ML20096D588

NRR Technical Newsletter.Volume 4,Number 1

ML20096D588
Person / Time
Issue date:	05/31/1992
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-BR-0125, NUREG-BR-0125-V04-N1, NUREG-BR-125, NUREG-BR-125-V4-N1, NUDOCS 9205180137
Download: ML20096D588 (16)

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hN Frank L Witt joined the NRC as a chemical engineer in nants on diesel generator materials. Frank knew about 1976 after having spent 23 years workmg for the General contaminants in plant piping thermal insulation and was Electric Company. Actively involved in the nuclear power the NRR represeruative for revising Regulatory Guide 1.36 fiend at GE, Frank spent considerable time with the and ASTM Standard C-16 on thermal insulation. He had Knowles Atomic Power Dwision. Many of GE's nuclear responsibility for examining microbially influenced corro-chemical engineering systems m use today evolved from sion (MIC) in nuclear power plant water sptems and in Frank's pioneering work.

monitoring methods to avoid MIC. Frank was active in a Frank participated in all of the chemical engineering func-number of technical associations including the National tions at NRC during his too-brief tenure. He headed up Association of Corrosion Engineers, the American Society of Tesong and Materials, and the Electric Power Research the effort on post accident sampling systems (PASS) and Institute lie was also on the ASTM Core Committee for represented the agency at the Tenth Annual PASS Own-Nuclear Grade Coatings.

ers Group Meeting as a keynote speaker. Frank was also responsible for the full system chemical decontamination Frank's friends and colleagues at NRC and throughout the programs at the NRC, coordinating the NRC teams in nuclear power industry will miss him personally and his

- dealing with the Westinghouse Owners Group, the General ever-present contributions to the chemical engineering Electric BWR Owners Group, and the Combustion Engi-field will be missed by engineers who knew him only neering Owners Group. Frank was responsible for intro-through his work ducing hydrogen water chemistry into PWRs ard BWps for control of intergranular stress corrosion cracking Frank leaves his wife, Louise, and three children, who (IGSCC). Frank was also considered to be the NRC ex.

were fortunate to have had a husband and father who pert on primary and secondary water chemistry in PWRs.

made an importam contribution to this country He was the NRC expert on contaminants in diesel genera-tor fuels and hbricants and the effect of these contami-Thomas IL Murley I

9205180137 920531 PDR -NUREC i

DR-0125 R PDR

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of fuel instead of a neutron poison. TI:ey enter the core from the bottom and are inserted to bring the reactor to IN TllIS ISSUE critkal and withdrawn to shut down the reactor. A core thermal fuse is used to hold the lower and upper sections of the core together. Because the core thermal fuse con-Dedication to Frank Witt i

tains fuel with greater uranium density than the rest of the by T. E. Murley..........

1 core, it is the hottest part of the core. %e core thermal An Introduction to Non-Power Reactors fuse is designed so that its polyethylene moderator will

-- by A'exander Adams......

.2 melt if an accident occurs, causing the upper and lower sections of the core to separate and providing a backup Inspections, Tests, Analyses, and Acceptance f

shutdown mechanism.

Criteria 4

(2) Critical Experiment Facility by Tom 'Boyce Decommissioning Regulations for Power One critical experiment facility at the Rensselaer Polytech-Reactors nic Institute (RPI) remains licensed by NRC. This was by Richard F. Dudley..

5 once a very common reactor type, used to acquire the many basic measurements of critical fuel behavior neces-i U.S. Program for Advanced Liquid sary to support nuclear research and power reactor design.

l Metal P.eactor (ALMR) Development It is a very low-power (100 watts) reactor capable of many l

by Stephen P. Sands and core configurations. The reactor core sits in a pool of light

.6 water that acts as coolant, moderator, and reflector. In Geoffrey R. Golub addition to controi mods, the poc! can be q.*!ckly drained Enhancing Safety Using PRA and IPE to remove the moderator from the reactor and shut it down.

Techniques and Results by Dennis F. Kirsch RG V.

8 (3) Argonne Nuclear Assembly for University.

j Training (Argonaut) Reactor (Figure 2)

NEWS 1 E' ITER CONTACT.

Anna May Haycraft, NRR, 504-3075 l

These are low-power (10 kW to 100 kW) reactors with e'

J high-enriched (93% U-235) fuel contained in ahaminum

_w clad Materials Testing Reactor (MTR) type plates. The i

fuel is placed in fuel boxes through which the cooling An Introduction 'I,o Non Power water flows. The water in the fuel boxes and graphite sur-Reactors rounding the fuel boxes act as the moderator. Semaphore-type control rods swing in between the fuel boxes. Massive Alexander Adams, Jr.

concrete blocks weighing several tons shield the reactor and must be unstacked to gain access to the core.

Non Pouer Reactors, Decommissioning and Environmental Projects Directorate (4) Training Reactor. Isotopes Production. General Because of tneir low power levels and inherent safety fea.

