IAEA-TECDOC-938
Predicted radionuclide release from marine reactors dumped in the Kara Sea
ReportSourcethe of Term Working Groupthe of International Arctic Seas Assessment Project (IASAP)
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PREDICTED RADIONUCLIDE RELEASE FROM MARINE REACTORS DUMPE KARE A TH ASE DN I IAEA, VIENNA, 1997 IAEA-TECDOC-938 ISSN 1011-4289 © IAEA, 1997 Printed by the IAEA in Austria April 1997 FOREWORD
In 1992 the news that the former Soviet Union had, for over three decades, dumped high level radioactive wasteshalloe th n i s w waterKare a causeth Se af o s d widespread concern, especialln yi countries with Arctic coastlines. The IAEA responded by launching an international study, the International Arctic Seas Assessment Project (IASAP) orden i , asseso t r e potentiath s l healtd han environmental implications of the dumping and to examine the feasibility of remedial actions related to the dumped wastes.
199y Russiae Ith 3nMa n Federation provided informatio IAEe th Ao nt w aboulo hige d th t han level radioactive waste dumped in the Arctic Seas ("White Book-93"). According to the "White Book-93 totae th "l amoun f radioactivito t y dumpe Arctie th n di c Seas morswa e tha PBq0 n9 e .Th items dumped included six nuclear submarine reactors containing spent fuel, spent fuel from an icebreaker reactor reactorn ,te s without fuelliquid soli d an leve, w dan d lo l waste nucleae Th . r reactors fuee th l d fro icebreakee an mth r reactor were dumpe shalloe th n di w bay f Novayso a Zemlyn i d aan the Kara Sea.
Within the framework of the IASAP project the Source Term Working Group was established, wit objective hth f determinineo informatioe gth n neede dimpacn i abou e waste us tth r assessmenefo t calculations in evaluating the usefulness of possible remedial actions. This involves having the knowledge of the inventory of radionuclides in the dumped objects, and protective barriers provided to them either by initial construction or prior to dumping, and of the likely behavior of the barriers wit he marin timth n i e e environment e e groueffortth Th f .p o s have been focusee dumpeth n o d d reactors and especially the dumped reactor fuel which contains the highest inventory of radionuclides.
The present report summarizes the work carried out by the Source Term Working Group of IASAP during 1994-1996. The report is based on the studies concerning the initial and current radionuclide inventories, operational histor constructiod yan reactore th f no . Sivintses Y carrie y b t dou v Russiae th f o n Research Center "Kurchatov Institute", Mosco . YefimoE Institute d th w an f o vf eo Physic Powed san r Engineering, Obninsk, Russian Federation workine Th . g group convened five times and evaluate result e studiee dth th developef d so san d model predictior sfo potentiaf no l releasee th o st environment e calculationTh . s wer Royae eth carriet a l t Navadou l College, Greenwichy b , UK , . LynnN . WardeJ , Lawrence S.Timmd th nt an a d ean s Livermore National Laboratory, California, USA, by M. Mount.
IAEe Th A wishe expreso st gratituds sit thosl participateo al o et wh Sourc e wore th th f kn o di e Term Working Grou preparatiod pan thif no s report IAEe Th .A officer responsibl thir efo s wors kwa K.-L. Sjoeblom of the Waste Safety Section, Division of Radiation and Waste Safety. EDITORIAL NOTE
In preparing this publication press,for IAEAthe staffof have pages madethe up from the original manuscript(s). The views expressed do not necessarily reflect those of the governments of the nominating Member States or of the nominating organizations. Throughout textthe names Memberof States retainedare theyas were when textthe was compiled. Theof use particular designations countriesof territoriesor does imply judgementnot any by publisher,the legal the IAEA, the to status as of such countries territories,or of their authoritiesand institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement recommendationor IAEA. partthe the of on CONTENTS
1. INTRODUCTION ...... 7 1.1. Background ...... 7 . 1.2. International Arctic Seas Assessment Project ...... 7 1.3. Objectives of the Source Term Working Group ...... 8 1.4. Disposal summary ...... 8 .
2. SOURCE TERM DEVELOPMENT ...... 8 . 2.1. Background ...... 8 . 2.1.1. Submarine pressurized water reactors ...... 8 . 2.1.2. Submarine liquid metal reactors ...... 10 2.1.3. Icebreaker pressurized water reactors ...... 0 1 . 2.2. Characteristic steae th mf so generating installations ...... 1 1 . 2.2.1. Submarine pressurized water reactors ...... 11 2.2.2. Submarine liquid metal reactors ...... 14 2.2.3. Icebreaker pressurized water reactors ...... 18 2.3. Reactor operating histories ...... 20 2.3.1. Submarine pressurized water reactors ...... 0 2 . 2.3.2. Submarine liquid metal reactors ...... 0 2 . 2.3.3. Icebreaker pressurized water reactors ...... 0 2 . 2.4. Radionuclide inventories ...... 22 2.4.1. Submarine pressurized water reactors ...... 2 2 . 2.4.1.1. Fission products and actinides ...... 22 2.4.1.2. Activation products ...... 3 2 . 2.4.2. Submarine liquid metal reactors ...... 4 2 . 2.4.2.1. Fission product actinided san s ...... 4 2 . 2.4.2.2. Activation products ...... 4 2 . 2.4.3. Icebreaker pressurized water reactors ...... 25 2.4.3.1. Fission product actinided san s ...... 5 2 . 2.4.3.2. Activation products ...... 25 2.5. Disposal operations ...... 36 2.5.1. Submarine pressurized water reactors ...... 36 2.5.2. Submarine liquid metal reactors ...... 7 3 . 2.5.3. Icebreaker pressurized water reactors ...... 8 3 . 2.5.3.1. Reactor compartment ...... 38 2.5.3.2. Core barre spend an l t nuclear fuel ...... 8 3 .
