged.This is consistent with the pol-Removal Requirements,' the staff recommended in icy guidance in NRC's " Safety Goals for the Operation of NURI.G-1289 that the frequency of events related to Nuclear Power Plants" and in the memorandum frorn C.
decay heat removal failure leadmg to core damage should J. IIeltemes dated August 20,1991. Similarly, the crite-be reduced to such a level [aixiut 1.0U.-5/ reactor-year rion for cost effectiveness was assumed to be $1000 per.
(ItY)l that the probability of such an accident in the next person-rem averted.
4 i
5 1
I i
4 5
m
..,--.e-,,-
,,..w--n-
--r
,-,--n m
,,,-,,,n
,m+
-,m w
3 ALTEltNATIVE ItESOLUTIONS As the result of scoping and seteening analy ses, the possi-psia. I lence, afte r generator maintenance, m er 35,000 scf ble significant sources of risk associated with combustible may be required for purging cmbon dioxide and reaching gases within the scope of GSI-106 were reduced to the the desired hydrogen purity and pressure conditions. As-following areas (see Section (1):
suming a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period for this infrequent purging and filling operation, an average flow rate of about 100 scfm (1) the hydrogen distribution systems for the volume would be needed.
control tank (VUI) and the electric generator at pWlts The hydrogen supply line leads through the turbine build-ing to the final pressure regulator at a hydrogen control (2) the hydrogen storage facilities at PWRs station located below the generator and one or two levels below the operating deck. This station also has controls -
for venting and purging and instrumentation for monitor-3.1 Ilydrogen System Characteristics ing generator conditions. At the bottom of the typical ~
for Vollline control Tank and generator, there are connections ror about nine iines Gellerator Applicatiolls leading to the control station (gas sample lines, carbon damde and hydrogen supply dnes, and dram Imes). l.caks The hydrogen distribution systems for VCT and genera, or breaks in any of these lines would result in large re-tor applications at PWits have several characteristics per.
leases of hydrogen, which would not be limited by the flow tinent to the selection of corrective measures to reduce restrictitm provided by pressure regulators. In addition, risk. For both applications there is a relatively small aver, the wal oil unit that supphes od, to the generator shaft seals is located on one of these levels.
age hydrogen consumption rate and an infrequent need for the use of larger quantities over short intervals. In addition, the normal operating conditions and gas vol.
The low average consumption rates for both the VCTand umes result in the storage of large quantities of hydrogen generator mean that monitoring consumption rates with in the generator, severat hundred standard eubic feet (scf) flow totalizers or changes in system pressure with time in the V(T, and negligible quantities in the piping, (e.g., for small supplies or operation with normally iso-lated VCI's or generators)could be a uscful procedure for The average hydrogen flow rate to the VLT over the long detecting hydrogen leaks in some cases. This would be term is probably less than about 0.1 scf per minute (sefm) supplemented by walkdowns of the systems with hydro-for plants without a recombiner in the waste gas system gen detectors. Another approach is the use of permanent and is about I scfm for plants with recombiners. larger hydrogen detectors with alarms.
short-term maximum flow rates occur during manual ad-justrnents when hydrogen is added to the VLT, while For the VCI', the very low average consumption rates for maximum local flow rates (corresponding to a postulated some plants are such that a limited amount of hydrogen loss of letdown flow from the reactor coolant system) may could supply the consumption needs for an appreciable be over 20 scfm for some plants. During startup from a time.The low short. term maximum flow rates also indi-cold shutdown,0ver 700 scf of hydrogen may be needed to cate that small restricting orifices or excess flow valves purge the VLT and obtain the desired hydrogen condi-(e.g., excess flow check valves or valves actuated by signals tions in the VCf and reactor coolant, from flow elements) with low setpoints might be used to limit the hydrogen release rate so that the average con-The hydrogen used to cool the electrie generator is circu.
centrations in local areas with normal ventilation are kept lated through integral water-cooled heat exchangers by well below the lower flammability limit of 4 volume per-fans located near the ends of the rotor. Shaft oil seals cent. A value of 2 volume percent is recommended in prevent the entry of air and the escape of hydrogen. A SRP Section 9.5.1, Revision 3 (NUlWG-0800).
carbon dimide syrtem is provided to purge hydrogen from the generator before maintenance and air from the gen-I or generators with normal leakage, the daily consump-erator before adding hydrogen after maintenance. The lion of hydrogen is probably less than a few percent of the average hydrogen consumption rate,which increases with hydrogen stored in the generator. Ilence, the slow reduc-generator size, is probably less than about 0.4 scfm for tion in generator pressure with normal hydrogen con-l, systems with small leakage, larger short-term flow rates sumption would permit isolation of the hydrogen supply will occur when hydrogen is added to a generator that is execpt for periodic (e.g., once-a-shift) adjustments to normally isolated from the supply. The volume occupied keep within required hydrogen conditions for the genera-by the hydrogen in generators anges from about 3500 to tor load. An alternative approach is to provide a small 6000 cubic feet and is maintained at pressures of about 85 flow limiting device with a bypass section for use on the 7
NURIE1364
-_m infrequent occasions when higher flow rates are needed volumes (e.g., NURl!G-1370, NUREG/CR-2475, and i
(e.g., after generator maintenance).
NURl!G/CR-5275). NURIiG/CR-5275 describes tests on the effects of hydrogen concentration, obstacles, and transverse ventin wcreconductedin'g on llame acceleration and DDT that 3.2 1ly(Irogen L.,onibilSi10:1 the l l.Ahll! facility, a half-scale model Characteristics of the upper plenum volume of a PwR ice condenser containment. These tests showed that flame acceleration immediate ignition of the hydrogen released from a leak at 12 volume percent was negligible DDT was first ob-or break results in a diffusion flame that could damage served at 15 volume percent, with no transverse venting nearby equipment because of thermal effects. Any delay but with obstacles present. Ilowever, with no obstacles in ignition could result in the buildup of hydrogen-air and a large degree of transverse venting, DIJr did not mixtures, which could then ignite and lead to either occur at 28 volurne percent. NURI!G/CR-5275 provides deflagrations or detonations. Dellagrations are premixed a methodology for a qualitatisc classification of the po-Ilames that advance into the adjacent unburned mixture tential for DDT, given an estimate of the local hydrogen at subsonic speeds. If the deflagration speed is much less concentration and knowledge of kical compartment than sonic speed, thermal damage to equipment with geometry. Local detonations in large, dry PWR con-little mechanical damage could occur. Ilowever, the tainments are discussed in a memotandum from a
flame can be accelerated (e.g., by turbulence produced by li. S. Heckjord dated Alarch 24.1992: NURl!G-1370;and flow over obstacles in the flame path). As noted in NURI!G/CR-5662.
NURI!G/CR-5275, if the deflagration speed increases to i
over 100 m/see, shock wases are generated if the Predictions of the consequences of postulated releases deflagration speed is large enough, a deflagration to-from hydrogen distribution lines are strongly dependent detonation transition (DDT) coutd occur. Ilydrogen deto-on plant-specific compartment geometries and equip-nations, which are premixed flarnes advancing into the ment locations and involve large uncertainties in predie-unburned mixture at supersonic speeds, can cause ther-tions of such factors as local hydrogen concentrations, mal damage and extensive mechanical damage because of igmtion locations, DDT, and subsequent blast damage, the associated overpressures and impulses (NURl!G/
I'or example, inadequate mixing and stratification may CR-5275).
result in local detonable mixtures even though average hydrogen concentrations are below the detonation limit.
As noted in Technical Note 690 (National Bureau of in view of these uncertainties, the alternatives considered l
Standards, 1976), the wide flammability limits for for reducing risk primarily insolve ways to prevent signifi-l hydrogen air mixtures (about 4 to 74 volume percent for cant hydrogen releases.
upward flame propagation and 9 to 73 volume percent for downward propagation)and low ignition energies suggest 3.3 1)esign Feattires at Operatiiig a high probability of ignition of flammable matures from random ignition sources. Ignition sources such as small l*lantS USed,I,O Reduce Risk lirom sparks and flames can initiate deflagrations, whereas di-1Iydrogen Systems for ihe Volunie rect initiation of a detonation in a detonabic mixture Control Tank and Generator usually requires much stronger (shock wave) ignition sources such as a high energy spark or explosive information from final safety analysis reports, supple.
(NURl!G/CR-2475). The required energy source for di.
mented by information from plant visits and surveys, indi-rect initiation of a detonation is relatively low for a cated significant plant-to-plant differences in existing hy.
stoichiometric mixture (about 30 volume percent hydro-drogen system design features to reduce the risk from gen in air) but increases rapidly as the concentration ap-hydrogen line leaks or breaks. Some plants havc onc large proaches the detonation limits, Since the presence of storage facility to supply hydrogen for both the VCT and high energy sourecs needed to initiate detonations near the generator but do not provide excess-flow protection the lean limit is unlikely in the auxiliary building or at (NURPG/CR-3551). Ilence, in the event of a system distribution levels of the turbine building, deflagration failure, a potential exists for discharging large amounts of i
would be the most likely initial rnode of combustion for hydrogen into the auxiliary building or turbine building this range of hydrogen concentrations Depending on before the storage facility can be isolated manually. In this
- such factors as hical hydrogen concentrations and com.
case, the maximum release rate probably is controlled by partment geometries, DDTeould then be the mechanism the flow restriction provided by the pressure regulators for initiating local detonation.
operating in parallel at the storage facility. The normal ventilation air flow rates for the buildmg as a whole are i
The lean detonation limit for hydrogen-air mixtures at large. Ilowever, the hydrogen retcase rates as a result of
(
standard conditions is 18 volume percent for laboratory these leaks or breaks could be too large for the dilution i
scale tests (National Bureau of Standards,1976). Ilow-capability of normal ventilation in the local building areas
)
ever, tests and analysis mdicate lower limits for larger involved and could result m estimated local average l
i NURiiG-1364 8
- L. -
_ ~. -
- _ _ - - _ -. - -. - _ - _ - -. _ ~
i t
i hydrogen concentrations that are well into the detonable of a pressure regulator oroperate with the VCI'normally range, parte.ularly if allowance is made for nonunifonn isolated from the hydrogen supply, except for periodic i
mixing, l'or such releases, damage to compartment walls (e.g., daily) manual adjustments of hydrogen condiuons, and blast wave and flame front propagation via ventila.
At one plant a single 200-scf hydropen botlle is connected tion ducts, fire barriers, and passageways could result m to the VCT, but is isolated except for daily adjustment of loss of safety-related equipment.
conditions in the VCI', At a number of plants, small flow-limiting devices (maximum flow rates ranging from Undetected leaks might also result in the buildup of unac-about 8 to 25 scfm) are used to limit flow to the VC1',
ceptable amounts of trapped hydrogen in overhead re-These maximum flow rates were usually low enough so gions with an inverted. pocket geometry and inadequate that normal ventilation in the compartment with the leak local ventilation because of low ventilation rates or inade-or break would result in an average compartment hydro-quat e intet and exhaust locations. Ilydrogen has a density gen concentration below the lower flammability limit.
on:j about 7 percent of air density and has a large diffu-One plant combines a small storage facility for the VCI' sion coef ficient. E lence, small leaks at lower elevations of with a small restricting orifice.The configurations to limit a compartment might be expected to mix quickly with the flow rate include excess flow check valves, restricting air and be removed with cormal room ventilation. Ilow-orifices, and valves controlled by a flow element in the eser, hydrogen from moderate releases could stratify and VCI'line. The valves controlled by the flow element in collect in overhead areas. In addition, much of the hydro-the VCI'line fait closed and are also closed on an auxiliary gen distribution piping is located in overhead areas.
building isolation signal when normal ventilation air flow to the building is lost. At some plants secondary pressure As a result of a review of more than 400 industrial acci-control stations are located outside the turbine and auxil-y dents (nuclear and non-nuclear) involving hydrogen,1 ac-iary buildings. Since normal leakage results in small rates tory Mutual itesearch Corporation recommended in its ordecrease in generator pressure,a number of plants aiso report that hydrogen safety standards emphasize the im-operate with periodic manual makeup to the generator, portance of hydrogen monitoring forleak detection.The Other features include seismic piping supports, scismic basic approach of the National Aeronautics and Space Category I piping, or sleeved piping for portions of the Administration (NASA) to reducing risk from hydrogen distribution system with the guard space between the fires or explosions includes (1) prevention of leaks, pipes ventedilirectly outside the building. At a few plants, (2) monitoring to detect leaks quickly and take corrective the hydrogen lines had been relocated after start of com-actions. (3) prevention of accumulations of leaked hydro-mercial operation to minimi/c the amount of piping in the gen by plen uful ven tilation, and (4) climination of ignition buildings.
sources, but assuming that ignition sources are present (NASA.1968 and 1992, and the hydrogen safety stan-Color coding is useful du ring walkdowns for leak tests and dard' to be published by NAS A).