_ Atomics (TRIGA) Reactor (Figure 3) tures, the 61 NRC-regulated non-power reactors (NPRs) -

Th.is is the most common NPR design. These are low-to are located in urban areas. In facti the majority of NPRs medium-power reactors (20 kW to 1500 kW) with U-ZrH are located on university campuses and are used for train-fuelmoderator (either 20% or 70% U-235 enriched) m ing and research. NPRs in thia area are located in the fonn of pins clad with aluminum or stainless steet The Gaithersberg at the National Institute of Standards and reactor is mode <ated by water and the hydrogen m the Technology, in Bethesda at the Armed Forces Radiobiol, fuel. TRIGA reactors are either water or graphite re-ogy Research Institute on the grounds of the Navalllospi-flected. %c core sits at the bottom of an open pool and is tal, ir' College Park at the University ci Maryland, and in c oled by natural convection. The reactor can be safcly the D.istrict of Columbia at Catholic University. The power Pulsed to very high power levels for sery short periods by levels of NitC-licensed _ NPRs range from 0.1 watt to 20 pneumatically ejecting a specially designed control rod gw, from the reactor core. TRIGAs have been routinely puhed with reactivity insertions of 5.00$ resulting in pulse power NPR types are primarily distinguished by their fuel design, greater than 4000 MW. The pulse is shut down by a strong The ma}or classes of NPRs are described below, prompt negative fuel temperature coefficient..As the fuel heats up, the hydrogen atoms in the fuel-moderator vi-(1) - Aerojet-General Nucleonics (AGN) Reactor brate and transfer energy to the neutrons in the fuel in-(Figure 1) creasing the probability that they willleave the fuel without The AGN has the simplest design. It is a compact, low.

causing fission to occur. These reactors also use Doppler i

power (0.1 to 5 watts), self-contained homogeneous core broadening of neutron absorption as the fuel heats up to dedgn with low-enriched (20% U-235) powdered uran;um shut down a pulse, oxide fuel embedded in a polyethylene moderator. A graphite neutron reflector and a lead and water shield sur-This is the only U.S. NPR design supported by the manu-1

round the 10-inch diameter core. The control rods consist facturer and a ailable for sale.

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(5) PULSTAR the loss of very expensive heavy water and also prevents the heavy water from at:sorbing light water from the air. In some cases, a decrease in heavy water purity to 90.M will j

The PULSTAR reactor __ frorn American Machine and.

Foundary (AMF) Atomics aho was designed as a pulsing result in a reactor (Mt cannot he _ made criticah IIcavy reactor, 'f hese are medium-to high-power -(1 MW to 2 water is used because it h an ident moderator and has o_

MW) reactors with low-enriched-(4% to 6% U-235) ura.

very "ow neutron absorption cross section. Tins allows the nium-diokide pin fuel clad in zirconium, The fuel ele.-

core to be designed with large gaps between fuel elements.

. ments resemble small pressurhed water reactor fuel ele.

that can accommodate experiments. lleavy water reactork ments.. The two PULSTAR reactors no longer pulse be.

aho have vet v high thermal-todhst-neutron ratios, cause of the cost of maintaining pulse equipment and op-s

- erator proficiency, Like the TKlGA, pulsing was initiated

_by pneumatically ejecting a pulae control rod from the re-ateQpe fool Reactors O,lgum M actor cerem The pulse was strat down by-the Doppler l.

broadening of neutron absorption as the fuel heated up.

These reactors have a wide variety of power levels (0.1 The core ats at the bottorn of an open pool and is cooled

. watt to 10 MW) with plate type aluminum-clad fuch Origi-by forced convectlan.

nauy operated with high-enriched fuel, a number of plate 1-reactors have been converted to low-enriched fuel us part of the Government's program to reduce the amount of

. (6) Tank Reactor (Figure 4) high-enriched uranium at NpRs. The reactor core sits at j.

the bottom of a open pool and is ccMed by forced convec-This is normally a high power (5 MW to 20 MW) reactor tion. These reactors are used for a variety of pmposes, for with plate-type high-enriched (9M U 235) aluminum clad example, beam experirnentation, isotope production, neu-MTR+ type fuel and forced cor. vection cochng. Ileavy tron radiography, and neutron activation analysis. A vari-water may be used as a coolam and moderator or as a ant on this design, the core of the reactor at the University reflector. To gain excess to the reactor core, which is in a of Missouri in Columbia is in a pressure vessel that sits in sealed tank, the tank top must be removed. Although all an open pool. 'The contiot rods are outside the reactor l.

tank reactors do not use heavy water as a coolant, the core and pressure vessel and adjust the amount of neu-I

- tank was an original design feature to keep the heavy trons that reach the external reflector and reflect back water system sealed from the environment. This prevents into the core.

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InspcCtions, Tests, Analyses, status Of ITAAC Resiew and Acceptance Criteria The General Electric Company (GIO Ms been selected to be the lead plant for which to develop the first ITAAC Tom Bo}ce' PDST dunng the design review I r the Advanced Boiling Water Reactor (ADWR). In 1991, the staff held numerous meet-l ings with GE and NUM ARC on this subject, which led to Introduction the deselopmer't of a set of " pilot" 1TAAC which were initially submitted for the staff to renew in September The Standardiranon branch of the Division of Advanced 1991. A full ITAAC submittal is espected in early 1992 Reactors and Special Projects has been coordmating the after agreement is reached on the P ot ITAAC. Staff ap-il

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Office of Nuclear Reactor Regulation's (NRR's) review of proval of the fullITAAC submittal o scheduled for the f all several new reactor designs which are intended to be h-of 1992, censed under the new licensmg process for a facihty estab-lished under 10 CFR Pact $2. These new designs, together Types Of ITA AC with Part 52, will provide the foundation for the next gen-GE is deseloping sesetal types of ITAAC for the GE eration of reactors to be hcensed in the Umted States. The ADWR. ITAAC are being developed for the approxi-inspections, tests, analyses, and acceptance criteria inately 140 systems of the ABWR desyn. Examples of (ITAAC) hase emergcd as a distinctise part of the design these'" system" 1TAAC are the ITA AC requirements for reviews because of the unique role which they play in this sy3tems such at the residual heat removal sistem and the new licensmg process. The concept of ITAAC ts also new, reactor protection system. GE also is developing "genenc" and thus has been a topic of considerable interest to in-ITAAC for issues that apply to many systems, such as dustry, the Commission, and the staf f during the course of those for the ennronmental quahfication of key compo-the design reviews.