3. MODELLING STRATEGY ...... 1 4 . 3.1. Method assumptiond san s ...... 1 4 . 3.1.1. Initial assumptions ...... 41 3.1.2. Route water sfo r ingress ...... 2 4 . 3.1.3. Modelling methods ...... 42 3.1.3.1. Calculation of k factors ...... 42 3.1.3.2. Analysi geometrf fuen so pi l y ...... 3 4 . 3.1.3.3. Analysi f activatioso n product release ...... 4 4 . 3.1.4. Calculation of activity release rates ...... 44 3.2. Model construction ...... 5 4 . 3.2.1. Submarine pressurized water reactors ...... 5 4 . 3.2.2. Submarine liquid metal reactors ...... 45 3.2.2.1. Corrosion of coolant and damaged fuel in the left board steam generator ...... 46 3.2.2.2. Corrosion of thermal shields ...... 46 3.2.2.3. Corrosio corf no e ...... 9 4 . 3.2.2.4. Corrosion of control rods ...... 52 3.2.3. Icebreaker pressurized water reactors ...... 52 3.3. Corrosion rates ...... 3 5 . 3.3.1. Metals ...... 54 3.3.1.1. Steels ...... 4 5 . 3.3.1.2. Europium hexaboride ...... 54 3.3.1.3. Lead-bismuth eutectic ...... 54 3.3.2. Reactor fuels ...... 54 3.3.2.1. Uranium-aluminum pressurized water reactor fuel ...... 4 5 . 3.3.2.2. Uranium-dioxide icebreaker pressurized water reactor fuel ...... 4 5 . 3.3.2.3. Uranium-beryllium liquid metal reactor fuel ...... 5 5 . 3.3.3. Fillers ...... 55 3.3.3.1. Organic materials ...... 55 3.3.3.2. Concrete ...... 5 5 . 3.4. Release scenarios ...... 6 5 . 3.4.1. Submarine pressurized water reactor releases ...... 56 3.4.2. Submarine liquid metal reactor releases ...... 59 3.4.3. Icebreaker pressurized water reactor releases ...... 1 6 . 3.4.3.1. Fission produc actinidd an t e release ...... 1 6 . 3.4.3.2. Activation product release ...... 61 3.4.3.3. Icebreaker total release rates ...... 1 6 . 3.4.4. Kar totaa aSe l release rates ...... 3 6 .
4. RESULTS AND ANALYSES ...... 63 4.1. Reliability ...... 63 4.1.1. Information on the steam generating installation structures and materials ..... 63 4.1.2. Radionuclide inventory ...... 6 6 . 4.1.3. Values of the best corrosion rates ...... 66 4.1.4. Degree of pessimism used in the models ...... 66 4.2. Sensitivity ...... 7 6 .
5. REACTOR CRITICALITY ...... 68 5.1. Contro corrosiod ro l n ...... 8 6 . 5.2. Structural changes ...... 9 6 . 5.2.1. Corrosion of the stainless steel supporting structure ...... 69 5.2.2. External corrosion ...... 71 5.3. Ingres f wateo s r ...... 1 7 . 5.4. Consequences of criticality ...... 72 5.5. Further work ...... 2 7 .
6. REMEDIAL CONSIDERATIONS ...... 3 7 . 6.1. Remedial measures ...... 73 6.1.1. Reinforcement of the existing containment barriers ...... 73 6.1.2. Recovery of the dumped spent nuclear fuel for land storage ...... 74 6.2. Structural integrity of the spent nuclear fuel containers ...... 74
7. CONCLUSIONS ...... 4 7 .
ABBREVIATIONS ...... 78 DEFINITIONS ...... 79 REFERENCES ...... 81 CONTRIBUTORS TO DRAFTING AND REVIEW ...... 83 1. INTRODUCTION
1.1. BACKGROUND
e sprinInth f 1993go Russiae th , n report, Facts Problemsd an Related Radioactiveo t Waste Disposal Seasin Adjacent Territorythe to Russian ofthe Federation released[1]was , findingThe . s presente than di t report wer resule scientifia eth f o t c study commissione Octoben di re 199th y 2b PresidenOffice th Russiae f o eth f o t n Federation. Relate Arctie th o dt c area White th , e Booke th s a , report was later called, reported that:
(1) between 1965 and 1988,16 marine reactors from seven former Soviet Union submarines and the icebreaker Lenin, eac f whicho h suffered some for f reactomo r accident, were dumpe t fivda e sites in the Kara Sea; ) (2 between 196 1991d 0levean w ,lo l liquid radioactive wast discharges ewa White e t siteda th n si , Barents Kard an , a Seasd an ; ) (3 between intermediat196d 1991d an 4an w ,lo e level solid radioactive wast dumpes ewa t siteda s Barente inth Kard san a Seas.
Of the discarded marine reactors, six of the 16 contained their spent nuclear fuel (SNF). In addition, approximately 60% of the SNF from one of the three icebreaker reactors was disposed of in a reinforced concrete and stainless steel (SS) shell container. The vast majority of the low and intermediate level solid radioactive wast disposes ewa container n i f do unknowf so n compositione Th . Kara Sea disposal sites for the 16 marine reactors and low and intermediate level solid radioactive waste varied in depth from 12 to 380 m. In particular, the icebreaker reactors and part of their SNF were reportedly disposed of in Tsivolka Fjord at an estimated depth of 50 m.