for avoiding inadvertent damage to hydrogen piping. In-formation on color coding of hydrogen piping was ob-At typical gaseous hydrogen storage facihties, a pressure tained during visits by Idaho National lingineering I abo-control station with two pressure regulators operating in mtory (INiii.) personnel to 13 plants in 1989 and from a parallel is used to decrease the hydrogen supply pressure staff survey. Of the 36 plants covered 20 use color coding (typical maximum pressures from 1500 to 2400 psig) to for the VCI'and generator lines,2 use color coding only about 100 psig in the distobution lines to the auxihary and for the generator lines, I uses color coding only for a turbine buddings. Overpressure protection for these lines hydrogen water chemistry (llWC) installation, and 13 do is provided by relief valves discharging outside. Pressure not use color coding.
regulators are then used to control pressures.at about 20 psig at the VC1' and 70 psig at the generator. At some The amount of hydrogen in the hydrogen supply piping is plants with a single hydrogen stmage facility for one or very small and can be neglected (e.g.,100 feet of 1-inch-two units, excess flow valves are provided downstream of diameter supply piping contains only a few standard cubic l
the pressure control station at the storage facility. Ilow-feet of hydrogen), flowever, a few hundred standard co-ever, the excess flow valve has a high setpoint (e.g., shut-bie feet could be released from the VCI' as it depres-i off flow rates from 80 to 330 scfm) that was probably surires, and about 15,000 to 25,000 scf could be released selected to meet the flow mte requirement for the infre-from generators. The final pressure regulators may limit quent purging and charging operation for the electric backfh)w to breaks in the hydrogen supply line from these generator, Ilence, breaks or leaks in the distribution lines sources, but could be backed up by check valves.The use for the VCl'or the generator with flow rates up to these of excess flow check valves to stop flow from the hydrogen shutoff values would not be stopped by this excess-flow storage facility, supplemented by check valves in the lines protection. Other plants use separate, small storage fa-near the VC.I' and the generator to stop the backflow of cillties that are nonnally connected to the VCl' by means hydrogen in these components to leaks or breaks'in the hydrogen supply lines in the auxiliary and turbine
- NSSl WINatl.
buihlings, was proposed in a recent individual plant 9
N UltliG-1364
examination (lPE) for a PWit with the hydrogen control tures obtained from a walkdown of the areas containing station on the level below the generator (Yankee Atomic the hydrogen piping and components.
lilectric Company,1989).'lhis approach protects the sup-ply line, reduces the contribution of the storage facility to Methods for leak detection that have been used at plants releases, and limits potential iocations for large releases include monitoring for abnormal consumption of hydro-to the generator itself (e.g., shaft oil seals) and to the gen using integrating flow meters or monitoring system sample, drain, and other lines below the generator.
pressure changes with time (for small or isolated sup-plics), and use of permanent and/or portable hydrogen detectors. If normally isolated or limited hydrogen sup-In surnmary, alternative design features for limiting the plies are used, the rate of decrease in pressure could be consequences of larger breaks or leaks m the hydrogen appreciable for relatively small leaks and the total supply lines for the VCI and generator that have been amount that could be leaked without early corrective used or considered f or use at operating plants and glo not action would be limited. For normally isolated hydrogen require major system changes mvolve several basic ap-supplies, existing pressure-sensing instrumentation and pmachs low-pressure or other alarms could be ustd for the VCr and the generator. For the limited supplies, the final 1.imit Supply Amount-use a limited supply nor-pressure regulator to the VCr would maintain VCr pres-e mally connected to the VCr to restrict the total sure until the supply was nearly exhausted. However, amount that could be released during a single event existing instrumentation and alarms could be used for j
monitoring the supply pressure. For large supplies, auto-Limit i Iow Itate From Supply-limit the maximum matic detection such as the use of permanent hydrogen e
flow rate from the storage facility to a line break so detectors with alarms to the control room may be needed that normal veritilation would keep local average hy-to identify significant leaks in time for corrective action.
drogen concentrations below the lower flammability This type of detection is covered in the IIPIRI guidelines limit (EPIll,1987) and the N ASA approach (N AS A,1992, and E"
Isolate Storage Facility-limit the available time e
when discharge from the storage facility could occur by using normally isolated storage facilities 3.4 Alternatives for Volume Control I imit Hackflow of Contained Hydrogen-use check Tank and Generator Ilydrogen e
valves or restricting orifices if needed to limit back-1)istrilMition Systems and flow of hydrogen to breaks in the supply lines from Ilydrogen Storage Facilities the pencrator and possibly the VCr The features that have been used or proposed for use in
.Ihese features would reduce the frequency of occurrence operating plants were considered in selecting alternatives and/or the consequences of events involymg larger breaks for the VCr and generator hydrogen systems. For the or leaks m the supply line that result in hydrogen release hydrogen storage facilities, the rek) cation of the storage rates so large that there is msufficient time for corrective facility to increase the separation distances was evaluated manual actions. The features also tend to reduce the for cases involving safety-related air intakes and struc-significance of buildup of trapped hydrogen because of tures. The addition of a blast shield was selected as a the low limits on hydrogen release rates or total amounts means of reducing the consequences of a detonation at of hydrogen available for release during an event and the the facility. Automatic closure of air intakes was selected additional time available for corrective actions. In the as a means of preventing the ingestion of a flammable auxiliary building, trapped hydrogen would probably be mixture at the intake following rupture of a hydrogen limited to the immediate areas containing the hydrogen header at the storage facility.
system because of the additional ddution by ventilation m adjoining compartments.
The following alternatives were evaluated:
Even with low leak rates, significant quantities of hydro-Alternative 1-Take No Action gen are involved. For example, a leak rate of 10 scfm probably results in an average concentration in rooms Under this alternative there will be no new regulatory with normal ventilation that is well below the lower flam-requirements. Consistent with existing regulations, this mability limit for hydrogen-air mixtures, but involves the alternative does not preclude a licensee, or an applicant release of nearly 5000 scf per shift. Hence, anyjudgments for an operating license, from proposing to the NitC staff on the significance of trapped hydrogen at a particular design changes intended to reduce the risk associated plant must be based on plant-specific information such as with the hydrogen storage and distribution systems on a local confirmed ventilation rates and on construction fea-plant-specific basis.
NUllEG-1364 10
L Alternative 2-Install inw Sctpoint Incess flow Valves, 111ternative 4-Enclos e Safety-Related Equipment lacated in Restricting Orifices, or Check l'alves in flydrogen Supply the Turbine fluilding in filast-and Fire l'roofStructures Line to the Generator This alternative provides protection against the conse-This alternative provides protection against leaks or quences of leaks or breaks in the supply line or at loca-breaks in the hydrogen supply line from the point of entry tions near the generator.
into the turbme building to the hydrogen control station below the generator, including any branches from this l
line to other buildings. A low setpoint excess flow valve or
,titernative 5-Install Low Scipnint ihress How I'alve or restricting orifice is provided in the supply line outside Ihe Restricting Oryice in flydrogen Distribution System to VCT budding to limit flow from the storage facility to the leak and Provide llydrogen Detectors, (/Nceded or break. A check valve or restricting orifice could be provided in the supply line near the generator,if needed This alternative entails the use of a low setpoint excess to prevent backflow from the generator to the leak or flow valve or restricting orifice (e.g., sized for 150 percent break. The alternative also limits flow from the storage of maximum daily flow rate)in the low-pressure hydrogen i
l facility to leaks or breaks in the piping near the generator supply line outside the auxiliary building to limit the rate and at the generator itself. Preoperational testing and of hydrogen flow from the storage facility to the leak or periodic retesting of the execss flow valve to ensure that it break.1he limit should be low enough so that normal
[.
will function properly and administrative controls or de-ventilation in the compartment with the leak or break sign features and training to prevent inshrtent opera-would keep the average compartment hydrogen concen-tion with the flow. limiting device bypassed are part of this tration well below the lower flammability limit.This alter-alternative. I'trger leaks near the ge ierator would be native includes (1) preoperational testing and periodic indicated by the reduction in generator pressure because retesting of the excess flow valve to ensure operability of the limit on makeup flow from the storage facility.
and (2) administrative controls and features to prevent fluildup of significant amounts of trapped hydrogen from inadvertent opening of any bypass around the excess flow smaller leaks in turbine buildings may not be a concern valve or restricting orifice and to isolate the supply manu-for most PWR plants because of the open construction at ally if normal building ventilation is lost. In addition, each level and between levels (e.g., open hatches, floor protection against leaks at flow rates up to the setpoint of grating, and stairs). Ilowever, administrative controls the excess flow valves, such as the use of hydrogen leak should be provided for monitoring hydrogen consumption detectors with alarms to the control room or other suit-for indication of leaks and for corrective actions.
able measures, should be provided for areas where leak-ing hydrogen could be trapped in unacceptable quanti-ties. Trapped hydrogen may not have a significant effect Alternatirc J-l'roride Manual Makeup of Ilydrogen to on risk for most plants because of such factors as location Generator and Check Valre or Restricting Oryice at the of vital equipment and adequacy of ventilation at the low Generator release rates. I lence, two options were considered for this alternative. Alternative 5 costs do not include the costs for hydrogen detectors. Alternative SA costs include the This alternative provides protection against leaks or additional costs of permanent hydrogen detectors.
breaks in the hydrogen supply line from the point of entry into the turbine building to the hydrogen control station below the generator, including any branches from this line to other buildings. It entails operation with the hy-Alternatire 6-1.imit the Guantity of flydrogen Normally Connected to the VCT drogen facility normally isolated from the generator by an isolation valve outside the building and periodic manual adjustments of hydrogen conditions in the generator. A This alternative entails a limit on the total amount of check valve would be provided in the supply line near the hydrogen in the storage facility that is normally connected generatorif needed to prevent backflow from the genera-to the VCT. The limit is such that the release and subsc-ter to leaks or breaks in the supply line. The Wiernative quent firc or detonation of the hydrogen in the supplyand also prevents flow from the storage facility o leaks or the VCT would not cause unacceptable damage to safety-breaks at or near the generator when the generator is related systems in the auxiliary building. Administrative isolated. Reduction of generator pressure with hydrogen controls. design features, and training should be provided release would provide warning of larger leaks or breaks, to prevent inadvertent operation with a larger amount of while monitoring hydrogen consumption by monitoring stored hydrogen. Administrative controls should also be changes in generator pressure or frequency of manual provided to monitor hydrogen consumption forindication makeup would indicate the presence of smaller leaks.
of leaks and to prevent the buildup of unacceptable Administrative controls would be used to prevent inad-amounts of hydrogen in areas where hydrogen could be vertent operation with the hydrogen supply connected.
trapped.
l 11 NUREG-1364 I
1
<fiternative 7-l*rovide Nonnally twlated Supply IVill: 1)aily prevent the buildup of unacceptable amounts of hydro-3/anual 3/aAcup of tlydrogen to Ilic ICT pen in areas where hydrogen could be trapped.
- llternatire S-Relocate flydrogen Storage Facility To 3/cet
.Ihis option entails isolatson of the hydrogen storage Separation I)istance From Safety Related Structures Jacility from the VUI except for brief daily operations to adjust VCF conditions. Administrative controls, design
<llternatire' 9-lintall lilast l><' lection Silic/d at flydrogrn f
features, and training should be provided to prevent Storage Facility inadvertent operation without isolating Ihe hydrogen sup-
~
ply. Administrative controls should be provided to moni-
,titeinatire 10-Install flydrogen Analy:cr-<lcluated Air tor hydrogen consumption for indication of leaks and to intaAc I.ourres at Safety-Rclared Air IntaAes 1
1 N URiiG-1364 12
4 TECllNICAL FINDINGS l
4.1 IIllrO(lllClioll hydrogen for coohng the cicetric generator was small and
)
did not warrant a more detailed study. This was based on l
P N regulatory analysis is based in part on work per-the small risk resulting from the absence of safety related ed by Idaho National lingineering 1.aboratory equipment in the turbine building at the six IlWR plants I
.) under a technical assistance contract and re-reviewed in the study. The screening analyses in liGG-l I in INEl, 1991: INI 1,1992; and NUREG/
NTA-90S2 and NUREG/CR-5759 also indicated that I
(
a759. The work included the following:
the risk associated with the off-gas system in HWRs, the waste gas system in PWRs, and the station battery rooms Surveys of HWR and PWR plant information perti-and portable bottles of combustible gases in both llWRs e
nent to GSI-10h that included the following sources:
and PWRs was very small and did not need to be consid-cred further in this study.
linal safety analysis reports (FSARs)
The hydrogen storage facilities at PWR and HWR plants probabilistic risk analyses (PR As) were initially sercened by comparing the actual separa-tion distances between stompe facilities and safety-plant informanon i. nun site vistts related structures and air intakes with the separation infonnation from equipment manufacturers distances provided m the Electric Power Research Insti-tute (EPRI) guidelines (EPRI,1987). The !!PRI separa-A resiew of observed combustible gas esents (pre-tion distances for safety-related structures are the dis-e cursors) at U.S. nuclcar power plants.
tances needed to pr even't unacceptah!c st ructural damage A nsk evaluation of hydrogen storage facilities at fnun postulated hydrogen detonations at the storage fa-e PWRs and HWRs with respect to damage to safety-cihty. lhe EPRI separation distances for safety-related related buddings caused by explosions at the storage air intakes are the distances needed to prevent the inges-locanon and hne breaks at the storage location re.
tion of a flammable hydrogen-air mixture at the intake sulung in flammable hydrogen-air mixtures at following a postulated pipe break at the facility and re-safety 4 elated air mtakes.
lease of a jet of unburned hydrogen. Although a number of plants did not pass this screening test, this does not A detaded PR A by INIf.1,0f the risk associated with directly imply high risk because credit for plant-specific e
the use of hydrogen for the volume control tank mitigating features could result in a low hazard. These (VL l > and main generator during full-power opera-mitipting features included (1) use of reinforced-tion for a representative (base case) PWR plant. A concrete storage buildings with one open siJe facing away f our-loop % esunphouse plant was selected because from safety-related structures and air intakes, (2) mterl it was representative of a signdicant portion of the vening non-safety-related structures. and (3) thfferences PWR popuhttion, had a I,esel 3 PI completed and detaded fire analysis.j A, and had a in elevation of storage area and air intakes. An informal I he IN El. base survey showed that some of the plants that were screened case PR A was then supplemented by a generic were not confieured to allow credit for such mitigating analysn based on senyitivity stuches of the variation features. lieneb, more detaded, plant-specific risk esti-m plant nsk resolungIrom observed changes in loca-wes were also deemed necessary.
tions 01 sal,ety-related equipment and plant feed-and-bleed W&H) capabihties. A similar approach As a result of the above evaluations, the more detailed was followed in a scoping PR A for BWRs of the risk risk and cost-benefit analyses under the scope of GSI-106 associated with hydrogen distribution to the main were rcJuced to generator. A IlWR/4 wah a Mark I contamment was selected for the BWR hase case plant.
hydrogen storage facdities for PWRs e
hydrogen distribution system for the VCT in PWRs Screening analyses for the risk associated with port-e e
able bottles of cornbustible gas and with other hdrogen distribution system for the electoc genera-e sources of hydrogen, including the (I) waste gas sys-tors in PWRs tem at PWRs, (2) off-gas system at HWRs, and (3) battery rooms at PWRs and HWRs.