nents. GE has identihed seseral areas as design interfaces with the rest of the facihty. These include site-specific ele-L nder the 10 CFR Part 52 "ene-step" licensing process, ments such as the semce water intake structure and the the NRC must be able to mak e all safet) fmdmps for a ultimate heat smk. A utihty referencmg the design wdl be facihty before the first showl breaks ground for construc-required to meet these interface requirements by subrmt-tion. This is no small challenge. In contrast, the plants ting "fatihty" ITA AC that wdl be prouded in the apphca-currently operaung were licensed under the "two-step" h.

tion for a COL censmg process under 10 CFR Part 50, whereby the NRC would first grant a utility a construction permit to budd a GE prouded dengn informanon hasing a lesel of detail Wat w s msu c rm ng a f facility and then issue it a second hcense to operate the ti n in certam are s of the reuew. gal safety determina-facihts after it was built. Under the Part 52 process, the GE did not proude a NRC would issue a single combined hcense (COL) for a high lesel of detad mihese areas because they are areas of facdity to be both budt and operated. Iloweser. to conbrm rapW chan@g Mn&g or because they represent as-that the facihty was built satisfactonly pnor to its actual budt or as procured type informauon, l'hese areas include operation, the concept of ITAAC was developed ITAAC instrumentation and controls, piping design, radiation pro-consist of various inspections, tests, and analyses which tection, and control room design GE is deselopmg design

.he uuhty performs for significant aspects of the facihty, accept nce critena (DAL) 3pecifymg the requirements wa muu met M umn' deQn umk in thew areas md then measures the results against prescribed accep-tance criteria. For example, if the flow rate of a pump has Resiew issues particular safety significance for the design, the appbcant would perform various mspections, tests, or analyses and 1.

I.esel of Detail And Extent of Standardization. The compare the resuhs to acceptance enteria in order to ser-level of detail in the certified design and the ITAAC ify the flow rate. Prior to fuel loading, the ITAAC provide ud! greativ af f ect the extent of standardizauon in fa-tne basis for the Comm:suon to confirm that the plant was ciht es referencmg the ceruhed deugn During the re-i built and wdl operate in accordance with the approsed de.

new proces< a graded approach to the cerufied de-sijn in the COL ugn informanon based on its safets sigmficance has been taken to proude the appropnate safety benehts of standardvanan Part 52 Design Resiew And Certification Process 2.

Separation of Tier 1 and Tier 2 Information. The Under Part 52, after the staff performed an extenme re-rule certifung a desgn will contain a sufficient lesel of view of a design and : Jued a fmal safety evaluation report, design detad so that the ru!e prosides for a standard-the Commission would isstm a final design approsal 17ed design and proudes for early resoluuon of design 8

(FDA) for the design. The Commission wotud then mAe issues, while allowmg the fleubihts to accommodate a rf : certifying the design and the ITAAC for the design necessary changes to a facday dunng its construcuon in order to provide fm aso!unon of all des.;n mues, and operaung hh ume. Thus, the rule w di only ceruf>

makmg them no b.iger subject to liugation. The ceruhed a selected poruon of the informanon (cahed Tvr I dmign can the* ce referenced by an apphcant for a Col..

informanon) <uhnutted by the deugner Tlus portion Sine'. the des!gn n not subject to hupation, this w dl of the deugn mformanon would be senhed by the stn amhne the process for hcensing new facihties In addi-11 A AC and would hase ury high thresholds for tio;, when referenced by multiple hcen3ees, a ceruhed de-thanges The rematnder M the deugn mformat on, sign would become the standardned ursign f or f uture nu typicalls that contained m the standard safets analy us clear plants.

iepon (SS AR L would he coritndled h a " 5u.54hke" 9

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t process, and would include a requirement to ensute.

heahh and safety, The NRC conducted technical analysis

- that the snange cud r,ct affect any Tier 1 information.

' to support the rule and concluded that three basic meth.

1 ods of decommissioning were acceptable:

4 T Overlap With Traditional Inspection Process Un-c der Part 50, Some areas of ITAAC may overlap with

. DECON:

Promptly decontaminate and I

the' existing regulations of 10 CFR Part 50 and the dismantle the facility, traditional inspection process conducted while the fa-cility is being constructed. The ITAAC are not in.

SAFSTOR: place the facility in an isolated,

. tended to supersede or substitute for existing regula-safe storage condition allowing tions and. inspections, but instead will provide a com-radioactivity to decay before plimentary process by which to verify that the facility dismantlement.

will conforrn with the certified design. Thus, all re-i ENTOMBt Entomb the facility by encasing

- quirements such as those found in Quaht) Assurance r di active materials in structurally Programs derived from Part 50, Appendix B, will still long-lived concrete or other material be in effect, inspection activities will be coordinated in and store until decay allows release a manner similar to that conducted in the readiness for unrestricted access, review pilot program for Vogtle.

The time limitations for completing decommissioning may =

Su, mary not allow for the ENTOMB alternative when long-lived The designers are developing the first ITAAC for the lead radionuclides are present. A power reactor which has op-

. designs being considered for standardized desigr. certifica-erated for 40 years will hkely have inventories of tion under 10 CFR Part 52. The ITAAC provide the basis nickel 59, nickel-63, and niobium-94 (half lives of 80,000 to confirm that key aspects of a facility licensed under Part years,92 yea % and 20,000 years, respectively) inside the 52 are t.iih and will operate in accordance with the reactor vessel, which would normally preclude tising the

. Atomic Energy Act-and the Commission's regulations.

ENTOMB alternative.

Since the ITAAC implement many issues associated with standardized reactor designs, they will play a significant Financial Assurance role in the licensing process of the next generation of nu-The NrtC required each power reactor hcensee to submit a clear power plants m the United States.

report by July 26,1990, to inaicate the manner in which it t

would ensure that funds would be available to decommis-DCCOninliSSIOntng RegillatlOnS for sion the facility, Acceptable funding methods allowed by Power Reactors the regulations include making a prepayment, establishing an external sinking fund, or obtaining a surety method Richard F. Dudley, PDNP (bond, letter of credit, insurance, or other guarantee).