The White Book also reported estimate totaf so l radioactivit time disposalth f eo t y a , these were:
(1) 8.5 x 10" Bq of fission products in the SNF; f activatio o 10x q 17 5B 3. n ) product(2 reactoe th n i s r components; unspecifie f 10x o q 9 14B 8. leve w ) d lo (3 origil e liquith n ni d radioactive waste whicf o , ove% hr50
was discharged in the Barents Sea; and
4 (4) 5.9 x io B1 q of unspecified origin in the low and intermediate level solid radioactive waste, f whico s discardeove% hwa Kar95 e r a [1]th aSe n .d i
With a few exceptions, information was provided on the radionuclides present in the wastes and no estimate was made of the current levels of radioactivity in the dumped wastes.
1.2. INTERNATIONAL ARCTIC SEAS ASSESSMENT PROJECT
International concern ove possible rth e healt environmentad han l effects both shor long-termd tan , local, regional, and global, from disposal of these aforementioned radioactive wastes in the shallow waters of the Arctic Seas prompted the International Atomic Energy Agency (IAEA), as part of its responsibilitie Londoe th o st n Conventio f 1972no initiato ,t Internationae eth l Arctic Seas Assessment Project (IASAP )IASA e [2]Th . P formally bega Februarn no y 1,199 Oslon 3i , Norwa meetine th t ya g on the Assessment of Actual and Potential Consequences of Dumping Radioactive Waste into Arctic Seas, organized by the IAEA in cooperation with the governments of Norway and the Russian Federation.
The stated objective IASAe : th to f e so P ar
) (1 asses riske sth huma o st environmene nth healto t d han t associated wit radioactive hth e waste dumpe Kare Barentd th aan n di s Seasd an , ) (2 examine possible remedial actions relatedumpee th o dt d wasteadviso t d whethen sean o r they are necessary and justified [2].
To meet these objectives, the work was organized into five working areas:
(1) source term, ) (2 existing environmental concentrations, ) (3 transfer mechanism modelsd san , (4) impact assessment, and ) (5 remedial measures.
1.3. OBJECTIVES OF THE SOURCE TERM WORKING GROUP
The Source Term Working Group was established to prepare a detailed inventory of and release rates for the radionuclides dumped at each Kara Sea disposal site. To this end, inventories were calculate spene th r td fo nuclea r fuel (SNF activated )an d component timt sa disposaf eo projected an l d forwar timen i d ; protective barriers f anyi , , were evaluate r theidfo r potential effec radionuclidn o t e release a numbe d an f ;mode o r l scenarios were develope o predict d e potentiath t l release th f eo radionuclide inventory into the Kara Sea.
This document summarizes the efforts of the Source Term Working Group to complete the tasked objectives under the IASAP. It presents a detailed discussion of the fission product, actinide, and activation product inventories at each Kara Sea disposal site and a detailed description, with assumptions modele th f o , s use predico dt t potential releas radionuclidee th f eo s intKare oth a Sea. Results of the release scenario models, reliability of the model input parameters, and an analysis of the sensitivity of the results to changes in the protective barrier lifetimes and SNF corrosion rates are then presented. The potential for recriticality of the SNF bearing cores and considerations for potential remedial measures are next addressed. Finally, conclusions are drawn with respect to the radionuclide releases at each Kara Sea disposal site from the SNF and activated components.
It should be noted that this document discusses only the estimates of the inventory and release f radionuclideo s associated wit marine hth e reactors dumpe Kare leveth w an lo di lSea e liquiTh . d radioactiv intermediatd an ew wastlo d ean level solid wastes discarde Barente th n Kardd i san a Seas wer t includeeno thin di s study.
1.4. DISPOSAL SUMMARY
Table I presents a summary of pertinent disposal information for the marine reactors dumped in the Kara Sea [1]. Figure 1 shows a map of the Northeast coast of Russia with Novaya Zemlya and the approximate locations of the five disposal sites.
2. SOURCE TERM DEVELOPMENT
2.1. BACKGROUND
followine Th sectiono gtw s describ accidente eth s that occurred aboar nucleae dth r submarines and icebreaker that led to their ultimate disposal in the Kara Sea.
2.1.1. Submarine pressurized water reactors [3, 4, 5, 6]
Six of the seven nuclear submarines contained two pressurized water reactors (PWRs) each. Eleven of these PWRs were dumped into the Kara Sea between 1965 and 1988: eight within and three without their reactor compartments (RCs) thesl Al . e nuclear submarines suffered some for reactof mo r accident; however, many specifics of the reactor design, maximum thermal power, compartment layout, detailed operating histories, and accident scenarios remain classified. TABL . EPERTINENI T DISPOSAL INFORMATIO MARINE TH R ENFO REACTORS DUMPE] KAR[1 E A TH ASE DN I
Disposal site Yeaf ro Factory Dumped unit Disposal Disposal Number of reactors Total activity (PBq) disposal number coordinates' depth2 (m) • Without spent With spent time Ath t e nuclear fue] nuclear fuel of disposal Î994 Abrosimov Fjord 1965 901 Reactor compartment 71° 56.03' N 20 (10-15) 2 3,0 0.73 55° 18.15' E 285 Reactor compartment 71° 56.03' N (10-150 2 ) 1 1 12 0.66 55° 18.08' E 254 Reactor compartment 71° 55.22' N 20 2 0.093 0.009 55° 32.54E ' 1966 260 Reactor compartment 71° 56.03' N 20 2 0.044 0.005 55° 18.08E ' Tsivolka Fjord 1967 OK- 150 Reactor compartmenN ' t 10 an. d26 ° 74 50 3 0.6' 20 2.2 special container with fuel 58° 36. 15' E Novaya Zemlya Depression 1972 421 Reactor 72° 40' N 300 I 1.0 0.29 E ' 10 ° 58 Stepovoy Fjord 198Î 601 Submarine 1.253 ° N ' 72 50 (30) 2 1.7 0.84 55° 30.25E ' Techeniye Fjord 1988 538 Reactors 73° 59' N 35-40 2 0.006 0.005 66° 18' E
Total 10 6.6 37 4.7
1 Disposal site coordinates for all units except those from factory number OK-150 are from reference [1]. Disposal site coordinates for the OK-350 units are from reference [4]. disposae Th 2 l depth froe sar m reference [1]; thos parenthesin ei s were obtained during joint Norwegian-Russian scientific cruise 199n si 1994d 3an . 3 Thermal shields, hardware approximatel d discardeF an , SN f o specian di % y60 l container. Techeniyt Fjord
Kara aSe
Zemlya Depression
FIG. 1. Approximate locations of the marine reactor disposal sites in the Kara Sea on the northeast coast of Russia.