4.2 Ren,ew Of U.S. Nuclear power Cost-benefit analyses to determine possible plant e
modifications to rbduce risk that could be cost effec _
l lant Ilydrogell I,recurSOr$.
INI L conducted a i terature search to identify hydrogen The scopmg PR A for llWRs in EGG-NTA4082 (IN El.,
events at U.S. nuclear power plants through Apnl 1990.
1991) mdicated that the risk associated with the use of Table I gises the number of hydrogen events for the 13 N UR EG-1364
hydrogen storape facdities and the hydrogen distnhution Volume Control Tank System lhents systems for the VCI'and the electricgenerators A hydro-pien esent is identified as an unburned hydrogen leak a A total of 11 events involved unburned hydrogen leaks fire. or an explosion. None of the events in Table 1 m.
Inno the WI' cowrtm systern of PWih [ including the vohed dam.me to safetv-related equipment. Up to the supply piping m the primary auxiliary buildmg (PAH){
date of this literature search (Apnl 1990). IN!!L esto None of these events resulted m either a fire or an explo-sion. Ten leaks were detected when associated gaseous mated that the total number of reactor operating years mcludmg shutdown time [ reactor 9 ears (l(Ys)l was 917 radioactivity was detected. Ten of the leaks were hard-ware related; one was caused because a sampling procc-l(Ys for PWin and 14241(Ys for PWlh and llWits.
duw was inadequate. Of the hardware-related events, Shutdown ome was included because sorne events oc one wm caused by a leaking diaphragm in the hydrogen curred dunnp this tune.
supply regulator, seven were valve related, and two were caused by a leaking V( T vent header.
Electrical Genciator lhents Storage Faed. it) ISents l'ourteen PWR events and two llWR events involved Threc events occoned at hydrogen storage facilities.Two by dnmen associated with generator cooling (includmg the of Ihose occurred at PWIb and one ocemred at a llWit.
supply piping within the imbmetenerator bm! ding). Fw
()ne of the three events involved an explosion. Although penence from HWRs was included because the eqmp-the other two events involved fires. it is aho possible that ment and operauons for generator cooling systems are there was an initial explosion with a sub. sequent fire. No similar f or both types of plants. Of these 16 events, 7 major unbur ned hydmpen releases were reported. None mvohed unhmned leaks 7 involved lires. and 2 involved of the three events resulted in damage to any safety-explosions. Of the nme fires or explosions (all of which related equipment or structures.
occurred in PWRst six occurred while the plants were at more than 90-percent power and led to a turbine and/or Recent Events at U.S. Nuclear Power Plants reactor inp.
Several hydrogen events have occurred in the turbine buildings at U.S. nudem plants since the cutoff date of IN1!!.attnlmted one of these nine events to leaks at the April 1990 that was used for this resiew to generate the hydrogen thstnhuuon system levels in the turbine build-imtiating event frequencies. A failure of a main trans-mg. The rest were associated with leaks near or at the funner at the N1aine Yankee plant in April 1991 resulted generator. Niost of the generator leaks were at the shaft in arcing below the generator that eaused a hydrogen Icak seal and wer e caused by such diverse mechamsms as fa ults and hre (Niaine Yankee Atomic Power Company,1991).
m the electrical power to the seal oil system, a clogged A tu bine overspeed incident at the Salem plant in No-s eal oil hher, loss of hydrogen cooling because of a loose vember 1991 resulted in damage to the low-pressure tur-temperature sensor, and failure of an oil pressure sensing bine and a hydrogen fire (Public Service lilectrie and Gas h n e.
( ompany,1991). In October 1991, at Nine Niile Point Unit 2 (Niagara N1ohawk Power Corporation,1991).
about 10,000 set of hydrogen was released from a broken Sescral bres m turbine hmiglines at forcien plants haye g
g g
Id dud been large and damagmg. t he event at \\ andellos I ui there was no fire or explosion. In December 1991, a Spain in October 19S9 (mernorandum irom R. L Pres 15-minute release of hydrocen from the seals of the jen-said dated September 4.1991) was caused by failure of a he Palisades plant occurred because of bleekage e o Jugh pressure turbine and involved oil and hydrogen hres.
M d od in a oil Ulter. There was no fire or explosion I he od and hydrogen fire at the N1aanshan plant m Ian (Consumers Power Compar,y.1991).
wan in 14X5 (reported,m Nuc/comcsIVerk on August 8and 22,1985; December 19,19S5; and February 6,1986) ap-parently was caused by loss of blading in the low-pressure 13 Metht>d of Analysis
~
turbine. I ailure of lubneating oil lines because of vibra-tion apparently caused the oil fire in the turbine building 13J Geiteral at the plant m N1uchleberg, Swit/criand, m 1971 (NRC Translation 2240). A more recent event at Chernobyl in The INEL analysis in NURl!G/CR-5759 of the risk detober 1991 was a large hydrogen fire that apparently associated with the hydrogen facilities and distribution was caused by an electrical fault (reported in Nuc/ conics systems for the VCT and electric generator for the base IVcck on October 17, 1991). INiii considered the avail-case PWR was based on a " vital area" analysis. The plant able information on hydrogen systems and events at for-r esponse, modeled by ty pical probabdistic risk assessment cign plants to be insufheient to rnake a meaningful evala-event trees and fault trees, was used to identify vital areas at on of event frequency for domestte application.
where hydrogen fires or explosions could result in damage NU Rl!G-1364 14
to safetya clated equipment with a significant condiuenal T/DI1R transierits can range from simple turbine trips to probabihty for core damage.The vital area analy sis for the steamline breaks. These transients involve loss of all base case phmt included the following:
forms of core cooling anJ 1oss of RCS coolant inventory at high pressure from the pt essuri/cr power-operated relief idenufication of the applicable accident sequences valves (PORVs) or safety valves. T/l OCA transients in-e volve failure of RCS rnakeup or recirculation following a determination of affected safety equipment reactor coolant pump (RCP) seal I.OC A caused by inade-e quate RCP seal cooling or a 1 OCA caused by stuck-open development of new accident sequences l.or which e
PORVs or safety valves. This category includes events hydropen hres or explosions;ue the mittating events sua a etion bidout ad le of componer,t cooling and tlut mclude the eff ects of postulated losses of
- water, safety equipment as a result of these fires or explo-
'i""N The mitating events considered were seismically induced I he detailed analysis by INiii. is limited to events that or random failures of hydrogen piping or components at occur when the rcactor is at full powcr. In NURl!O/
the storape facihty or in the by drogen distribution systems CR-5759, INI'l. concluded that power operation th rough in the audary and tu rbine buddings. In the tin hine build-hot standbv (Modes 1 throuch 3) should be bounded bv mg, events are considered to occur at the generator and these full power results. Th'is conclusion was based or}
knn fnun the hydrogen control station to the generator f
considerahon of system operability requirements, decay or at the hydropen supply Une to the control station.
heat loads reactor coolant system (RCS) coolant inven-tory, RCS coolant makeup capacity, containment integ.
4.3.2 Seismic Events nty, and operahihty ol vital auxih.uy systems. IN1!I. also
~
SacrM quWes suppo&d by the NRC (sce NURiiG/
concluded that the effects of hydrogen-induced events during hot shutdown (Wde 4), cold shutdown (Mode 5),
CR-5759, p. 51) hase shown that small-diameter piping and refueling (Wde 6) would be insignificant. This con, symms such as those used in the hydrogen systems have a very large seismic capacity (i.e.. failure unhkely for low-g clusion was based on consideration of the typical absence of hydrogen from the in-plant distnbution system, the low quthquakc4 Ahqugh uns condusion applies to the decay heat loads, inct cased coolant inventories, and low er piping runs, connecuons to equipment or connectmp pipe coodnt temperatures.
nay be sulneraNe to seismic fadure, particularly if there is inadequate anchoring or flexibility. At the base case plant, key portions of the hydrogen lines in the auxiliary
.Fhe equatians f.or the point esurnates of the core damage building are seismically quahhed. Ilence, no damage is f requency (UDF) lor the genene analyses have the pen-expected for the higher frequency, relatively low-level eral form of an milialmp event I,requency for a large carthquakes. I:or a severe (beyond-design-basis) carth-hydrogen release, /F, muhiplied by the probabihty that quake (e g.,0.Xg). INEl. concluded that the contribution t he release and resulting fit e or explosion damages sal,et} ~
of scismically induced hydropen fires or explosions should related equipment. licquipment damare), muhiplied by be relauvely small compared with the dacci damage the condiuonal probabihty that the loss of the equipment caused by the earthquake itself l'or example, a severe lor this scenano results in core damare,l'(core damage).
carthquake could initiate a loss-of-offsite-power event t hat is, because of fadure of ceramic insulators at the offsite power transformer, l.oss of pnmary coolmg equiprnent Cl)l'
//'/Nequipment damape)*/1 core damage)
(needed for the bleed-and-feed f u nction)could also occur because of failure of either the refueling water storage nnk (RWSTh PORVs, or piping (including failure of Imge quanuties of hydroren are not present near the reactor sessel and contr'ol roJ drive motorsprimary h)dnTen piping). The RWST, which has a lower seismic coolant ssstem pipmp, and steam penerators. llence, capacity than the hydrogen piping, should fail before the hydrogen piping is darnaged. llence, INEl. concluded pipe-break loss-of coolant accidents fl.OCAs), steam pencrator tube r uptures and anticipated-transient-that the core damage contribution during seismic events
~
awneiatcd with the seismically quahfied hydrogen piping without-scram (AIWS) events should not result from hsdrocen release esents. I he INiii. calculations were system in the auxiliary building at the base case plant liiniteI.I to the following two types of transient induced should be small relative to that already considered in the PR A and could be neglected.
core melt scenarios:
INiii. assumed that the frapdity of the hydrogen gas sys-(1) transients with failure of decay heat removal (DHR) tem for the main penerr. tor at the base case plant could be sy stems (T/D H R) represented by that for the dry small-bore pipe with threaded joints in the fire protection system at that plant.
(2) transient-induced i UCAs ( Ill ()CAs)
INEl. obtained an initiatmg event frequency of 15 N U R EO -1304
_A
s 4
$.31I-5/RY for seismically induced breaks of this piping age might occur to safety-related equipment in compart-using the Lawrence Livermore National laboratory ments with hydrogen lines or in adjoining compartments.
(I LNL) seismic ha/ard curves, For the generic study, INf!L assumed that this seismic fragility was representa.
Iarge amounts of hydrogen are also available in storage tive of the hydrogen storage facility and the distribution facilities to initiate or exacerbate a fire in the turbine system piping and components containing hydrogen for building at some plants if special features to reduce risk the auxiliary and turbine buildings.
(e.g., excess flow check valves or normally isolated sup-plies)are not provided. Anotherlarge source of hydrogen 43.3 Non-Seismic Events is that contained in the generator. A hydrogen fire in the turbine buildings might not normally be considered to For non seismic events, INliL estimated the initiating have the potential to directly affect nearby safety-related event frequencies for the VCI'and main generator hydro-equipment because of the large si/c and open design gen distribution systems and storage facilitics from the typical of these buildings Ilowever, delayed ignition of reported numbers of pertinent hydrogen events and the hydrogen escaping from a large break in the hydrogen associated number of reactor-years of oper ation lines or generator could result in the accumulation and (NURl!G/ cit-5759, Appendix II). 'I he mear initiating cventual ignition of large quantities of hydrogen and re-event frequency of a large hydrogen rCease involving the sult in the propagation of blast waves and fire fronts that VCT hydrogen system in the auxihm building is could cause direct damage to nearby safety-related equip.