Nearly all power reactor licensees have chosen to establish Introduction external sinking funds into which periodic deposits are made so that upon expiration of the facility license, suffi-On June 27, 1988, the NRC issued a package of revised cient funds will be available to pay all decornmissioning and new regulations to ensure the safe and effective de-costs.

commissioning of nuclear facilities. These regulations be-

- came effective 30 days later. However, the NRC gave li' The segulations specify a minimum amount of funding that censees until July 26, 1990, to submit reports indicating must be ensured [$10$M for large PWRs end $135M for the manner in which they will comply with the require-large BWRs (1986 ciollars)] and also provide a formula for ments for ensuring funding for. decommissioning. When annually adjustmg these amounts to account for the esca-the NRC issued these final decommissioning regulations in lating costs of labor, energy, and waste disposal. The regu-

1988, it concluded 10 years of performing technical, emi' lations also allow the licensee to use a site-specific cost ronmental, pohey, and legal analyses. The NRC also in-estimate if it is not less than the minimum value. These cluded in the rule its review and responses to public com-decommissioning costs do not include the costs necessary ments received on proposed regulations published in Feb-

.to store and dispose of spent fuel.

ruary 1985. Although these regulations apply to all NRC-regulated nuclear facilities, this article will only address Planning for Decommissioning

. their effects on nuclear power reactors.

The regulations in 10 CFR Part 50.75 require each licen-see to submit a preliminary decommissioning plan on or g

Acceptable Decommissioning Alternatives abut 5 years before the projected end of plant operation.

The regulations define " decommissioning" as safely re-ne preliminary plan is to include a site-specific cost esti-i moving a facility from service followed by reducing resid-mate, the decommissioning alternative anticipated to be l

ual radioactivity to a level that permits the releasc of the.

used, major technical actions necessary to decommission property for unrestricted use. They do not require the re-the facility, the current situation with regard to disposal of moval of non-radioactive structures or structures that have high-level and low-level radioactive waste, the criter.ia for I

been decontaminated to levels acceptable for unrestricted residual radioactivity, and any other site-specific factors use, The regulations spedfy that decommissioning must be that could affect planning and cost. If necessary, this sub-accomphshed within 60 years of the time when the plant is uuttal must also include plans for adjusting the level of Jshut down, ehhough a longer period may be allowed mder funds ensured to be available for decomnussioning. Al-certain circumstances if necessary to protect the public though the plan must contain sufficient detail to justify the 10 r

l 1.

l I

site-specific cost esumate, the prehminary decommission-nearly completed the process to approve the decomnus-ing plans need isot be approsed by the NRC, sioning plan for Shoreham.

In 10 CFR Part 50.52, " Termination of License," the Additional Rulemaking Needed NRC requires the licensee to submit an apphcation to sur-The staff's recent hcensing elforts on plants in the decom-render the beense within 2 years of shutdown and not later missioning process have shown that when man) existing than 1 year before the operating license expires. The ap-NRC regulations were promulgated, their applicabiht) to plic,uon to surrender the bcense must be accompanied by decommissioning was not always evaluated by the staff.

or preceded by a proposed occcmmasioning plan. This Thus, the statt has issued nm merous exemptions since plan must include the decommissionmg alternathe se-strict comphance with certain regulations is not necessary lected, a description of the controls to be used to protect to ensure adequate safety at decommissioning plants. NRR the health and safety of the public, a description of the has recently requested that the Office of Nuclear Regula-final radiation survey, an updated cost esumate and,

tory Research iniuate a rulemaking to consider moditymg comparison with the amount of funds alre4 collected, and clanf>tng requirements for decommiwomng plants in and a desenpuon of the technical specificathans, quaht) the areas U 10 CFR Part 50.59 apphcabihty, possession-i assurance provisions, and physical security plan prousions only bcense nsuance, emergency preparedness, secunty proposed for decomnussioning. After the NRC reviews and sa feguards, propert) damage habihty msurance, and approses this decommissiomng plan, the staff will n requirements for operator traimng and t equahtication and sue a decommissionmg order which amends the beense to for simulators. fitness-for duty, 10 CFR Part 50 Appenda estabhsh the specific regulatory requirements for decom-J leak testmg, maintenance, and hre protecuon.

missioning at the particular f acihty.

U.S. Program For Advanced Limid Radioiogical Release Criteria Metal Reactor (ALMR) 1)evelop% ot

~'he regulations do not include radiological release cruena Stephen P. Sands, PDAll, and smce final entena are to be deseloped by the Enuron-(leoffrey 11. (loluk PI)All memal Protection Agency (EPA). Until the EPA develops these criteria, the NRC wdl continue to use the surface Within the framework of the Advanced Liquid Metal Re-contamination hmits iti Regulatory Guide 1.56, "Termma-acmr (ALMR) Program. the U.S. Depanment of Energy uan of Operating Lttenses," 19R Smce IW1 for Part (DOE) selected the Power Reactor Innosatise Small Mod-50 reactor facibues, the NRC has also used as the hmit for u!e (PRISM) desyn as the ha nd metal reactor deugn to Itxed contaminanon by gamma-emitting radionuchdes. a sponsor for NRC design ceruncation. The conceptual value of 5 micro-Roentgen per hour above background at PRISM design was developed by Gencral Flectne (GE) a distance of 1 meter, Occasionally, the NRC pernuts li-Compan, m conjuncuon with an mdustnal team compris-ctmees to exceed the 5 micro-Roentgen per hour hmit if it mg Bechtel Power Corporanon, Borg-Warner Corpora-can be shown usiug reasonable occupancy assurnpuens tion, Fouer Wheeler Corporauon. and Cruted Engineers that the maximum dose commitment to an indindual and Constructory Inc. Research and deselopment support would not exceed 10 mrem per year. Recently, m a June n being supphed by the Argonne Nauonal Laboratory, En-2S,1991, stati requirements memorandum, the Comma erg) T echnok,gy f:ngmeennp Center, Hanford Engmeer-sion directed the paff of both NRR and NMSS to conunue me Des elopment Laboraton, and Oak Ridee Nauonal to use these exisung critena (which were m place just pnar LEborator). In addiuon, a steermg group of utihty repre-to the July 3,199 L pubhcation of the Below Regulator) sentauses was invuhed m the PRISM design ef fon.