A criticality accident aboard the submarine identified as factory number 421 is known to have caused over pressurization of the right board reactor pressure vessel (RPV). The fuel rods were reported to not be damaged; however, a decision was made to not re-use the RPV. As such, the SNF t removedno s wa .
2.1.2. Submarine liquid metal reactors [7, 9]
remainine Th g nuclear submarine, designate factors da y number 601, containe liquio dtw d metal reactor maximusW (LMRsM 0 7 m f )o therma l power eac used han d coolane Pb-Bth s a r ihea o t t transfer medium. The steam generating installation (SGI) began operation in December 1962 and operated successfull duratioe th firs e r th y tfo f ncoro e load. Both reactors were reloade Septemben di r 1967 and operated at 10% of full power until May 24, 1968 when a portion of the left board reactor core channels became blocked while the submarine was at sea. As a consequence, approximately 20% of the left board reactor fuel was destroyed and deposited in the associated steam generator (SG) and volume compensator via the sealed primary circuit. The submarine subsequently returned to base on power fro righe mth t board reactor, shu seales tabour o down wa n d o t Junan , , 1968e6 righe Th .t board t approximatela reacto y s restarteda rwa e f ful o on l Aprir n d% poweri fo y 20 n l ru 197 .d 2an Ultimately s discardewa t i , d along wit lefe hth t board reactor.
2.1.3. Icebreaker pressurized water reactor] 8 , s[6
Launche Leningran di 1959n di icebreakee th , firse r Leninth t s nucleawa r merchante shith pn i world. During 31 years in commission, the icebreaker had two separate SGIs. The first SGI contained three PWRs of 90 MW maximum thermal power each and operated from 1959 to February 1965, when during routine repair of the SGI, an operator error allowed the core of the center line (N2) PWR to be left without water for some period of time. As a consequence, a part of the reactor core was damaged due to residual heat. It is this first SGI that forms the basis for the icebreaker source term.
10 The following sections detail what is currently known or estimated about the seven nuclear submarines and icebreaker with respect to the characteristics of their SGIs, their reactor operating histories, their radionuclide inventories, and their disposal operations.
2.2. CHARACTERISTICS OF THE STEAM GENERATING INSTALLATIONS
Know estimater no d characteristic seve e SGIe th th f nf so s o nuclea r submarine icebreaked san r detailee ar followine th n di g sections.
2.2.1. Submarine pressurized water reactor] 6 , 4 , [3 s
Specific information about the SGIs associated with the 11 discarded PWRs remains classified. Generally, the entire SGI, including the SGs and circulation pumps, was located aft of the submarine sail in an isolated RC. The two PWRs were aligned vertically, either in a plane perpendicular to the keel or along the keel, and were surrounded by a water-filled steel shield tank. Biological shields were located above the shield tank and around each PWR; however, the specifics of their construction material unavailablee ar s .
Each nuclear submarin consisteR PW ea cylindrica f o d l steel RPV ,a reacto s rit cord an e associated support structure, and a series of radial and bottom thermal shields, the latter being employed to reduce heat and radiation effects on the RPV and subsequently extend its operating life. For submarine factory numbers 901, 285, 254, 260, and 538 the RPVs were made from type 15X2MC&A carbon steel, with approximate dimensions:
(1) 1.4 m diameter, heightm 7 3. , ) (2 thicm m k 0 wall12 thicm s m ) witk 6 (3 internaha l claddin f typgo e 1X18H9, TSS thicm m k 0 botto31 thic mm ) m (4 witk 5 internaha l claddin f typgo e 1X18H9d an , TSS (5) 400 mm thick lid.
Figur type showe2 th ecross-sectiof e useo s th submarinV n di RP a f no e factory numbers 901, 285, 254, 260 538d ,an .
For submarine factory number 421, the RPVs were made from a variety of carbon steels, with approximate dimensions:
(1) 2m diameter, (2) 3.4 m height, (3) 120 mm thick walls of type 15X2MOA-A carbon steel with a 7.5 mm thick internal cladding of SS, thicm m k0 botto11 typf m ) o e(4 12X2MOA-A carbon steethicm m lk d wit5 claddin7. an h, a SS f go (5) 390 mm thick lid of 25X2MOA carbon steel.
Figur type showe3 th ecross-sectiof e o usesth submarinV n di RP a f no e factory number 421.
e heigh diameted Th an t corR f eaco r eh PW remain s classifiedpurposee th r f Fo theso s. e calculations, each cors assume ewa e loade b 235e f 235o o dt U UTh dg .k enrichmenwit 0 5 h s wa t assumed to be 7.5% for the cores of submarine factory number 285 and 20% for all others [3]. Fuel rods were assume constructee b o dt U-Aa f do 1 alloy.
submarinr Fo e factory numbers 901, 285,254, 260 538d radiae an , ,th l thermal shields consisted f fouo r concentric cylinders positioned around each reacto e bottorth cord mean thermal shields consisted of four cylindrical plates positioned below the lower core plate. The four concentric
11 FIG. 2. Cross-section of a reactor pressure vessel of the type used in submarine factory numbers 901, 285, 254, 260, and 538 [3].