5.51!-4/RY. In the turbine buikling, the rr c an frequency mentflhese blast waves or fire fronts could also cause oil of a large hydrogen release is 611-3/RY f ar the generator leaks and fires, which could then spread and dmnage level and Ll E-3/R Y for the hydrogen 6stribution level.
safety-related equipment in lower levels of the building.
4.3.4 Damage Mechanisms large ou fires in the turbine building have always been considered because of the presence of large quanuues of In the evaluation of hydrogen storage facihties. INiiL turbine lubricating oil in the lower levels of the turbine assumed that the loss of safety related equipment was building. An event involving a break in the oil lines could caused by detonations with resultant blast damag : to result in an oil fire caused by burning hydrogen or ignition nearby safety-related buildings or by large releases of of the oil from prolonged contact with hot metal surfaces unburned hydrogen that resulted in flammable hydrogen-or insulation. llurning oil from the operating level could air mixtures at nearby safety-related air intakes. INEL then cascade to lower levels and spread to other fire assumed loss of equipment if the separation distances zones. Early shutoff of the pumped oil could help prevent between the storage facility and the buildings or air in-a large fire.1lowever, the oil supply system is designed to takes did not meet the EPRI guidelines and there were no be highly reliable because loss of the bearing supply could mitigating factors.
also cause extensive damage to the turbines and genera-torin the absence of a fire. Design features to increase the The EPRI calculations were made using the TNT equiva-reliability of the oil supply include turbine shaft-driven lency method, which equates the available amount of pumps and emergency pumps. After a turbine trip, oil hydrogen to an amount of TNT givmg the equivalent may be supplied to the break and fire for up to an hour as damage. The blast wave parameters are then obtained the turbine coasts down. 'ihis time can be reduced by from relations for TNT detonations. For these postulated breaking the condenser vacuum to slow the turbine and l
detonations in op;n air, liPRI assumed that the TNT-stopping motor-driven pumps.
I hydrogen equivalence k 20 percent on an energy basis (520 percent on a mass basis). On this basis,1000 scf of 4.3.5 Ilydrogen Explosion Methodology hydrogen is equivalent to 27.1 pounds of TNT This value I
of 20 percent was considered conservative for outside The conditional probability of damage to safety-related detonations.
equipment P(equipment damage), following a large hydrogen release was treated as the product of (1) the For local detonations in partially or fully enclosed vol-probability, P(delay), that early ignition following a large umes such as compartments and passageways in auxiliary hydrogen release wouki mt occur and (2) the probability, buildings, reflected shock waves from other surfaces P(blast), that, given the continued accumulation and within these volumes could cause additional damage to eventualignition of the released hydrogen, the resultant the target surface Other factors to consider include the blast wave or fire front would incapacitate safety-related lower expected strength of interior walls and the weak-equipment.
ness of fire barriers and air ducts that woukt not be de-signed to withstand differential pressures occurring dur-In the NUREG/CR-5759 generic calculations, INiiL ing a detonation. Significant hydrogen releases could assumed that P(delay) was equal to 0.01 for large hydro-occur at some plants with large supplies and no special gen releases at the main generator level. This value was features to limit hydrogen flow to the break.1lence, dam-based on discussions with INEL combustion experts who NURiiG-1364
.16
concluded that the leakage of large quantities of hydro-west wall of the turbine building where it enters at the pen into the ionized air surrounding the generator is al-15-foot lesel it passes hori/ontally through fire zones in most certain to cause irmneduuc igmtion. For releases which the hydrogen seal oil unit, sample and vent lines, into the auuliary buildmg and releases from the hydrogen and hydrogen control panel are located and then passes distnbutom system in the tmbme buildmp, INiii. as.
sertically through a fire zone at the 37-foot level to the sumed a wdue of 0.1 lor P(delay )to renett the presence of main generator at the 53-foot level. Each elevation has sipmheant amounts of elecuical eqtupment. such as one large open area that is dnided into fire /ones for pump and fan nunors, that could prouJe ignition sources convenience. The fire zones on each les el are not sepa-f or defLgranons m these buildmgs. INEl. assumed a ge-rated by any physical baniers llowever, different eleva-nerie value of 0.1 for l'tblast) lor the ausihary building and nons are separated by concrete floors.
for the penerator and distnhuuon system levels m the turhme builJmp t his choice was based on tonsideration None of the fire zones in the turbine building contain of the spatial imeracuons b c., t elative locations of bydro-safetyaclated equipment. The closest fire zones contain-pen sources and s.dety.related equipment)and the hkch-mg safety-rehtted equipment are in the control building, hooJ of blast mn ew or Dres d. unaging redundant safety-which is directly adjacent to the southeast corner of the related eqmpment.
turbine budding. A fire mne on the 15-foot level in the control buddmg contains the switchgear room in which 4.3.61:ncertainty Analpis vitiil 4s0 N buses and power and control cables for most saf ety related pumps are located.The wall separating the INEl. perhirmed the un1.crtamt3 aruh ses for the base turbme and control 1 uildings is about '00 feet from the case pLmt and pennie anal sci usmp[ the dNobution hydrogen lmes at thi' level. A fire zone en the 37-foot 3
rom the hydrogen hnes, contains funcuons f or random varah'esmJ the Monte Carlo stm.
elevation, about 60 feet f pling techmques budt mto the suustical 6 RISK com.
the cable spicading roon in which power md contrra puter pnpun (Pahsades Corporation,19SS) Some un.
cables for most safety rela ed pumps and vah's are lo-certaints mfor manon was obtained f rom the htensee's cated. A fue zone on the turuue hall floor at the N-ioot PR A.10r models with random event values bascJ on data elevation contains the control roota in which the control f orn the review of hvdwen esents at U.S. nuclear power p.meh and instrumentation for most safety-related sys-plank, the d&tribution f'uncunn and its parameters were tems.ue located.This level has the largest open volume calculated usine the methods reconunended in NURI:(it in the entire plant. The roof is about 100 feet above the CH-2300 liniertamtics m the salues of P(delay) and operating deck. and the wall is constructed of insulated f(blast i nere treated as staustical uneettamties. l oenor-metal sandwich pancis and contains large windows for mal dntnbuunns wet e auigned to these probabilitiek and natural hphting.
an error f;.ttor of 10 was awiened to reflect the larce modchnu tacertaimv.
The only T/DilR event considered for the turbine build-ing was a main steamline (MSI.) break caused by a large hsdrocen explosion on the operating deck.The estimated 4.4 Allilly sis Of Ilitse ('Inse l'Wl( l'lillit ebre damage frequency (CDI ) for this event was neglici-hie.
'l be base caw plut has a hun loop nue! car steam supp!)
system prouded by Westinehouse I lectric Corporanon.
The T!J.OCA events evaluated for the turbine building The reactor is beensed to operat" at a thermal power of were both random and seismically induced failures of the 3071 mepawatts thermal, corresponding to a turbme-hydrogen distribution lines at the 33-foot level that result pencratm output of 471 mepawauvelectrie.
in loss of the equipment in the cable sprearJing room (CSR). 'l he CSR was considered to be the most suscepti-4.4.1 Turbine lluiltling ble fire /one because of the small separation distance from the hy drogen lines and the types of fire barriers and The hydrogen stmage facihty supplymg the main penera-construction of the separating wall at the 33-foot level. A tor at the base case plant is located about 135 feet west of large fire in the CSR would result in the loss of power and the turbine buildmp. The facdity has 45 storage cylinders, control cables for most of the plant safety-related equip-each containine about i150 sef of hydrogen at a pressure ment (e.g., high-pressure injection pumps, component of about 1500 psig. I ighteen cylmders ar e on line as the coohnp water pumps. and residual heat removal and active supply. Another IS cyhnJers supply the reserve recirculation pumps) Assummg a value of unity for the mamfolth and 9 cylinders supply the emergenev reserve conditional probability of core damage given loss of the mamfold. The hydrogen hne to the main generator is not CSR. the point estimates of the mean CDF from non-equipped wit h an ewess flow valve or rest neting onhee to seismic and seismically mduced failures of the hydrogen automatically linut the rate of hydrogen flow to a large distnbution system would be about 1. l E-5/RY. I lowever, leals or break 6 the turbme buildmg. A 1.5-inch-dumeter because the susceptibility of the plant to large fires in the nyJropen.upply Ime runs from the storage faedity to the FSR had been recogni/ed, an alternate safe shutdown 17 NUREG-1364
system (ASSS) was installed at the base case plant in the 4.4.3 Ilvdrogen Storage Facilities early 19xus.The power and control cables for the ASSS, which performs a number of critical safety functions, are INI:1. performed a vital area analysis for the generator routed independently of the CSR. INiii. obtained the and V( T supplies and the hydrogen truck skid facility for probability of nonrecovery from a T/1.OC A condition, the VCT.The main generator supply met the !!PRi crite-using the installed ASSS, from the original base case ria for safe separation distance from safety-related strue-plant fire PR A. 'the mean value of 0.0% consists of a tures and air intakes. The V(T supply and truck skid human error contribution (0.039) and hardware-related supply met the !!PRI criteria for detonation but did not failures du)7). Ilence, with credit for use of the ASSS, the meet the criteria for air intakes. Ilowever, consideration Cl)l for these cvents was estimated to be about of intervening buildings and differences in elevations 5F 7/RY.
along possible paths from the supplies to the safety-related air intakes indicated a negligible likelihood that a safety-related air intake would ingest a combustible 4,4,2 l'rimary Auxiliary lluilding hydrogen-air mtxture.
The hydrogen supply system for the VLT at the base case 4.4.4 Uncertainty Analysis 1(esults J
pkmt is mJependent of the supply system for the main pencrator. It consists of a 12-bottle supply located next to in summary, the evaluation of the risk from hydrogen the primary auxiliary buildmg (Pall). The supply contains system failures at the base case plant mdicated that the a maumum of abou't 2400 scf of hydrogen at about 2000 mean com damne imluency for hydrogen events at the psig. 'there is no excess flow valve or restricting orifice to storage facihties and in the turbine and auuliary buildmgs limit in drocen flow rate following a large leak or break in is is than IF-6/RY and is negligible when compared the au'xihAy buildmg. A hydrogen truck skid about 200 wah the values from other causes identified in the plant EN A' feet away,1rovides refilhng capahdity through a 1-inch-diameter field-run pipe. The 12-bottle bank is used for normal service, and the tr uck, which is normally isolated-4.5 Generic Analysis is used a, a backup.
The generic analysis addressed the risk from hydrogen t udon ymms fm N wM es dif@f erent from those of the base The hydrogen supply pip.; enters the PAH at the 92-foot Mo{ age a es an clevation, trasels hori/ontally for about 25 feet, enters gn daram case plant, sus as locadon of Mmpn supply facilities and travels about 20 feet through a pipe chase, and then and distnbution systems relauve to vital equipment [c.g.,
enters the VCT eubicle and tank at the 98-foot elevation.
auxiliary feedwater (Al:W) system, diesels, electrical in the vital area analysis for this plant, most fire zones in sw gar, cane spreading room, essential service water the Pall were screened out either because no safety-
@f eed-and-bleed (F&II) capability. As mlater system) and
)g e m,an component cooling w related components were present or the fue /one was the base case h>cated two or more les els below the zones containing the phmt analysis. the generic analysis was based on the fact hydrogen lines.
that the hydrogen system is not near the reactor pressure sessel and control rod dnvc motors, the prirnary coolant The only non seismic T/DilR event considered was a piping, or the steam generators. llence, INiii. did not l
general turbine trip es ent caused by a small fire or explo-consider pipe-break I.OC As, steam generator tube rup-sion.The estimated CDl for this event was insignificant.
tures, and KlWS-type scenarios due to hydrogen events.
The initiating event frequency for seismically induced failures of the hydrogen lines in the Pall was considered 4.5.1 hiethod l
negligible because of Ihe seismic qualiheation of Ihe lines.
I A non-seismie T/1.OCA cvent in the Pall was considered The generie approach involved a vital area analysis in because the component cooling water (CCW) heat ex-which hydrogen release and fire or explosion scenarios changers in Fire Zone 1%7A were located on the same were related to specific system failure scenarios. The level as the VCT. !!owever, in view of the limited hydro-generie analysis did not include the level of detail used in gen supply to the VLT, partial shielding by intersening a ty pical plant-specific vital area analysis, but was directed concrete walls, and the separation distance from the hy-at key systems used to mitigate the two categories of drogen lines, the probability that the CCW heat exchang-transients. In particular, for transients with loss of decay ers would be lost because of a hydrogen fire or explosion heat removal (T/DilR), attention was focused on the at that level was considered to be very small. As noted AFW system and systems needed for F&B cooling. For previously, seismic events were considered to be neghri-transient induced 1 OCAs (T/l OCAs), attention was fo-ble contributors to risk in the PAH of the base ese plant cused on normal and emergency ac power and the essen-because of the seismic qualification of the supply line to tial SW and component cooling water systems. If this the VCT.
genene approach is used, a plant-specific design feature i
N URFG-1364 18
that produces a special outlier vulnerability will not be mal barrier heat exchangers and by RCP seal injection idenufied.