Concern Pohey Statement), in making hcensmg decisions involving decatr.missioning.

DOE chose to spomor the PRISM deugn as part of its Nanonal Energy Strategy because of the design's potential Prematurely Shutdown Plants for enhanced safety through the use of passne safety sys-tems and greater safety rna r gins, reduced cost through The decommnsioning regulations descnbed herein do not modeint devin and construcuon, and powible future con.

specificalh address plants that prematurely shut down be-tnbuuan to high-!evel waste management through dewtop-y fore their operatmg beenses expire. The NRC found ses-ment of an actimde rec >chng capabm3. Ahhough this last

\\

eral concerns after the Fort St. Vrain, Rancho Seco, and ahernauve has not yet been proposed m the current apph-Shoreham plants aH shut dowri prematurely and entered canon, acumdes separated from light-water reactor spent the decommtssoning process. The most ugmhcant of fuel could be burned in an ALMR fast flux core. This these concerns were (1) the nature and extent of the NRC process oHers the potenual to reduce the hich-level waste review undet the National Enuronmental Pobty Act (llLW) storage requirements for hght-wa:er-reactor spent (NEPA), specifically whether the ernironmental impacts fuel f rom thousands to hundreds of sears.

of replacement power alternatnes must he compared with the environmemal impacts of returning a shutdown plant The NRC is reuewmg the PRISM conceptual deugn and to power operation, (2) the time penod for accumulating wiH nsue a fmal Preapphcauon Safen Evaluanon Report the fundi far decommusionmp and (31 the requirement (PSERI. In order to obt.un N RC apprmal of its prototspe, to pay annual fees (10 CFR Part 171)= Oser the past see DOE plans to subnut a Prehminary Safety Av essment Rc-eral years, the staff has worked unh these hcensees and port (PS AR) m 1945 and a i mal Safen Msessment Re-with the CommWon to resohe al.agnibcant uncertamues port in A W DOE aho plans to apph for standard deugn in these areas. 't he staf f is now conducune decomnussion erinhc *un i

20n3 nuer a proun pe demonstranon mp renews to thexe prematurch shutdown plaras and has

~l ni. se phin, a re hated i m tlm current DOL coah to dem-

!I

f5 onstrate the commercial potential for the ALMR by 2010, (IllTS), lilts sodium is circulated by a centrifugal pump.

as called for in the National Energy Strategy.

The !!!TS operates at a higher pressure than the primary loop so that, if the IHX breaks, the sodium would not flow

- The PRISM Plant Design '

out of the reactor vessel. The !?ITS transfers heat to the SO system, which provides saturated steam at 965 psi to The PRIShl plant design consists of three separate power the turbines, A sodium-water reaction protection system blocks each rnade up of three reactor rnodules (Figure 1).

m gW Weca I rea ns ween s dium and Each module has a theimal output of 471 MWt and an water in the SG.

electric output of 155 MWe for a total (plant) output of 1395 N1We. Options for one or two power blocks are also Core and fuel possible; The PRISM design contains three turbines, each

_ supplied from a power block, The nuclear steam supply The reference fuel for the ALMR is a uranium-pluto-sy en (NSSS) for PRISM consists of the primary sodium -

nium-zirconium (U-Pu-Zr) alloy. The ferritic alloy llT9 is loop and the secondary (intermediate) ' sodium loop, used for dadding and channels to minimize swelling l

which receives heat from the primary system and transfers caused by long bumups. The PRISM core is a heterogne-it to the steam generator. The steam generator is the inter-ous arrangement of driver fuel and blankets The PRISM face for sodium and water systems.

design has six control rods. The refueling schedule for PRISM calls for replacing one-third of the core every 18 Reactor Module months. The f ael designers cite neptive reactivity feed-backs, better heat transfer properties, and competitive The reactor module consists of the containment system, c sts as advantages of metal luel, llowever, a drawback is the reactor sessel, the core, and the reactor's internal that the metal fuel has a considerably lower melting point components. The reactor vessel encloses and supports the Wan oxide fuel. Also, the neutronic design. employed to core, the primary sodium coolant system, and the internal achtese the negative feedbacks necessary for a highly de-components. The sessel is located just inside the contain-arable passive shutdown feature results in an undesirable ment vessel, which is located below grade in the reactor p sitive void coefficient. Houver, the negative tempera-silo. The reactor vessel is made of 2-inch-thick, type 316 ture coefficient and operation at temperatures well below stainless steel. The reactos vessel is penetrated only in the s dium boiling are intended to make core vaiding a highly closure head, it is supported by the floor structure, and unMely event. The Argonne N,auonal Laboratory (ANL) the floor structure is supported by seismic isolator bearings ts cont nuing t develop the metal fuel as part of the Inte-to reduce horizontal movement during seismic events.

gral Fas' Reactor (IFR) Fuel Program for the PRISM de-M "'