12 o-
60 ' o i i l
aS.
§2• Anticorrosion coating I cylinders were made of type 1X18H9T SS of, from the core surface outward, 25-, 40-, 40-, and 20-mm thickness, respectively. The four concentric cylinders and the inner surface of the RPV wall were separate primary-circuiy db t wate , fro e corof r mth e surface outward, 5-md 20- , an 20-m 6- , 5- , thickness, respectively e loweTh .r core platd fouan er cylindrical plates were mad f typo e e 1X18H9 , froe of cor mth S e TS surface downward, 28-, 45-, 40-, 50- d 40-m,an m thickness, respectively. The lower core plate, four cylindrical plates, and the inner surface of the RPV bottom were separated by primary-circuit water of, from the bottom surface of the core downward, 30- 47-, 60-md an , m5- thickness, 5- , 5- , respectively. Above eac uppee h th cor e reactore ear rin th f go r core barre foud an lr additional cylindrical plates lowee comprisinth ,o tw r balance uppege th th f ero core barrel. The four cylindrical plates were made of type 1X18H9T SS of, from the core upward, 150-, 150-, 12-, and 25-mm thickness, respectively. The four cylindrical plates and the inner surface of the lid were separated by primary-circuit water of, from the core surface upward of, 400-, 70-, 74-, and 22-mm thickness, respectively. Above each RPV lid is an additional plug consisting of, from the RPV lid upward, 650 mm thick B4C and 120 mm thick type CT-3 steel.
For submarine factory number 421, the radial thermal shields consisted of four concentric cylinders positioned around each reacto e bottorth cord man e thermal shields consiste f seveo d n cylindrical plates positioned belo lowee wth r core plate foue Th r. concentric cylinders were madf eo an unspecified SS of, from the core surface outward, 25-, 25-, 30-, and 30-mm thickness, respectively. foue Th r concentric cylinder innee walth V d rl RP surfac sweran e th ef eseparateo primary-circuiy db t wate , fro core of r mth e surface outward, 4.5-, 10-, 10-, 15-md 10-an , m thickness, respectivelye Th . lower core plate and seven cylindrical plates were made of type 8X18H10T SS of, from the core surface downward, 90-, 25-, 25-, 25-, 25-, 25-, 25-m25-,d an m thickness, respectively lowee Th . r core plate, seven cylindrica l innee platesbottoV th d rRP surfacman ,e th wer f o ee separate primaryy db - circuit water of, from the bottom surface of the core downward, 60-, 70-, 12-, 12-, 12-, 12-, 12-, 12-, and 65-mm thickness, respectively. Above eac uppee h th cor e r ear cor o additionae tw plat d ean l cylindrical plates, the latter two comprising the balance of the upper core barrel. The upper core plate cylindricaano dtw l plates were madunspecifien a f e o , fro core of mth S e dS surfac e upward, 50-, 120-, and 210-mm thickness, respectively uppee Th . r core plate cylindricao ,tw l innee platesth d r surfac,an e of the lid were separated by primary-circuit water of, from the core surface upward, 15-, unknown-, 95-, and 60-mm thickness, respectively.
2.2.2. Submarine liquid metal reactor] 9 , [7 s
The entire SGI, including the SGs, circulation pumps, and primary circuit volume compensators, was located aft of the submarine sail in an isolated RC. The two LMRs were aligned vertically in a plane perpendicula keee werd th an l eo rt surrounde lead-watea y db r tank shield SGse Th ,. circulation pumps, and primary circuit volume compensators were enclosed in lead lined structures.
alsR Eaco LM hconsiste a cylindrica f o d l RPV a reacto, s associaterit cord an e d support structure, and a series of radial and bottom thermal shields. The RPV was made from SS, with approximate dimensions:
(1) 1.8 m diameter, (2) 3.7 m height, and thicmm k walls30 . (3)
Externa outee wer th cylindricaV o ro lt e surfactw RP e th f eo l channe eaclm regionhm tha0 3 tf swero e formed throug additioe hconcentrith o respectively, tw f no mm 0 2 c cylindricad an . m m shellS 0 lS 1 f so Figure 4 shows the cross-section of a RPV of the type used in submarine factory number 601.
The height and diameter of each LMR core was approximately 800 mm and 780 mm, respectively. A radial reflector was present and consisted of, from the core surface outward, 10 mm
14 FIG. 4. Cross-section of a reactor pressure vessel of the type used in submarine factory number 601 [7].
15 thick SS, 65 mm thick BeO, and 8 mm thick SS. Above the core was a special shield plug through whic three hth e emergency protection rods (EPRs) contron te , r compensatioo l n rods (CCRs)4 2 d an , emergency cooling tubes (ECTs) passed e EPR CCRd Th .an s s extended int e cord weroth an e e enclosed in special steel channels of approximately 20 mm inside diameter while the ECTs extended through portion aboud radiaan e f th t o s l therma CCRn l te shield e s Th wer. e approximatelm m 7 1 y inside diameter each and arranged in the core as follows: centee coree th th t f ,a o r e on ) 1 ( (2) three evenly spaced at a radius of 97.5 mm on a 120° arc, and (3) six evenly spaced within the annulus between radii of 97.5 and 292.5 mm.