(which is pnivided by the charging pumns). Ilut loss of the CCW system leads to loss of the charging purnps; there-4.5.1.1 T/DilR Transients fore, both means of cooling the RCP seals are lost if the ECW system is lost. Since the oil coolers of the high-INEl. separated the TiDllR transients into several cate-pressure safety injection pumps and the charging pumps pories according to the location of the AFW estem and are cooled by CCW, loss of the CCW system could also the plant's F&ll capability. In sorne plants, Al W and lead to loss of the capability to mitigate the small! OCA 1 &ll systems are located in separate buildings away from caused by seal failure (NUREG/CR-5759),
the hydropen storage facihties and distribution systems in these plants, no interactions would be expected he-4.5.? Generie Anahsis RcStills tween the initiating event and AFW system failure,llow-E2 J Tm bice Ituilding and Auxiliary lluil ling ever, in some plants, AFW systems are located in the
^""
ausikary or turbine buildings. For T/DilR transients m the turbine buildmg. INEl. assumed that a hydrogen fue The Til)llR and TiLOCA transients are categori/cd ac-or explosion suflicient to cause loss of the Al W system con.hng to the locatian of the Ai/W, CCW, and other vital wuuld also cause loss of the main feedwater (Mi W) sys-equipment relative to the hydrogen sy4 cms and the tem. Ilence, actions to recover the MFW system on to plant's feed-and-bleed capahdity. 'lhese combinations depresson/e the steam generators and use the conden.
are cosered by the configurations given in Table 2. Ta-I sate system would be incifective.
ble 2 summari/es the mean core damage ficquencies for the specific PWR configurations considered by INEl. in in some phmts. if Al W cooling fails, operators can start a evaluating the in-plant risk from the hydrogen supply am!
i I
hightewmc salety injecoon pump and manually open distribution sy stems and the specific acciJent scenanos to the power-operated relief valves (PORVs) to provide be considered in these generic evaluations.
1 NH coohng to the pnmary system. Almost all Westim phouse M) and llabcock & Wilcox (ll&W) phmts have 4.5.2.2 Ilydrogen Stnrage Facility Analysis this capahihh. l'or W and il&W plants, INEl. assumed An informal survey of all plants showed that about 30 that Al All ladure) was equal to 0.045. This value is the plants did not meet the EPRI criteria lor separation dis-amage of four different F&ll failure probabilities from tances for safeterelated air intakes or structures. Of different plant PR As and depends almost exclusively on these plants, mitigating factors were insufficient at three human error probab:hties and not on design considerd-plants. The results of additional evaluations by INhl. of tions. llecause newer Combustion Engineering (Cl ) re-these three plants are as follows.
actms do not hase Pt)RVs. coohng the core usmp the F&ll function is not possible. INI 1. also concluded in At Plant A. the hydrogen storage facility (with 14,400 scf NURE(UCR 57591 hat older CE reactors have marginal of hydrogen in two tanks)is located about 30 feet from the F&ll conhng capability and cited previous analyses main steam enclosure, which contains motor operated (NUREU<CIU4471 and NURI (bCIU5072) that indi-Al W valves, main feedwater (N1FW) piping, and rnain cated Ihat only a short time window wonlJ be availabic for steam stop and safety valves. The scenario considered was initiating successful 1 < recovery operations. In view of a random rupture of a hydrogen storage tank resulting in a the higher prionty awirned to other recovery operators detonation causing loss of the MFW and AFW systems.
INI.I. assumed in this penene study that F&ll cooling was Recovery is made using I &ll capability. Since the actual not a siable heat remmal mcchanism for CE plants.
separation distance was a few feet greater than the EPRI separation distance criterion, the probability ofloss of the MI W and Al W systems, giver, a detonation, was as-4.5.1.2 Til OCA Transients somed equal to 0.1. The frequency of all hydrogen fires The majority of PWRs probably do not have systems such and explosions, based on data in licensee event reports, as emergency diesels and essential SW systems in budJ-was 2.51W3/RY. I rom these values and a value of 0.045 ings where larpe hydrogen systems are located. However, for the failure probability of I &ll coohng, the estimated INEl.noted that in at least one PWR, a key portion of the mean CDF from the uncertainty analysis is IE-5/RY.
component coohng water (CCW) system (CCW heat ex-The hydrogen facility is also near the air intake for the changers) is located in the turbine building. In most pumphouse containing the essential SW pumps.The air-l PWRs, the ( CW system is located m the auxdiary budd-intake scenario considered was a random failure of the ing. At many PWR plants (including, the base case plant),
hydrogen header upstream of the pressure regulators loss of the CCW system (e.g., caused by a large hydrogen leadmg to ingestion of a combustible mixture at the air explosion fading CCW heat exchanpers) would seriously intakes and a fire or explosion causing loss of the SW affect the plant's capahdity to cool the reactor coolant pumps and consequent loss of the motor-driven Al W pump (RCP) seals. At many plants, cooling of the RCP pumps. The turbine-driven AFW pump would still be seals is performed redundant!y by CCW flow to the ther-available.The mean CDU for t his scenario is 2.51!-8/RY.
19 N U Rl!G-1364
Plant 11 did not meet the separation criteria for either large releases of the hydrogen stored in the generator, structures or air intakes. The hydrogen storage facility The reductions in CDF for the alternatives are given in contains 85 standard hydrogen bottles (about 200 sef Tabic 3. The values in the table are the mean values each). At least four bottles are on hne. The hydrogen obtamed hom the uncertainty analysis.
supply is not seismically quahfied and was assumed to have the smne fragility as that used for the hydrogen 4.7 Dose C,onsequence Analys,s i
system piping in the generic calculations for the auxiliary and turbine buildingsflhe hydrogen supply is close to the The dose consequence analysis by INiil for this safety two diesel generators. A storage bay containing a large issue was based on the information developed for Safety
~
quanuty of turbine lubncatmg od islocated between the issue A-45 (NUltI!G/ Cit-4762). For that study, Sandia hydrogen facihty and the diesel generators. ihe scenario National I ahoratories used Calculation of Itcactor Acci-considered was a scismic event causing loss of offsite dent Consequences (CitAC) Version 2 computer code power because of failure of ceramic msulatms at the results for integrated doses (in person-rem per event) of fsite power transformer and a seismically induced rup-
'd by the population around the plant out to 50 ture of a hydrogen storage bottle. Since a hydrogen fire or detonation can also cause a large fire at the adjacent we bound beesunjate, and uppu Nund vah m
lubricating oil storage area, loss of both diesel generators 1p were calculated for each of seven release categories.
u or the generic PWit analysis of NUlt!!G/ Cit-4762, a was assumed. Ilence, the seismic event could result in a containment event tree with three containment functions station blackout. The mean values of the initiating event and six containment sequences was used. Each accident frequency and the CDF for this scenario are 5.3fi-5/f tY.
quena mntributing to the C Dl was mapped to one or For the a'ir mtake analysis, the concern was also the prox-mme et the mntainment accident sequences, hacn con.
imity of the diesel generators. The scenario considered tainment sequence was then mapped to containment fail.
was a random failure of a pipe at the hydrogen storage um modes wie asociated protnbilities of occunence per area during a loss of offsite power that causes loss of both event and release fract ons, diesel generators because of ingestion of a flammable mixture at the air intakes. The point estimate value of Cl W for this event is 6.61!-7/ItY.
4.8 Cost Analysis Plant C did not meet the liPRI separation criteria for air The cost-benefit analysis for the various alternatives fol-intakes. The concern was the proximity to the hydrogen lowed the guidelines of NUltliG/ Cit-3568 and NUltliG/
storage facility of the air intake for the buildingcontaining CR-4627, Revision 2. Costs were calculated using the 4160-V switchgear.1,oss of this switchgear would FORl! CAST 2.1 (Science & Engineering Associates, cause loss of most safety-related equipment. Important Inc.,1990).
cxceptions are the turbine driven AFW pump and the diesel generators. The scenario considered was a random AU plant cost estimates are given in 1991 dollars and failure of a pipe at the facility causing releases of un.
include implementation md recurring costs. IN!!L as-burned hydrogen, infestion of a combustible mixture at sumed that modifications would be made during normal the air intakes, and an explosion resulting in loss of the plant operations or scheduled shutdowns and that no switchgear. Recovery from this event is made with the replacement energy costs would be incurred. The cost turbine-driven AFW pump. The pomt estimate value of analyses covered the following component costs. These UDF is 2.51F6/RY.
costs are discussed in IN!!!,1992; Science & Engineering Associates, Inc.,1992; and NUlt!!G/CR-5759.
'o of cyuipment, materiai, and siructures 4.6 IlO(lllClions in Core namage e
instaHation and removal costs and associated over-IIreyt!CllCy for IluprOV0mCH(
head l
Altcritatives engincenng and quality assurance costs e
Sections 4.4 and 4.5 give estimates of the hydrogen-radiation exposure costs e
initiated CDF contributions for the base case plant and for other generic PWR plant configurations before any health physics support costs procedural or hardware changes are made to reduce risk.
hansce costs for rewriting procedures, staff train-e Configuration I scenarios are characterized by a hydrogen in g, nd oWu tuhmcal subtasks event at the generator floor level leading to a general licensee recurring costs plant transient or steamline break. INiilm assumed that a
the contributions of the Configuration I accident scenario NRC implementation costs e
were not reduced by any of the alternatives for the turbine NRC recurring costs building, since none of the alternatives could prevent e
NURl!G-1364 20 l
l
onsite averted costs representing the aserted onsite 0.05. Table 5 gives similar results for a remaining plant e
property darnages, includmg allowances for cleanup, hfe of 40 years to illustrate the change in values for a plant repair, and replacernent energy costs with a license renewal for an additional 2O years. A nega-tive value indicates a cost savings is predicted because of Table 4 gises the point estimates of the total cost for each the inclusion of onsite averted costs.The calculations are alternative rninus the onsite averted costs for a t emaining described in detail in I N iii., 1992, and Ntiltli(1/
plant hfe of 20 years and a liest-estimate discount rate of
('ll 5759.
t l
l l
l 2I NUlW(i-1361
5 COST-ilENEFIT ANALYSIS The cost benefit methodology for analyzing the various breaks in the lines near the generator and at the genera-alternatives considered for this safety issue is based on the tor itself. It applies to Configu rations H, Ill, and IV, which requirements of the backfit rule (10 CFR 50.109)and the involve leaks or breaks at the hydrogen distribution sys-guidance in memoranda from 13 S. Beckjord dated tem levels in the turbine building. Configuration 11 ap-May 10 and November 18,19S8, and NURI!G/CR-356S.
plies to Habcock & Wilcox (B&W)and Westinghouse (W)
One consideration in the decision-making process is the plants and the loss of the auxihary feedwater (AUW) cost-benefit ratio for each alternative, evaluated in terms system located in the turbine building. Alternative 2 is not of cost in 1991 dollars per person-rem averted, which may cost effective for this configuration because of the small be compared to a guideline such as $1000/ person. rem.
delta core damage frequency (CDF) at these plants that The costs used in the cost-benefit ratio are (1) the total have feed-and-bleed (F& B) eapability. For Configu ration cost of an alternative without consideration of the onsite Ill, insolving loss of the AFW system located in the tur-averted costs (OS ACs), given by the sum of the first nine hine budding at Combustion !!ngineering (CH) plants, componem costs listed in Section 4.S. and (2) the nel cost and Con figuration IV, involving loss of vital equipment in of an alttrnative, which equals total cost minus the the turbine building at all PWRs, INEL obtained a ge-OSACs. Tne other consideration in the decision-making neric best-estimate value of about Ifi-5/RY for delta process is the magnitude of the risk reduction achieved by CDF. If onsite averted costs are included, there is a cost the alternative action. In the following sections, each savings for Configurations tit and IV. Tabic 6 shows there alternative is described and the results of a cost-benefit is nearly a 100-percent chance that the cost-benefit ratio assessment are given.
is less than $1000/ person-rem for these configurations.
IN!!Lassumed that Alternative 2 did not reduce the CDF IN!!L used the @ RISK computer program (Palisades for Configuration V, which is applicable to hydrogen re-Corporation,1988) to evaluate the uncertainty in the leases at or near the generator.
cost-benefit analysis with and without OSACs. Input to the program included the cost of the alternative, the benefit in offsite person-rem averted, and OS ACs. Tables 5.3 Alternative 3-Provicle blanual i
6 and 7 give the results, which were obtamed from the cumulative distributions for the cost-benefit ratio for a blakeup of IlyClrogen to Generator remaining plant hfe of 20 or 40 years.These tables give an(1 Check Valve or Restricting the chance that the cost-benefit ratio is more or less than Orifice at the Generator
$ 1000/ person. rem.