- The' reactor core is supported by a beam structure at the bottom of the reactor vessel..The reactor vesse! also con-Reacthity C ntr I and Shutdown Systems tains support for storing up to 30 spent frel and blanket assembbes. The upper head of the rearcor vessel is the There are six control rods in the main reactivity control closure head, consisting of 12-inch-thick, type 304 Ma6 and shutdown system. Inserting any one of the six will shut J less steel. It would assist in mitigating (Le effects of hypo-the core down. The control rods can be inserted using (1) thetical core disruptive accidents (HCDA). The closure the plam control system (PCS), for normal insertion, (2) head also supports the intermediate heat exchangers the safemgrade reactor protection system (RpS) for rapid

(llIXs) and the electromagnetic (EM) pump % The reac-insertion, and (3) gravity drop into the core. If this system L

- tor vessel is about 62 feet high and just under 20 feet in fails, the operator can send boron balls into the central diameter.

location of the core which causes shutdown independently c

of the control rods. The PRISM design also includes pas-

. Nuclear Steam Supply System (NSSS) sive mechanisms for controlling reacuvity: three gas ex-pansion modules (GEMS) consistiry, " tubes, closed at the The main components of the NSSS in i,RISM are the re-g g

g g

gg actor module, pnmary sodium loop. EM pumps, IHL in-pumps ate running, the static pressure is high, causing the termediate sodium Icop and steam generators (SGs).

level to use to a high point in the GEM. Ilowever.

so mth the pumps o.ff, the static pressure and sodium level The sodium in the pnmary loop circulates from the core drop, which lnereases neutron leakage. The reactivity' outlet to the shell side of the IHX.to the EM pump and change kom the GEMS between these two states is about l-then to the. core inlet This primary sodium loop is con-L tained completely within the reactor vessel, which is tr-

- metically sealed to prevem leakage of the primary coolant.

f<esidual lleat Removal The high saturation temperature of sodium allows large margins to voiding and !ow system pressure during normal Normal cochng through the non-safety-grade condenser is operation. The EM pumos provide the primary sodium used for residual heat removal (RHR). If the condenser -

circulation. Conventional pump coastdown is not possible becomes unavailable, the safety-grade reactor vessel auxil-because the EM pumps have no moving parts. Ho vever, a iary cooling system (RVACS) is used for - RilR. The synchronous machine pmvides flow coastdown through RVACS operates through the direct natural circulation air th; EM pumps. Flow coastdown is very importart for pre-cochng of the containment vezel. The design-basis acci-venting sodium voiding during a loss of power wahout re-dent invokes a loss of all RilR except the RVACS Analy-actor scram, sis has shown that the RVACS' heat removal rate is suthcient to maintain the temperatures of the internal l

Reactor-generated heat m the pomary loop is transferred structures under American Socie,v of Mechanical Engi-through tae lilX to the intermediate heat transfer system neers (ASME) Level C conditions (1200xFL The PRISM l

12

~.

design also contains the non-safety-grade auxiliary cooling number of concerns and open iwues that needed to be syrtem (ACS) to assist the RVACS, The ACS uses natural resched before the design could be approved.

t circulation of the steam generator to remove heat indi-rectly from the reactor vessel. The ACS on be used in in May 1990, DOE submitted Amendments 12 and 13 to combination with the RVACS to reduce the tooldown the PSID to r ddress the issues and merns identified in time.

the draft PSER. Major design ci,anges were made, includ-ing the following:

The frequency,inagnitude, and duration of high tempera-ddmg a low-leakage, pressuresretrining containment tures reached during RVACS accident scenarios may be of concern to safety, Therefore, researchers will examine the dome above the sessel head effect of these temperatures on the life of internal compo-

• sueng ning the core support and sessel head pene-nents in the PRISM prototype reactor. However, the sim-plicity of the RVACS will make it difficult to defeat. The tration structures to better contain postulated severe RVACS has two major classes of failure modes; the deg-CDdi'f"PLIV* **'"I" radation of RVACS suifaces to lower heat transfer capa-bihties and the blockage of RVACS flow passages. For adding a diverse reactivity shutdown system, using bo-6 ron balls at the center of the core flow blockage scenarios, the duration of the flow blockage will determine the outcome of these events.

demonstrating a seismic-capability of 0.5g peak ground acceleration abuse the design-basis ground ac-Containment celeration of OJg adding three gas expansion mod-The containment is unconventional, consisting of the con.

ules (GEMS) at the core periphery for increased reac-tainment vessel (guard vessel) and a containment dome tivay margin during loss-of-flow esents j.

above the reactor closure head. The containment systern The PRISM design.is currently one of four preapplication is a low-leakage, pressure-retaining boundary separate l

from the reactor vessel. The containment vessel is located reviews being conducted by the Advanced Reactors Pro-just outboard of the reactor vessel and has no penetra-ject Directorate in NRR. The objectises of the preapplica-tions, A break in the reactor vessel wall will allow sodium ti n reviews are nientified m SECY-4!-202 as foHows:

to spill into the containment vessel, but the core will still l

remam covered. The atmosphere in the reactor vessel Identify major issues that could require the Commis-above the so-c'ium pool is hehum. Ilowever, the atmos-sion ta provide policy guidance before the staff it.iti-ates acuons.

phere in the containment dome is normal air. The de-signer performed a preliminary analysis to determine ra' identifv major technical issues that the staff could re-e dicactive releases for HCDAs witt a simultaneous break in sohe in the context of existing ryulations and Com-l the reactor vessel, However, the source term for the nussion pokcy, and for which additional Commission t

PRISM design remains an open issue.

guidance is not considered necessary.