The three EPRs were approximately 20 mm outside diameter each and evenly spaced at a radius of 195 mm on a 120° arc. Both the EPRs and CCRs were made of europium hexaboride (EuB6). Figure 5 shows a schematic layout of the CCRs and the EPRs within the core of submarine factory number ECT4 2 e s Th 601wer. e approximatel insidm m outsid m e0 y7 m diamete 0 e8 diameted an r d an r evenly spaced on a radius of 641.5 mm. There was approximately 300 mm of the Pb-Bi coolant reactoe th f botto e o reactobetweee rth p corth d to f m o e an r nth shiel d plug. Each reactor cors ewa loade enrichmen n 235f a o t Uda g witk 0 h9 f 90% o t . Fue pelletd ro l s were constructe f U-Bdo e alloy in a BeO ceramic matrix and were of. approximately 10 mm diameter. The pellets were covered with a 0.1 mm thick layer of Mg and clad in SS of 0.5 mm thickness. The resultant fuel rods were 11 mm outside diameter and arranged on a 13 mm triangular pitch. Considering the fuel rod pitch and core diameter, and accounting for the presence of the three EPRs and ten CCRs, there were approximately 3000 fuel rod R coren eaci s LM h. Typically e distributio th ,R cor eLM materialf o n s wa s approximately 54% fuel, 36% Pb-Bi, and 10% SS. Figure 6 shows cross-sections for CCR and EPR channel type th ef suseo submarinn di e factory number 601. Figur showe7 sschematia c layoue th f o t ECT cord submarinsan n ei e factory number 601.
10-mlayeS S mr
BeO reflector
m Site of emergency protection rod
, Site of control or compensation rod
FIG.Plan5. core reflectorviewthe and liquida of of metal reactor submarineof factory number 601, showing the approximate position of the emergency protection rods and the control or compensation rods.
16 Sectio : Channea nDO f o l Section EE: Channel of an contro r compensatioo l d ro n emergency protectiod nro
Cross-sections6. FIG. controlfor compensationor emergencyand rod protection channelsrod ofthe type used submarinein factory number [7]. 601
17 radiae Th l thermal shields consiste f nindo e concentric cylinders positioned around each reactor core between the outer surface of the reflector region and the inner surface of the RPV wall; the bottom thermal shields consisted of two cylindrical plates positioned below the lower core plate. The nine concentric cylinders were made of an unspecified SS of, from the reflector region surface outward, 11.5-, 34-, 36-, 36-, 36-, 120-, 36-, 36-, and 25-mm thickness, respectively. Cylinders five, 36 mm thickness, and six, 120 mm thickness, were adjacent to one another with the latter containing 24 evenly insidspacem m e0 d 8 diamete r hole accommodato st ECTse enine th Th e. concentric cylindere th d san inner surface of the RPV wall were separated by, from the reflector region surface outward, nine annular Pb-Bi coolant channelthicknesm m 3 f so s each. Figur showe8 relative sth e layoucoree th f ,o t BeO reflector, SS thermal shields, Pb-Bi coolant channels, ECTs, and RPV wall in submarine factory cylindricao numbetw e Th r 601l . platebottoe th f smo thermal shield were madunspecifien a f eo S dS of, fro e lowemth r core surface downward d 100-man , 200-m thickness, respectively e loweTh . r cylindrical plat adjacens innee separatewa s bottoth V wa o rt t d RP surfac me an dth frof eo uppee mth r cylindrica thicm m kPb-Be 0 zonth l10 f plateia o coolanty eb .
2.2.3. Icebreaker pressurized water reactor] 8 , [6 s
The entire first SGI, including the SGs and pumps, was located in the center of the ship in an isolated RC. The three PWRs were aligned vertically in a plane perpendicular to the keel and were surrounded by a large water-filled steel shield tank. The biological shield located above the three PWR s mad swa f limonit eo e concret heaa d t ean resistan t compositio f graphitno borond ean .
Each icebreake consisteR PW rcylindrica a f do reactoa l RPV, associates rit cord ean d support structure, and a series of radial thermal shields. The RPV was made from type 12X2MOA carbon steel, with approximate dimensions:
(1) 2m diameter, heightm 5 d an , ) (2 thicm m k 0 wall10 thicm s m ) witk 5 (3 internaha l claddin f typgo e 1X18H9. TSS
The height and diameter of each core was approximately 1.6 m and 1 m, respectively, with each core containing 219 cylindrical technical fuel channels (TFCs) (fuel assemblies) on a 64 mm triangular pitch. Within 30 of the TFCs were central channels for six control rods, ganged together in two groups scra4 f three2 o d man , rods, ganged togethe groupx si n i f rfour so . Driveeighe th groupd r ro t sfo s TFCs9 containe9 21 e 18 , th f fue6 O werd 3 containe 0 lidl V 3 . rod e d RP situate san e fue0 th d3 n ldo rods. Overall, there were 7,704 fuel rods in each core. The fuel rods were constructed of UO2 sintered cerami diameterm cm pellet 5 4. f , so enriche 5.0o d%t 235pellete UTh . s were cla Zr-Nn di b S alloS r yo of 0.75 mm thickness with a 0.1 mm thick gap. The resultant fuel rods were 6.1 mm outside diameter.
The operating temperatures of the water in the main inlet and outlet tubes, the UO2 in the center of the fuel rod, the water between the TFCs, and the fuel element cladding at one-half of maximum power were 261°C, 313°C, 1400°C, 280°C, and 700°C, respectively. Water flow rates in the N2 reactor primary circuit bow and stern loops were 458 and 407 tonne-h"1, respectively, and the operating kg-cm"0 pressur18 s 2.ewa
The radial thermal shields consisted of five concentric cylinders positioned around each reactor coree fivTh e . concentric cylinders were mad f typeo e 8X18H10 , fro e corof mth S e TS surface outward, 19-, 15-, 25-m21-d , an 25-m, thickness, respectively five eTh . concentric cylindere th d san innewalV rlRP surfacwere th ef o eseparate primary-circuiy db t wate , fro e corof r mth e surface outward 10-md , m39-an - ,, 13-0 ,7- 4 , 15-thickness, , respectively f theso o e fivTw . e concentric cylinders were welded to the upper and lower rings, which connected with the upper and lower plates fixatioe TFCe th use th r forf o d fo st nreactoo e mth r core barrel uppee lowed Th . an r r platee th f so core barrel were also constructed of type 8X18H1OT SS and were 250- and 185-mm thick, respectively. Overall approximate th , e dimension core th ef so barre ldiameteheight m werm 4 3 1. ed . an r
18 Water envelope
Steam envelope
Emergency cooling tubes
Reflector
Core
FIG. Plan7. view ofliquida metal reactor of submarine factory number 601, showing approximatethe layout of the emergency cooling tubes and core. For clarity, individual thermal shields and Pb-Bi coolant channels are not shown.