Alternative 3 entails operation with the hydrogen facility 5.1 Alternative 1-Take No Action normaHy isolated from the generator, except for manual adjustment of hydrogen conditions m the generator, and the installation of a check valve in the hydrogen line near Under this alternative there will be no new regulatory the generator to prevent backflow to the hydrogen line requirements. Consistent with existing regulations, this alternative does not preclude a licensee from proposing break. Alarge portion of the cost forth,salternativeisthe i
to the NRC staff other changes intended to reduce the recurring costs associated with manual makeup to the generator.
risk associated with the hydrogen storage and distribution systems on a plant-specific basis.
Alternative 3 applies to Configurations U, Ill, and IV.
INI!L assumed that this alternative did not reduce the 5.2 Alternative 2-Install Low CDU for Configuration V, which is applicable to hydro.
Setpoint Excess Flow Valves, gen releases at or near the generator. The alternative is not et beneficial for Configuration H (loss of the AFW RcStriClilig Orifices, or Check system in the turbine building for E and H&W plants)
T,alveS in the ll,(Irogen Supply because of the small della CDU for these plants that have 3
Line to the Generator F& H capability. For Configurations IU (loss of the AFW system in the turbine building for CE plants)and IV (loss This alternative provides protection from large breaks in of vital equipment in the turbine building for all PWRs),
the hydrogen supply line to the hydrogen control station the generic best-estimate delta CDF is 1.Oli-5/RY. If at the generator by limiting flow from the hydrogen stoc-onsite averted costs are included for these configurations, age facility and backflow from the generator.The alterna-Table 6 shows that there is nearly a 100-percent chance tive also limits flow from the storage facility to leaks or that the cost-benefit ratio is less than $1000/ person-rem.
23 NURl!G-1364
5.4 Alternative 4-Enclose Safety.
ment in the AH (sec Table 2)JFhe alternatives with onsite Related E4quipment Located in aserted costs included are not cost effective for Conhgu.
ration VI (loss of the AFW system in the AB for W and
'I urblile lluild,llig in lllast-and H&W plants) because of the small delta CDF at plants fire-Proof Structures with F&B capability. For Configurations Vl!(loss of the AFW system in the AH for Cli plants) and VIII (loss of -
'Ihis alternative, which applies to Configurations II, Ill.
vital equipment in the All for all PWRs), the INiil esti-IV, and V, entails structural modifications to protect the mate of a generic value for delta CDF is about vital equipment from damage by large fires or explosions.
0.511-5/RY. The alternatives are cost effective for these INiil, initially considered it as a possible alternative for configurations if onsite averted costs are included. Table plants with higher risks because of safety-related equip-6 shows that the chance that the cost-benefit ratio is less ment in the turbine building. The alternative is not cost than $1000/ person-rem is about 100 percent for Alterna-clfective with or without onsite averted costs because of live 5 and 88 percent for Alternative SA.
the large costs and the generic delta CDFs estimated for these configurations. Table 6 shows rnore than a 99-percent chance that the cost-benefit ratio is greater d.6 Alterinat.ive 6-L.imit flie Quantity than $1000/ person-rem.
of Ilydrogen Normally Connected to Volume Control Tank 5.5 Alternative 5-Install Low This altemative involves a hmit on the total amount of Setpoint Excess Flow Valve or hydrogen in the storage facility that is normally connected IleStricting Orifice in flydrogen to the volume control tank at any time. The limit is such tha th u a n uh quent Gre or dctonadon of the Distrihtition S}' Stem to Volume hsdrogen m the supply and the \\ C'l, would not cause Control Tank and Provide unacceptable damage to safety-related systems in the
~
llydrogen Detectors, ir Needed auxihary building.
This alternative involves the use of a low setpoint excess This alternative also applies to Configurations VI, Vil, now valve or restricting orifice in the hydrogen supply line and VIII, which involve loss of the AFW system or vital outside the auxiliary building (AU). The purpose is to equipment in the auxiliary building. For Configuration VI limit the rate of hydrogen flow from the storage facility to (loss of the AFW system in the AB for E and ll&W larger breaks orleaks to a low value so that normal venti-plants), the alternative is not cost effective if onsite lation in the compartment with the leak or break can keep averted costs are included because of the small detta CDP the average hydrogen concentration well below the lower at these plants that have F&B capability. For Configura-flammability limit. Protection against the accumulation of tions VH (loss of the AFW system in the AH for Cli unacceptable amounts of trapped hydrogen from leaks at plants)and VIII(loss of vital equipment in the All forall flow rates up to this maximum flow rate would also be plants), the generic value of della CDF is 0.511-5/RY.
j provided if necessary, Trapped hydrogen may not be a This alternative is cost effective for t hese configurations if concern at most plants if the low setpoint excess flow onsite averted costs are included. Table 6 shows that valve or restricting orifice is used. Ilence, as discussed in there is about a 100-percent chance that the cost-benefit Section 3.4, the costs for this alternative were estimated ratio is less than $ 1000/ person rem.
for two cases. Alternative 5 costs do not include the instal-lation of permanent hydrogen detectors. Alternative SA costs include the additional costs of permanent hydrogen 53 mrndive 7-Provide Normall}*
leak detectors (see !!PRI,1987 and Science & lingineer Isolated Supply With Daily ing Associates. Inc.,1992) that provide an input to a local Manual Makeup of IIVdrogen to panel sending an alarrn to the control room. The costs Volume Control Tanii melude the costs of (1) the hydrogen detectors, (2) peri-odic testing and maintenance of the excess flow check Thisalternative involvesisolation of the hydrogen storage valves and hydrogen detectors,(3) administrative controls facility from the VCT except for brief daily operation to and/or design features to prevent the bypassing of the adjust VLT conditions. The dominant cost for this alter-excess flow check valves or restricting orifices, and (4) ad-native is the recurring plant cost for manual makeup.
ministrative controls to provide for manual isolation of the hydrogen supply if normal ventilation to the AB is This alternative applies to Configurations VI, VII and lost.
Vill, which involve loss of the AFW system or vital equip-ment in the auxiliary building. For Configuration VI(loss These alternatives apply to Configurations VI, Vn. and of the AFW system in the AH for E and H&W plants),
Vill, which involve loss of the ARV system or vital equip-the alternative is not cost beneficial if onsite averted costs NURl!G-1364 24
~-..-. --.
. ~__
are included because of the small delta CDF for these of flammable hydrogen-air mixtures at safety-related air plants that have F&H capability. For Configurations Vil intakes, applies to Plants A. H, and C. Since the delta and Vill, the generic best-estimate value of delta CDF is CDFs are small because of the small initial CDFs, this -
0.51i-5/RYJThis alternative is still not cost effective be-alternative is not cost effective. Table 6 shows that the cause of the higher costs. Table 6 shows that there is a chance that the cost-benefit ratio is less than SI-percent chance that the cost-benefit ratio is less than
$1000/ person-rem is less than 8 percent if onsite averted
$ 1000/ person-rem.
costs are included.
5.8 Alternative 8-Relocate Ilydrogen Storage Facility To Meet d.11 Other Alternat.ives h,eparat. ion Distance 17 rom Safety-Related Structures Other alternatives are possible that could also reduce an This alternative involves the relocation of Ihe hydrogen existing plant vulnerability at either the auxiliaty building, storage facility to reduce the probabihty that a detonation the turbine building, or the hydrogen storage facility.The at the storage facility will damage safety-related equip-alternativcs discussed in Sections 5.1 through 5.10 pro-ment.The mean delta CDFs for this alternative are about vide a perspective on a range of options and could serve as 1.0E-5/RY for Plant A and 5.3E-5/RY for Plant H. If guidance on the cost effectiveness and benefits to be onsite averted costs are included, the alternative is not expected from other possibilities.
cost effective for Plant A, but is cost effective for Plant B.
Table 6 shows that the chance that the cost-benefit ratio is less than $1000/ person-tem is 6 percent for Plant A and w percent for Plant n.
5.12 Isife Extension Considerations 3.9 Alternative 9-Install lllast
'lhe NRC staff is developing the regulatory requirement Deflection Shield at Ilydrogen for the renewal of operating licenses. A license may be Storage Itacility renewed for an additional 20 years if the licensee meets the specific requirements of the license renewal rule.The This alternative involves the addition of a blast shield to effect of license renewal on the evaluation of this safety protect safety-related equipment from a detonation at issue was included by repeating the calculations for a the hydrogen storage facility.The alternative was consid-remaining plant life of 40 years (current remaining life of cred only for Plant A. since a blast shield could not be 20 years plus a license renewal of 20 years). The results used at Plant H beca use of insufficien t space. The alterna-are shown in Table 7.The increases in onsite averted costs tive does not apply to Plant C, which did not meet the and benefits with increase in remaining plant life cause a separation distance criteria for air intakes. The best-decrease in the cost-benefit ratio (more cost effective) estimate delta CDF for Plant A is 1.0E-5/RY. Table 6 that can be offset, in part, for some alternatives by the shows that the chance that the cost-benefit ratio is less increased contribution of recurring costs.
than $1000/ person-rem is about 100 percent if onsite averted costs are included.
F 10 Alternative 10-Install Ilydrogen 5.13 New Reactors Analyzer-Actuated Air Intake Imuvres at Safety-Related Air The implementation of the resolution of GSI-106 in-i Iniakes cludes a reccmmendation that the Standard Review Plan (NURIIG-0800) be modified to include new guidance on
?
This alternative, which involves the addition of shutters hydrogen storage facilities and distribution systems for actuated by hydrogen detectors to prevent the ingestion the VCT and generator at future PWRs and HWRs.
i s
1
)
A j
25 NUREG-1364
l 1
1 6 DECISION RATIONALE 6.1 IIllrOdtlCliOH much larger estimated costs because of recurring costs associated with manual adjustment of VC1'conditionsc As noted in Section 5, the estimated cost savings and cost-benefit ratios vary significantly for the alternatives Plant;Specilie Conditions in Ausiliary and Turbine associated with the hydrogen storage facilities and the lloildmgs hydrogen distribution system for the volume control tank A significant number of hydrogen events have occurred (VCl)and electric generator at PWils-and continue to occur at U.S nuclear power plants. In recent years, most of the significant events involving hy-Ilydrogen Storage Facilities drogen fires, explosions, or large hydrogen releases have occut red in turbine buildings, llowever, hydrogen events The estimates for the hydrogen storage facilities indicate in the turbine building are not expected to be significant that Alternative 8, which involves the rekication of the sources of risk for most plants. For T/LOCA transients, storage facility, results in a della core damage frequency vital equipment is not expected to be in the tuibine build-(CDiz)of about 51i-5/It Y for Plant 11 and is cost effective ing for most plants. For T/DUR transients, the risk is if onsite aver ted costs are included.170r Plant A, Alterna-small at the Westinghouse and llabcock & Wilcox plants tive 9, which involves the addition of a blast shield, results because of their feed-and-b!ced capabilities (additional in a delta CDF of about II!--5/ItY and is cost effective if recovery operations include recovery of the main feed-onsite averted costs are included. Alternative 10, which water system or depressurization of steam generators and involves the air intake louvres, is not cost effective be-the use of condensate). At most CII plants there would cause of the small values of CDF associated with the air also be a minimal risk from the T/DIlit transient because intakes for these plants.
the AFW pumps are located (1)in a separate building or (2)in both the auxiliary and turbiac buildings. Hence, a sinele hydrogen event would not be expected to result in
,I.m.b.me lloild.mg bMW M',I buver, for the small number of plants The generic estimates mdicate that Alternative 2, which that could be susceptible to core damage resulting frora involves the insudiation of flow-limiting devices, and Al-T/l OCA orpossiblyT/DilR transients becausc of events ternative 3, which invokes operation with a normally in the turbine building, the magnitude of the generic isolated generator, would be cost effective in reducin'g estimates of CDF (lli-5/itY) and cost-benefit ratio indi-risk for Combustion lingineering (CH) plants with the cates that some alternatives considered m this analysis auxiliary feedwater (AFW) system in the turbine building could be warranted, or all PWRs with vital equipment in the turbine building Although there have been a number of leaks, there have and hydrogen releases at the hydrogen dlstnbution sys.
l>cen no fires, explosions, or large hydrogen releases in tem level (Lonfigurations lil and IV of,Iable 2). Ihese the auxiliary building. Ilowever, some plants are consid-alternalges result m a best-estunate delta C DF of about cred to be susceptible to core damage because protective E5/Ri. Alternative 7, w hich involves protection of the features are lacking to prevent large hydrogen releases safety-related equipment, is not cost eff ective.
or, possibly, the buildup of significant amounts of trapped hydrogen in the auxiliary building. The hydrogen events Auxiliary !!nilding considered potentially significant are those resulting in loss of vital equipment and a reactor coolant pump seal lyr the auxiliary building, the evaluation indicates that 1.OCA. An appreciable number of plants have corrective Alternatives 5 and 5A and Alternative 6 would be cost n casures such as normally isolated supplies, limited sup-effective for CE plants with the AFW system in the auxil-plies, flow-limiting devices, and leak detection equipment lary building or all PWRs with vital equipment in the and procedures. Therefore, a relatively small number of auxiliary building (Configurations VII and VH1 of Table plants may need changes to reduce risk, The generic CDF 2)if onsite averted costs are included. The generic best-and cost-benefit ratios for these events indicate that some estimate delta CDU for the alternatives is about alternatives could be warranted for those plants that do 0.5E-5/RY. Alternative 5 involves the use of a flow-limit
- not have protective features.
ing device in the supply line to the VCT. Alternative 5A involves the use of a flow-limiting device in the supply line The generie delta CDFs for the various alternatives are to the VCTand permanent hydrogen detectors. Alterna-based on sensitivity studies of the base case plant to better tive 6 involves the use of a limited hydrogen supply to the quantify values for other existing plants that (1) have large VCF. Alternative 7, which involves the use of a normally hydrogen facilities that are normally connected to the isolated supply,is not cost effective. This alternative had VCI'and generator via pressure regulators and (2) do not the same delta CDF for Configurations VII and VUI, but have low setpoint excess now valves or restricting orifices 27 N UREG-1364 s
m
..___.,....-..__..--___.__._.-_._m
l to limit the rate of flow f rom the facility following a large corresponding increase in the CDF (e.g., about break in the hydrogen supply line in these buildings.