NRC Activities

• Identify research and developmem that is needed to On July 8, '1986, the Commission pubhshed " Regulation remlw noted issues.

of Advanced Nuclear power Plants, Statement of Policy" (51 FR 24643), and in June 1988, the NRC issued "De-Key technical and policy issues being considered during velopment and Utilization of the NRC Pohey Statement on the preapplication review include the following:

the Regulation of Advanced Nuclear Power Plants? The Metallic Fuel Performance. The proposed reference Commi>sion encouraged the NRC staff to interact early fuel design is a U-Pu-Zr metanic fuel with steel allay l

with designers of advanced reactors to estabhsh licensing IIT9 cladding. The staff needs more information on guidance that applied to such designs. In accord with the the phase change ei high temperature, the fuel extru-policy statement, the staff would review conceptual designs sion durine a significant overpower event, and the l'

before receiving any formal application for a construction eutectic fo'rmation and interaction with the cladding.

permit or standard plant review and certihcation.

ANL is ruanaging the DOE fuel research program.

which is still in the development stages and requires in November 1986, the NRC received the PRISM Prehmi' close monitarmg.

f nary Safety Information Document (PSID) for this preap-l plication review. The Office of Nuclear Regulatory Re-Positise Void Coefficient. The stoposed core design search ;RES) reviewed the design in accordance with the has a positive void coefficient tqat could result in a guidance of NUREG-1226, " Development and Utilization large positive reactivity addition should sodium boiling 2

- of the NRC Policy Statament on the Regulation of Ad*

occur in the center of the core. Redundant and di-

"anced Nuclear Power Plants," and a draft Preapplication verse coolant flow pumps and GEMS were designed to Safety Evaluation Report (PSER) was issued to DOE in deal with this problem The NRC staff is evaluating the l

September 1989 after it was reviewed by the Advisory preventive and mitigative features of the PRISM de-l Committee on Reactor Safeguards (ACRS), although the sign to assess the safety significance of this issue-Commission did not give its formal approval of the draft PSER. The NRC staff and the ACRS concluded that the liypothetical Core Disruptive A'ccident (llCDAn l.

PRISM design provided several features for making a nu-Potential HCDA initiators include a latge unprotected clear power plant safer and that design and development reacovay insertion from core saidmg. The proposed should be cc,ntinued The draft PSER also identified a reactor vesse and containment are ead designed to I?

i

l l

contain the postulated ilCDA energy without breach-llowever, this sptem alone could lease the plant at ing. The stati is tsaluating the magnitude of the me-elesated reactor temperatures for long periods after chanical energy released from an llCDA and assessing accideatt The staff is esaluaung the acceptability of the intepity of the primary spt'm, containment, reat-highly rehable, actis e, non-safety-grade, heat-removal tor vessel, and vessel head.

sptems to redure the frequency, magnitude, and du-ration of challenges to the passive sptem.

o Source Term. The preappbcant has proposed usmg a source term unhke that used for current light water Seismic isolation. The proposed design uses large e

reactors to account for the liquid metal reactor design.

seismic isolators to support the nucicar island to re-Release estimates from the metalhc fuel and decon-duce the magmtude of horirontal ground acceleration tamination factors for the r. odium pool and coser pas transmined to the safety-pade nuclear island struc-under the containment dome need to be supported b) tures, systems, and components. The staf f will con-experimental data.

tmue to evaluate the isolator system as additional de-gn mformauon and test resubs become available.

o Operator Staffing and Control Room Design. The proposed control room utilizes advanced instrumenta-Emergency Planning. The preapphcant has postu-uen and control sptems to support nine reactor mod.

lated a sery low probabihty of exceeding the Protec-utes from one room with as few as five hcensed opera, the Action Guidehne lower hmits at the site boundar) tors. The staff needs to esa)uate whether the propowd and has proposed reduced requiremenn for ofSite design and staf'ing can effectisely manage muhiple C'"Uff e ncy planning. 'T his is a pahey issue that the plants simulu.n ausly during accident, transient, arid Maff wiH raise to the Commission for further guidance shutdown situations.

before evaluating emergency plannir g for the PRIS$1 o Passise Residual lleat Removal System. The pro-posed design utih zes natural consecove air flou ateund the lower enntainment vessel through a hot '.nr The staff is resiewing the new design and evaluatmg key riser space. The sysAm is completely pesne and in mue and will mcorporate the results of its evaluation in a continuous operation, providmg increased reliability.

final PSER espected to be completed in Nosember 1992.

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Enhtmcing Safety Using Pita and IPE with the need for plant saf ety >> stems to be rehable and Tecliril(Ities and IlestiltS oper Ne and the need to periodically test and maintain the sptems. PR \\ techniques have been used in revising l'y Denttis l', Kirsch, Regi, a T

• Ch"IC I *PiIi'"li""5 I"' """ti " I"$t f"**"l^ t i"" ""d the staff expects other similar apphcations to increase in Introduction e huure.

The nuclear utility industry is performing indisidual plant Applicability To Daily Operational Decisions examinations (IPEs) using probabilistic risk assessment The use of the IPE methodology and results could contrib-(PRA) techniques, to quantify plan' nsk resulung f rom se-ute significanth to the daily planrung and scheduling of quences that can initiate several sesere accident events.

test and maintenance activities.

PRA techniques ena,le plant management to deterrnine how existing and proposed equipment and the procedures if au syMems are in seruce and are operable, the risk, or and practices for cperation and maintenance affect the c re damage frequency (CDF) (esents per year), is mini-overall plant risk. lhe licensee can use the results of the mah H wever, na incre m when eqmpment and systems IPE process to assess the effects on eperational safety of roust be remosed from op-rabihty status dunng pl.ait op-activities in seseral broad areas. For example, the IPE re-eranon for test nd maintenance acuutiet The 4ctual

.ults can be apphed to qualitatively assess the effects that magmtude by which nsk increases depends upon the rela-scheduled maintenance and testing base on aserage plant use unp nance f We component or sptem in the event risk, assess the effects of design modificatiom on plant tree and a duectly proporuonal to the out-of selvice ume.

risk, and optimize the technical specificaucus to mmimue Rnk als increases due to muluple component or system plant risk.

inoperabihues. These f actors cause nsk to sary continu-

)

ously throughout the op+ratmp cycle.