Edge of reactor vessel BeO Emergency reflector cooling tube
Regions filled with Furfurol(F)
990.5 Distance from centre line (mm)
3 mm PbBi channels
FIG. 8. Radial cross-section of a liquid metal reactor of submarine factory number 601, showing the layout and approximate dimensions of the core, BeO reflector, stainless steel (SS) thermal shields, Pb-Bi coolant channels, emergency cooling tubes, reactorand pressure vessel watts.
19 2.3. REACTOR OPERATING HISTORIES
Known or estimated reactor operating histories of the seven nuclear submarines and icebreaker e detaile ar followine th n di g sections.
2.3.1. Submarine pressurized water reactors [3, 6]
Specific information abou operatine tth g historie SGIf so s associate discarded1 1 wit e hth d PWRs remains classified. The year of start-up and shutdown and the fuel burnup for each PWR is currently known, as is the fact that all contained their initial fuel load at the time of final shutdown. Start-up dates covere yean te re periodth d from 195 19688o t , wit earliese hth t date tha submarinr fo t e factory latese th d numbet datan 4 e tha25 rsubmarinr fo t e factory number 421. Shutdown dates coveree dth seven year period from 1961 to 1968, with the earliest date that for submarine factory number 901 and the latest date again that for submarine factory number 421. The longest period of SGI operation was four years and is that for submarine factory number 254. While the shortest period of SGI operation thas i r submarin yeae fo d t on an r s wa e factory number421d an . 1 Fue90 s l burnuPWRe th r spfo varie f 125o d w 0fro lo MW- ma r submarindfo e factory higa numbe f 388o t ho 1 0 42 rMW- r dfo submarine factory number 254.
2.3.2. Submarine liquid metal reactor] s[7
Subsequent to the core reload in September 1967, each reactor operated for approximately 300 effective-full-power hours (875 MW-d). The total neutron fluence of all energies on the RPVs and internal reactor constructions durin * 10I operatio 3 20gSG 2. n-cm' s nwa 2 .
2.3.3. Icebreaker pressurized water reactors [6, 8]
The first fuel load lasted from 1959 to 1962 and consisted of 80 kg of 235U in each reactor core. The integrated power productions with this first fuel load were equal to 17.8 GW-d for the Nl PWR and 18.0 GW-d for the N2 and N3 PWRs. Refuelling occurred in 1963. The second fuel load lasted from 1963 to 1965 and consisted of 129 kg 235U clad in SS in the Nl PWR and 75 kg 235U clad in Zr- operatedR PW 2 PWRs3 N froN e d mTh an . Jul, 2 196N y19 Novembe e o 3t th n i b , 196N d 17 r 3an from Jun , 196e22 Novembe4o t , 196totaa 13 r 4lfo daysperio7 meae 26 Th 2 f .n dN o powee th f ro PWR during the second fuel load was 53 MW. At the time of the reactor accident, the integrated power production for the N2 PWR was approximately 14.2 GW-d thermal and the burnup was equal GW-d-tonne"4 9. o t initial heavy metal. Integrated power production secone th r dsfo fuel load were 1 equal to 22.5 GW-d for the Nl PWR and 17.5 GW-d for the N3 PWR. The total mean neutron fluence internas it d an l reactoV RP 2 r N construction e th n o s during operatio witI firs e secone th hSG tth f no d fuel load was 5.5 * 1020 n-crrf2.
Table II summarizes the current available information for the steam generating installations of the marine reactors dumped in the Kara Sea [3, 4, 6, 7, 8].