IE-5/RY).
'these generic calculations are characterized by a single
~,
set of values for initiating event frequencies and for the l
6.2 Relat.ionsliip to Otlier Generic probability of a delay in ignition, P(delay), and the prob.
abdity of blast damage to redundant safety systems, ISSUES P(blast). Depending on the relative h> cations of the safety equipment and hydrogen lines, and other factors such as Hecause the T/LOCA transients in this evaluation en-use of a limited amount of stored hydrogen or excess flow compass those hydrogen-induced system failures that valves and Icak detection, these values and, hence, the lead primarily to reactor coolant pump (RCP) seal delta CDFs for a given plant may be significantly lower or LOCAs, GI-23," Reactor Coolant Pump Seal Failures,"
i higher than the generic values. For example, the prob.
is related to GSI-106.The objective of GI-23 is to reduce j
ability of damage to component cooling water (CCW) the probability of RCP seal failures and, hence, make it a heat exchangers at the base case plant that are k>cated on small contributor to the total CDF, GI-23 could entail i
the same level as the VCT was considered to be very small the addition of a separate and independent cooling sys-because of thc limited amount of hydrogen, partial shield.
tem for the RCP seals and could provide part of the ing by several concrete walls, and a large separation dis.
resolution of GSI-106 because it would climinate most of tance. Other fire zones were climinated from considera.
the delta CDF for the T/LOCA scenarios.
tion because they contained no safety-related equipment or were located two or more levels below zones contain-6.3 Backfit Rule and Plant-Specific ing hydrogen components. Damage to a seismic Cate-gory i six-inch scivice water line located near a short COnSiderat. ions length of hydrogen line in a pipe chase was climinated As discussed in Section 2, the overall objective of from considerahon m the base case plant because of a GSI-106 is to ensure that the contribution from the usc of relatively large separation distance, lack of local igmtion combustible gases to the total CDF is less than about sources, and significant ventilation.
1E-5/RY The generic calculations indicate that some plants may have a CDF due to hydrogen events of more llowever, other plants may have such building arrange-than IE-5/RY. However, it is apparent that there are ments that the hydrogen lines are relatively close to vital large and diverse plant-to-plant differences in equipment considered in the risk analysis. Limited infor-relative locations of hydrogen systems and safety-mation on the kication of safety-related equipment rela.
e tive to the hydrogen lines and VCTin the auxiliary build-related equipment i
ing was obtained from site surveys conducted early in the hydrogen storage and distribution system safety fea.
e program. These surveys showed that rooms contaimng tures, operating procedures, and considerations of CCW pumps and heat exchangers were at the same level trapped hydrogen or at levels adjacent to the level containing hydrogen hnes reactor characteristics that affect risk from hydrogen e
at six of nine plants surveyed. Motor control centers at 5 of 6 plants and switchgear rooms at 4 of 14 plants were events also at the same or adjacent levels. A survey of the kica-tion of the auxiliary feedwater (AITV) system at 14 plants Hence, only plant-specific evaluations can determine the showed that the AFW pumps were in the turbine building extent to which a modification is justified.
at 3 plants, the auxiliary budding at 5 plants, and in other a
buildings at 6 plants. Of the five plants with AITV pumps 6.4 COneluSiOn m the auxiliary building, three had pumps at the same level or levels adjacent to those containing hydrogen In view of the observed large differences in plant-specific lines. An additional review of one of the plants showed characteristics affecting the risk associated with the use of that the hydrogen supply line to the VCT in the auxiliary hydrogen, and the marginal generic safety benefit that building was located next to the compartment containing can be achieved in a cost-effective manner,it is concluded the CCW pumps and heat exchangers. The generic esti-that this generic issue be resolved simply by making these mate of the CDF for the T/LOCA transient following loss results available in a generic letter.This information may of the CCW system is about 0.5E-5/RY, Ilowever, the help licensees in their plant evaluations recommended by j
proximity of:he hydrogen supply line to the CCW compo-Generic Letter 88-20, Supplement 4, " Individual Plant nents at this plant could result in a higher conditional Examination of External Events forSevere Accident Vul-probability of damage to safety-related equipment and a nerabilities," June 28,1991.
i NUREG-1364 28
~.
7 REFERENCES American National Standards institute, A13.1," Scheme
, from II. S. Ileckjord, N RC, to J. M. Taylor,1!xecu tive for the Identification of Piping Systems," 1975.
Director for Operations, NRC, " Resolution of Generic issue (GI) 121,'llydrogen Control for PWR Dry Contain-
-, Z35.1, "American National Standard Specifications ments,' " March 24,1992.
for Accident Prevention Signs," 1972.
, from C. J. licitemes, Chair Regulatory Analysis Code ofrederal Regulations. Title 10, "linergy," U.S. Gov.
Stecring Group, to J. M. Taylor, lixecutive Director for ernment Printing Office, Washington, D.C., revised peri-Operations, NRC, "Comrmssion Paper on Safety Goal
- odically, Implementation, August 20,1991.
, from R. l Pressard, NRC, for distribution," Lessons Consumers Power Coinpany, PNO-III-91-51, "Ilydro' Learned From the incident on October 19,1989, at the gen Leakage into Turbine lluildmg," Palisades, Dece -
Vandellos I Nuclear Power Plant (Spain)," September 4, ber 10,1991, ig9;,
I!!ectric Power Research Institute (llPRI),
National Aeronautics and Space Administration NP-5283-SR-A, " Guidelines for Permanent ilWR Ily-(NASA),Ilydrogen Safety Afanual. NASA TM X-52454, drogen Water Chemistry Installations-1987 Revision,"
F. II. liciles,1968.
September 1987.
, Lewis Safety Alanual. NAS A TM 104438, 1992.
Factory Mutual Research Corporation, Report COO-4442-4, " Compilation and Analysis of Ilydrogen National Itureau of Standards, Technical Note 690, "Is Accident Reports," October 1978.
Ilydrogen Safe?" J. Ilord, October 1976.
Idaho National Engineering 1.aboratory (INEl.), EGG,-
Niagara Mohawk Power Corporation, PNO-1-91-73,
" Uncontrolled Release of a Flammabic Gas in the Tur-NTA-9082,"Scopmg Risk Analysis of Ilighly C.ombusti-bine lluilding," Nine Mile Point Unit 2, October 24,1991.
ble Gas Storage, Supply and Distribution Systems m lloil-ing Water Reactor Plants," G. P. Simion et al., November 1991-Nuc/ conics Weck, August 8,1985; August 22,1985; Dc.
, !!GG-SSRl!-10198, " Risk Analysis of Ilighly Com-bustible Gas Storage, Supply, and Distribution Systems in Palisades Corporation, & Risk, Version 1.5, " Risk Analy.
Pressuri/cd Water Reactor Plants-Supplementary sis Modelling for the PC," March 1,1988.
Cost /Henefit Analysis," R. Van Ilorn, C. Smith, and G.
Simion, March 1992.
Public Service lilectric and Gas Company, LliR 311-91-017,"l'urbine and Generator Failure and Fire,"
Letter from J. Richardson, NRC, to G.11. Neils, IlWR Salem Unit 2, December 9,1991.
Owners Group II, " Acceptance for Referencing of Li-censingTopical Report Titled ' Guidelines for Permanent Science & Engineering Associates, Inc., Report SI!A ilWR llydrogen Water Chemistry installations,' 1987 89-461-04-A:1, "FOR!! CAST 2.1 User Manual," !!.
Revision," July 13,1987.
l#pez and F. W. Sciacca, April 1990.
Report SliA 91-554 01-A:1,"llackfit Cost I!stima.
~.,for the Resolution of Generie Safety Issue 106,, D.
Maine Yankee Atomic Power Company, LER tion 309-91-005," Plant Trip on Main Transformer Failure,"
Maine Yankee, October 25,1991.
bemer et al, Janumy 1, M U.S. Nuclear Regulatory Commission, Generic Letter Memerandum from II. S. Heckjord, NRC, to distribution, 89-01, " Implementation of Programmatic Controls for RRS Ufice Letter No. 2, " Procedures for Obtaining Radiological liffluent Technical Specifications and the Regulaton mpact Analysis Review and Support," No-Relocation of Procedural Details of Rl!TS to the Offsite I
vember 18,1983.
Dose Calculation Manual or to the Process Control Pro-
, from li S. Beckjord, NRC to distribution, Rl!S Office Letter No. 3, "Procedur and Guidance for the
,Information Noticc 87-20,"Ilydrogen I cak in Auxil-Resolution of Generic Issues," day 10,1988.
iary Building " April 20,1987.
29 NUREG-1364
i
, information Notice 89-44, " Hydrogen Storage on
-, N UREG/CR-3568, "A llandbook for Value-impact the Roof of the Control Room," April 27,1989.
Assessment," llattelle Memorial Institute, Pacific North-west Laboratory, December 1983.
, Notice of Solicitation of Public Comments on Ge-4 neric issue 23, ' Reactor Coolant Pump Seal Failure,' and
, NUREG/CR-4471, "Los Alamos PWR Decay-Draft Regulatory Guide; issuance, Availability,"lideral lleat-Removal Studies Summary Results and Conclu-Register, Vol. 56, p.16130, April 19,1991, sions," 1.os Alamos Scientific laboratory, March 1986.
, NUREG-75/094,
- Standard Format and Content of
, NUREG/CR-4627, Revision 2 " Generic Cost Esti-Safety Analysis Reports for Nuclear Power Plants," Octo.
mates " Science & Engineering Associates,Inc., Febru-ber 1975.
ary 1989.
, NUREG-0705, " Identification of New Unresolved
, NUREG/CR-4762, " Shutdown Decay Heat Re-Issues in U.S. Commercial Nuclear Power Plants," Feb-moval Analysis of a Westinghouse 'threc-I.oop Pressur-ruary 1981.
ized Water Reactor: Case Study," Sandia National labo-ratories, March 1987.
, NUREG-0800, " Standard Review Plan for the Re-view of Safety Analysis Reports," July 1981.
, NUREG/CR-5072, " Decay iIcat Removal Using Feed and Bleed for U.S. Pressunzed Water Reactors,
, NUREG-0933,"A Prioritization of Generic Safety Idaho National Engineering laboratory, June 1988.
Is ucs," and amendments, initially issued in December
, NUREG/CR-5275," Flame Facility-The Effect of Obstacles and Transverse Venting on Flame Accelera-
, NUREG-1289, " Regulatory and Backfit Analysis:
tion and Transition to Detonation for Hydrogen-Air Mix-t s
M M h M Mumm Unresolved Safety Issuc A-45, Shutdown Decay Acat
}
S*
Removal Requirements," November 1988.
, NUREG/CR-5662, " Hydrogen Combustion, Con-
, NUREG-1370, " Resolution of Unresolved Safety trol, and Value-Impact Analysis for PWR Dry Contain.
Issue A-48,'llydrogen Control Measures and Effects of ments," Brookhaven National laboratory, June 1991.
Hydrogen Burns on Safety Equipment, September 1989'
, NUREG/CR-5759, " Risk Analysis of Highly Com-bustible Gas Storage, Supply and Distribution Systems in
, NUREGICR-2300, "PRA Procedures Guide-A Pressurized Water Reactor Plants," Idaho National Engi-Guide to the Performance of Probabilistic Risk Assess' neering laboratory February 1993.
ments for Nuclear Power Plants," Institute of Electrical and Electronics Engineers, January 1983.
, Regulatory Guide 1.29," Seismic Design Classifica-
, NUREG/CR-2475, " Hydrogen Combustion Char-acteristics Related to Reactor Accidents," Brookhaven
, Safety Goals for the Operation of Nuclear Power National laboratory, July 1983.
Plants," Policy Statement, Federal Register, Vol. 51, p.
-, NUREG/CR-2726, " Light Water Reactor Hydro-gen Manual," Sandia laboratories, September 1983.
, Translation 2240, "The Fire at the Nuclear Power Plant Muchleberg (KKM) Switzerland," December 1989.
, NUREG/CR-3551," Safety Implications Associated With In-Plant Pressurized Gas Storage and Distribution Yankee Atomic Electric Company, " Yankee Nuclear Systems in Nuclear Power Plants," Oak Ridge National Power Station Severe Accident Closure Submittal," De-l Laboratory, May 1985.
cember 1989.