The results of calculations indicate that, whik plant safet)

To determme if the remosal of a safety related component systems are maintained or tested orlme, the plant's rnk of mcreases risk unacceptably, the hcensee should estabhsh core damage increms in a manner directly proportional an upper hm:t nsk guideline. Iloweser most hcensees base to the amount of time the system is unavailable. For thn not set done tho.

reason, the benefits of voluntary onhne tests and mainte-

~

nance shcald be weigned against the risk mcurred PRAs can be mo Shed to rdlect actual plant equ pment conheuration used to determine if or when to remove Outage Safety Assessments equipment trom seruce for testing and maintenance. For While IPEs and most other PRAs base only conuderea.

exarnple, certam suneillance t : sting, required by technical specihcauons, could be umed so that the necessary system plant configurations typtcal of normal full-power opera' tion, PRA techmques can be adapted to proude mughts ou'c comiNm a Mle as pmuble to the total rak in about nsk for configuration other than full power. Inmal ombinanon w:th all other equipment in test or mamte-nm u The mom progressne unhoes are consider-work in this area has been summarued in " Shutdown and mg a methodolon W awng the adusabihts of schedul-Low-Power Operanon at Commercial Nuclear Powe r Plants in the United States " NUREG-1449 (DRAFT RE-ine and performmg needed test and maintenance acuuties duch conuders the status of all safets systemt PORT), February 1992. During outages, the hcensee can

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utilize PRA techniques to identify and quahtausel> address In scheduhng suneillance testmp and mamtenance on accident scenarios that are of high relauve probab:hty and safety-related spiems to take place concunently (for ex-can resuh in sesere consequences. For example, to evalu-ample, during the day shdt when the largest ru aer of ate the special risk constderations associated with PWR maintenance personnel are onste), the licensee should mid loop operation, the hcensee can use PRA technique 3 conuder the eflects of those actiaties on the ability of the en quahtausely identdy and evaluate the relatise effect en plant to mingate events. Even though a strict interpreta-total risk resuhing from remonng equipment or systems tion of technical specifications may allow such a practice, from service during mid-loop operauon or from enter;ng this is not a conservanve operaung philosophy. N!ost uub-mid-loop operation with certain combinations et eqmp-ties understand the risks of that kmd of approach, llow-ment imavhilable. The licensee should carefully esaluate eser, the industry has not fulh devloped an effectne the advantages and dnadvantages of performing soluntar) methodology for scheduhng test and mamtenance acuvi-online tests and maintenance of components or sptems ties to mimmue nst or mamtam rnk within r ecom-that can affect nsk dunng mid-loop operauon.

mended guidelines. In presennse maintenance, recent in-terpretauon of the inspecuan N1anual, addressed n'un-To perform realistic outage safety assessments, the hcen-tary entry mto htmting condmons of operation (LCr ; to see can use risk-based failure modes and effects analysis perform presentne rnaintenance. Accordmg to thn mter-(FN1EA) of shutdown cond tions and the effects of special pretation, p esentne mamtenance ( PN1) procedures on plant conditions on the nsks attendant with those condi-equipment that h perfonned onhne should improve safety tions. RISK-BASED TECilNICAl. SPECIFICATIONS (that is, reduce nsk or improse rehabaht3 h and should not he performed as a matter of con emence. Thn interpreta-To mir.imize the nsk of potennal accident sequences at unn implies that unenhghtened con [ormance to PNI pro-the plants, PRA techruques and IPE results can be used to gram scheduhne and reqmrements is short-signted design the allowed outace omes and survedlance time m-tervals m the technical sIpecthcanons to reduce core dam-Mant I Aperience age frequencies. Tlus apphcation of rnk as*essment woulJ Repor< \\ has recenth esanuned the mdustr> N expenence prmide a logical process h> which to balance plant Jen that correLies the eutet N suem outage nme f resolong j

1 15

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frorn onlir e test or rnanntenahce) to the tracreate in risk ci The magnhude of a CDP inurase for any system.

(expressed an events per reactor 3 cat), llegion V exam.

assuming a ltercent unavailabihty. varies widely I

ined the three Safety sptems for which utilities teport un.

arnont plants (from a traction of a percent tu about availability data to the Institute of Nuclear Power Open.-

20 pr.rcerit) tions (INPO)1 high pressure safety injection, emergency diesel r,eneration, and auxiliary feedwater.1he region d.

The magnitude of CDP increases throughout the year used a baseline CDP for the $ptem. assurning no unasail.

for each plant with the number of hours the safety ability.1he region compared the baseline CDP to CDI?$

sptem is unacallable; fer a 1-percent sptem unavailabihty and the sptem un-Conclusion availabihty as tcported to INPO.

The industry is establishing a technically defensible ructh-1he study concluded the following:

ndology to n.inimlic risk. i.nprove reliability, and hnprove safety in the selection and implementation of surveillance i

The hccusees for plants examined had riot calculated ung, co$ lectin rnaldenance, and preventin mabte-

"# "" "II rema ns to be @ne in Ms a.

t b risk for cote damage based upon the actual area. Bus eHon can gready u.nprow meran plant ufety.

drnber of hours of unavailabthty. Accordingly. man-agement was not av<are of the true integrated risk.

ypg mt,thodology and results can be applied to a wide range of bcensee organitations and activit:es. The chal-b.

Jach plant has one sptem, or component, that lenge for the nucicht industry is to devise technically de-contributes the most to CDF increase for each I fensible mechanisms to use IpE results to effect real and percent of unavailabihty, rueusurable improvernents ir. plant safety.

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