NUREG-1364 30
TAllLES 1
31 NUREG-1364
.- - - -. - --...~ -
l i
).
Table i Number of hydrogen events at each plant location Event location Unburned (reactor-years)
Iteactor Explosions Fires leaks Total Turbine building Il W R, P W R 2
7 7
16 (1424)
Volume control tank in PWR 0
0 11 11 primary auxiliary building (917) liydrogen storage system IlWR, PWR 2
1 0
3 (1424)
Total 4
8 18 30 Sourec: NUREO/CR-5759.
Table 2 Generic plant configurations important accident Core damage scenarios applicable frequency /
Configuration Plant applicability to configuration reactor-year I
All PWRs with auxiliary feedwater T/DHR*
3.4 E-8 (AFW) and vital equipment outside T/LOCA*
- turbine and auxiliary buildings II Habcock & Wilcox and Westinghouse T/D11R 7.3 E-7 plants with AFW system at turbmc (feed and bleed credited) building distribution system Icvel 111 Combustion Engineering plants with AFW T/DilR 9.4 E-6 system at turbine building dis tribution (feed and bleed not credited) system level IV All PWRs with vital equipment at turbine T/LOCA 9.41!-6 building distribution system level
[ station ac blackout or loss of component cooling water (CCW) or service water (SW) system]
V All PWRs with vital equipment at turbine T/LOCA 5.2E-6 building generator floor level (station ac blackout or loss of CCW or SW system)
VI liabcock & Wilcox and Westinghouse T/DilR 2.011-7 plants with AFW system in auxiliary (feed and biced credited) building VII Combustion Engineering plants with AFW T/DilR 4.7E-6 system in auxiliary building (feed and bleed not credited)
VIII All PWRs with vital equipment in auxiliary T/LOCA 4.7E-6 building (loss of CCW or SW system)
- T/DHR = transient-induced loss of decay heat removal.
- T/LOCA = transient-induced loss-of-coolant accident.
Source: NUREG/CR-5759.
NURI!G-1364 32 s
,-,,--a, e
.-v-r-
..,w-..
,-,-.,.+-.-a n
.-. n---
.,----,.--a-n
,-r
.ww-
.~.
=. -
l l
Table 3 Delta core damage frequency per reactor-year for alternatives (calculated mean values)
Configuration Plant Alternalise il 111 IV V
VI Vil VIII A
11 C
Turbine building Alternative 2 7.0 11-7 9.4 E-6 9.4 11-6 NA NA NA NA NA NA NA Alternative 3 7.0 11 - 7 9.4 E-6 9.4 E -6 NA NA NA NA NA NA NA Alternative 4 7.0E-7 9.4 E-6 9.4 E-6 5.2 11-6 NA NA NA NA NA NA Auxiliary building Alternative 5 NA NA NA NA 1.911-7 4.61!-6 4.6 11-6 NA NA NA Alternative S A NA NA NA NA 1.9E-7 4.6E-6 4.6E-6 NA NA NA Alternative 6 NA NA NA NA 1.9E-7 4.6E-6 4.6E-6 NA NA NA Alternative 7 NA NA NA NA 1.9E-7 4.61!-6 4.6E-6 NA NA NA flydrogen storage facility Alternative 8 NA NA NA NA NA NA NA 1.0E-5 5.3 E-5 NA Alternative 9 NA NA NA NA NA NA NA 1.0E-5 NA NA Alternative 10 NA
'NA NA NA NA NA NA
- 6. lE-8 6.9 11-7 2.6E-6 Note: NA = not applicable.
Source: NUllEG/ cit-5759.
a l
33 NUlWG-1364 i
\\
Table 4 Cost of modifications minus onsite averted costs (point estimates) for remaining plant life of 20 years ($)
Configuration Plant Alternative 11 Ill IV V
VI Vil Vill A
15 C
Turbine building Alternative 2 8,400 -130,000 -130,000 NA NA NA NA NA NA' NA Alternative 3 95,000 -48,000 -48,000 NA NA NA NA NA NA NA Alternative 4 1,000,000 950,000 950,000 1,000,000 NA NA NA NA NA NA Auxiliary building Alternative 5 NA NA NA NA 9,200 -63,000 -63,000 NA NA NA Alternative S A NA NA NA NA 76,000 3,600 3,600 NA NA NA Alternative 6 NA NA NA NA 5,700 -67,000 -67,000 NA NA NA Alternative 7 NA NA NA NA 100,000 31,000 31,000 NA NA NA I
Ilydrogen storage j
facility Alternative 8 NA NA NA NA NA NA NA 340,000 -370,000 NA Alternative 9 NA NA NA NA NA NA NA -150,000 NA NA-Alternative 10
-NA NA NA NA NA NA NA 99,000 87,000 57,000 Note: NA = not applicable.
Sources: INI!L,1992, and NURIiG/CR-5759.
l l
I NURIIG-1364 34
.. -. ~ -. -.
-.. _.. -. ~ - -....
l Table 5 Cost of modifications minus onsite averted costs (point estimates) for remaining plant life of 40 cars ($)
3 Configuration Plant Alternative 11 III IV V
VI VII Vlli A
Il C
Turbine building Alternative 2 8,600 -190,000 -190,000 NA NA NA NA NA NA NA Alternative 3 130,000 -61,000 -61,000 NA NA NA NA NA NA NA Alternative 4 1,100,000 890,000 890,000 980,000 NA NA NA NA NA NA Auxiliary building Alternative 5 NA NA NA NA 11,000 -88,000 -88,000 NA NA NA Alternative 5 A NA NA NA NA 78,000 -22,000 -22,000 NA NA NA Alternative 6 NA NA NA NA 4,500 -95,000 -95,000 NA NA NA l
Alternative 7 NA NA NA NA 150,000 47,000 47,000 NA NA NA Ilydrogen storage facility Alternati<e 8 NA NA NA NA NA NA NA 275,000 -690,000 NA Alternative 9 NA NA NA NA NA NA NA -210,000 NA NA-Alternative 10 NA NA NA NA NA NA NA 99,000 84,000 41,000 Note: NA = not applicable.
Sources: INEL,1992, and NUREG/CR-5759, 35 NURLIG-1364
1 Table 6 Cost-benefit uncertainty results for remaining plant life of 20 years With onsite Without onsite averted costs averted costs Percent Percent Percent Percent probability probability probability probability Configu-Alter.
CllR < $1000/
CIIR > $1000/
CllR < $1000/
CIIR > $1000/
ration native person-rem person-rem person-rem person-rem 11 2
35 65 4
96 3
<1
> 99
<1
> 99 4
0 100 0
100 III 2
100 0
44 56 3
97 3
13 87 4
<1
> 99
<1
> 99 IV 2
100 0
44 56 3
97 3
13 87 4
<1
> 99
<1
> 99 j
V 2
NA NA NA NA 3
NA NA NA NA 4
<1
> 99
<1
> 99 VI 5
3 97 2
98 5A
<1
> 99
<1
> 99 6
4 96 2
98 7
<1
> 99
<1
> 99 VII 5
100 0
32 68 5A 88 12 9
91 6
100 0
31 69 7
51 49 6
94 VIII 5
100 0
32 68 5A 88 12 9
91 6
100 0
31 69 7
51 49 6
94 Plant A 8
6 94 2
98.
9 100 0
56 44 10 0
100 0
100 Plant 11 8
97 3
7 93 9
NA NA NA NA 10
<1
> 99
<1
> 99 Plant C 8
NA NA NA NA 9
NA NA NA NA 10 8
92 1
99 Notes: CBR = cost-benefit ratio.
NA = not applicabic.
Sources: INEl,1992, and NUREG/CR-5759.
NUREG-1364 36
Table 7 Cost-benefit uncertainty results for remaining plant life of 40 years 1
With onsite Without onsite averted costs averted costs Percent Percent Percent Percent prnbability probability probability probability i
Configu.
Alter.
CIIR < $1000/
CIlit > $1000/
CllR < $1000!
C11R> $1000/
ration native person-rem person-rem person-rem person-rem II 2
40 60 5
95 3
<1
> 99
<1
> 99 4
<1
> 99
<1
> 99 III 2
100 0
53 47 3
91 9
16 84 4
2 98
<1
> 99 IV 2
100 0
53 47 3
91 9
16 84 4
2 98
<1
>99 V
2 NA NA NA NA 3
NA NA NA NA 4
<1
>99
<1
> 99 VI 5
5 95 3
97 5A
<1
> 99
<1
> 99 6
14 96 4
96 7
<1
> 99
<1
> 99 Vil 5
100 0
39 61 5A 97 3
15 85 6
100 0
44 56 7
44 56 7
93 VIII 5
100 0
39 61 5A 97 3
15 85 6
100 0
44 56 7
44 56 7
93 Plant A 8
21 79 6
94 9
100 0
71 29 10 0
100 0
100 Plant 11 8
>99
<1 11 89 9
NA NA NA NA 10 1
99
<1 99 Plant C 8
NA NA NA NA 9
NA NA NA NA 10 15 85 3
97 Notes: CllR = cost-benefit ratio.
NA = not applicable.
Sources: INiil,1992, and NUREG/CR-5759.
37 NUREG-1364
- ~ _ _.
PEtC fOR M 335 U.6, NUCLEAR DEGULATORY COMMISSION 1, HLPORr NUMutR (2,89)
( Assigned by NHC, Add Vol.,
1 NRCM 1102, Supp., Rov, and Addendum Nurn-
- 21. m BIBLIOGRAPHIC DATA SHEET t*' * - " *nvl (See instructions on the reven,e)
- 2. illLli AND buulH LL
- 3. DATL HLPORT PUUUSHED Regulatory Analysis for the Resolution of Generic Safety Issue 106: Piping MONTH I
vrAn and the Use of 1lighly Combustible Gases in Vital Areas I
June 1993
- 4. rlN OR GRANT NUMBER S Au l HvH t b)
- 6. TYPE OF REPORT Technical C. C. Graves
- 7. PERIOD COVERED (inclusive Dates) 6, PLHf OHMiNO OhGANIZ ATION - NAME AND ADDRESS Of NHC, provide Dmston, Office or hegron, U,S. Nuclear Hegulatory Comenission, and rnaihng address; if cor tractor, provida name and ma,Hng address.)
Division of Safety Issue Resolution Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555 04 WONSOHING OHOANIZA h0N - NAME AND ADOHESS Of NRC. type "Same as atme"; if contractor, provide NHC DMsk>n, Office or Regton, U.S. Nuclear Regulatory Commlnsion, and maihng address )
Same as above.
- 10. SUPPULMLNTARY NO T E S
- 11. ADSTRACT (200 words or less)
IIighly combustible gases such as hydrogen, propane, and acetylene are used at all nuclear power plants. Ilydrogen is of particular imporlance because it is stored in large quantities and is distributed and used continuously in buildings contain-ing safety-related equipment. Iarge hydrogen releases at the hydrogen storage facilities or in these buildings could lead to fires or explosions that might result in loss of safety-related equipment.This report gives the regulatory analysis for the resolution of Generic Safety issue 106," Piping and the Use ofIlighly Combustible Gases in Vital Areas." Scoping analy-ses showed that the risk associated with the storage and distribution of hydrogen for cooling electric generators at boiling-water reactors (DWRs), the off-gas system at IlWRs, the waste gas system at pressurized-water reactors (PWRs), and station battery rooms and portable bottles of combustible gas used for maintenance at PWRs and llWRs is small. On the basis of generic evaluations, the NRC staff has concluded that several possible methods to reduce risk could provide cost-effective safety benefits at some plants. However, in view of the observed large differences in plant. specific charac-teristics affecting the risk associated with the use of hydrogen, and the marginal generic safety benefit that can be i
nchieved in a cost. effective manner, it is recommended that this generic issue be resolved siinply by making these results available in a generic letter. This information may help licensees in their plant evaluations recommended by Generic letter 88-20, Supplement 4, " Individual Plant Examination of External Events for Severe Accident Vulnerabilities,"
June 28,1991.
- 12. KEY WoRDS/DESCRIPTORS (Ust words or phrases that wlil assist researchers in lor:ating the reFort.)
- 13. AVAILADIUTY GTATLMENT Unlimited
- u. mniw CLAssnCATON combustible gases
- Pase) detonation Unclassified explosion (This Report) generic safety issue hydrogen Unclassified
- 4. NUMBLR OF PAGES pipmg probabilistic risk analysis regulatory analysis
- 16. PRICE NRC FORM 335 (2-89)
i 1
i 4
Printed on recycled paper Federal Recycling Program
NUREG-1364 REGULATORY ANALYSIS FOR Tile RESOLUTION OF GENERIC SAFETY ISSUE 106: PIPING JUNE 1993 AND Tile USE OF IIIGIILY CONIllUSTIBLE GASES IN VITAL AREAS UNITED STATES FIRST CLASS Mall NUCLEAR REGULATORY COMMISSION POSTAGE AND FEES PAID WASHINGTON, D.C. 20555-0001 USNRC PERMIT NO. G 57 OFACIAL BUSINESS PEN ALTY FOR PR!VATE USE, $ 300 I
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