20 TABLE ÎÎ. CURRENT AVAILABLE INFORMATION FOR THE STEAM GENERATING INSTALLATIONS OF THE MARINE REACTORS DUMPED IN THE KARA SEA [3, 4, 6, 7, 8]
Factory Reactor !)ÍU initial conditions' Steam generating installation RPV2 with SNF1 Typ« Position Load Enrichment Start-up Shutdown Burnup Disposal (kg) (%) date date (MW-d) date 901 PWR4 Left board 50 20 1961 1961 1710 May 1965 Yes PWR Right board 50 20 1961 1961 1670 May 1965 Yes 285 PWR Left board 50 7.5 1961 1964 2780 t 196Oc 5 No5 PWR Right board 50 7.5 1961 1964 2730 t 196Oc 5 Yes 254 PWR Left board 50 20 1958 1962 3080 1965 No PWR Right board 50 20 1958 1962 3880 1965 No R PW 0 26 Left board 50 20 1959 1962 1720 1966 No PWR Right board 50 20 1959 1962 1940 1966 No R PW OK-150 Left board 129" 5 Aug 1959 Oct 1965 40300 Sep 1967 No PWR Center line 75' 5 Aug 1959 196b Fe 5 32200' Sep 1967 No1 PWR Right board 75' 5 Aug 1959 Oct 1965 35500 Sep 1967 No R PW 421 Right board 50 20 1968 1968 1250 1972 Yes 601 LMR'Left board 90s 90 c 196De 2 May 1968 1 580'° Sep 1981 Yes LMR Right board 906 90 Dec 1962 Jun 1968 1 580'° Sep 1981 Yes 538 PWR Left board 50 20 1961 1963 1680 1988 No PWR Right board 50 20 1961 1963 1440 1988 No
1 Exceptin r factorgfo y numbers OK-15 601,d 0an 23SU initiai condition assumede sar . 2 Reactor pressure vessel (RPV). 3 Spent nuclear fuel (SNF) with thermal shields, hardware Furfurol(F)d an , . 4 Pressurized water reactor (PWR). 5 Thermal shield hardward san e only. secone th r dFo fue6 l load. 7 Burnup for the second fuel load was 14 200 MW-d. 1 Thermal shields, hardware approximatel d discardeF an , SN f o specia n di % y60 l container. ' Liquid metal reactor (LMR). 10 Burnup for the second fuel load was 875 MW-d. 2.4. RADIONUCLIDE INVENTORIES
Two independent estimates were prepared for the radionuclide inventories. One estimate was prepared by consulting members of the Group from the Russian Research Center "Kurchatov Institute" (RRCKI) Institute th d , f PhysicMoscoan eo , 8] Powed , san w [3 r Engineering (IPPE), Obninsk [7]. Another estimat s prepare State wa th ey db Institut f Applieeo d Ecology (SIAE), Moscown I [10]. preparin estimatese gth Groue th , p core modelPWRe th LMR d r an sfo s s were representee th y db icebreaker OK- 150 and submarine factory number 601 core models, respectively. While, the SIAE estimates use dWWER-100a 0 core cores R mode LM . represeno t d l Valuee an th R f so t PW bot e hth fuel burnup used in the Group inventory calculations came from RRCKI and IPPE records; those for SIAE were supplie Russiae th y db n Navy. Computer programs use botn di h inventory estimatee sar well established and benchmarked.
Results from the Group estimate, when compared to those from SIAE, showed the following:
(1) fission products are in good agreement for the icebreaker and no worse than a factor of 0.5 for the nuclear submarines, and (2) actinides agree within a factor of 0.5 for the icebreaker and are no worse than a factor of 0.1 for nucleae th r submarines.
Upon consideratio abovee th grouf e no th , p concluded that even though SIAE results e tenb o dt higher thereford an , e more conservative t represen, theno o y d bes e th t t ASAestimatI e th P r effortefo . First and foremost, the core models used in the Group estimate for the icebreaker and submarine factory number 601 represent the actual configurations; the SIAE models do not. Second, even though ther differencee ear s betwee core nth e configuration nucleae th f so r submarin icebreaked ean r PWRs, OK-15e th 0 mode mors i l e representativ true th e f coreo e configurations tha WWER-1000ne thath f o t . This is further substantiated by the fact that comparisons of the Group actinide results to those in other Russian reports [11] indicate difference moro n f eo s tha 20%n± .
e followinTh g sections detai methode th l s use estimato dt radionuclide eth e inventoriee th f so nuclear submarine icebreakerd san e result Th givee . sar Tablen ni XVIo t .I sII
Table III presents a summary of the estimated total 1994 radionuclide inventories of fission products, activation products actinided an ,marin e th n si e reactor , 12]8 . , s7 dumpe, [3 Kare a th an Se di
Table througV sI presenI hXV t individual summarie estimatee th f so d 1994 activit long-livef yo d radionuclides in the marine reactors dumped in the Kara Sea [3, 7, 8, 12].
2.4.1. Submarine pressurized water reactor] 6 , 5 , [3 s
2.4.1.1. Fission products and actinides
Due to classification issues, complete descriptions of the PWR nuclear SGIs are not yet available. Thus, even though the fuel for the submarine PWRs and the icebreaker were described as a U-A1 alloy and UO2, respectively, the model for the icebreaker SGI was assumed for the PWR SGIs. Fission product activitie nucleae th r sfo r submarine PWR quotiene basee th nuclease ar n th do f o t r submarine and icebreaker burnups followingivethe are nby and , g equation:
where
A¡(S) is the activity of the i-th radionuclide in the nuclear submarine (Bq), activite i-te th h th s i f A¡(Iradionuclidyo ) icebreakee th n ei r (Bq),
22 B(S) is the burnup in the nuclear submarine (MW-d), and B(I) is the burnup in the icebreaker (MW-d).
Actinide activities for the nuclear submarine PWRs are based on the product of the quotients of the icebreake nuclead ran r submarine enrichment nucleae th d sran submarin icebreaked ean r burnupsd an , followine givee th ar y nb g equation:
) (2 A.(S - ) •V 'V\E(S)|B(I)J l'
where
235e th UE(I s i enrichmen) icebreakee th n i t r (%), E(S) is the 235U enrichment in the nuclear submarine (%), and
all other terms are as previously defined.
2.4.1.2. Activation products
Again becaus f classificatioeo n issues, activation product activitie PWRe th basee r sar sfo n do calculationaa l procedure simila thao rt icebreakere t useth r dfo conservatismr Fo . assumes wa t i , d that
l internaal ) l (1 reactor constructions were made fro; mSS RPVe th s ) wer(2 e made froallow mlo y steel; (3) the fuel elements were clad in SS; (4) the total mass of SS in the internal reactor constructions was some 100 kg greater than that associated with the icebreaker disposal; ) (5 even though some RPVs dumped withou alsF theid oha SN t r thermal shield d internaan s l hardware removedactivatee th f o l dal , metal remaine r disposaldfo d an ; ) (6 neutron activatio internae th f no l reactor constructions constana t RPVa d s an swa t level ovee rth operating period of the SGIs.
Suffic sayo t t i e, that almos l activatioal t n products associated wit e nucleahth r submarine PWRs originate e internath n di l constructions made fro. mSS
Analysi f neutroso n activation shows that only four radionuclide f consequenco e r sar o n te t ea more years after reactor shutdown. The 14e Cyar ,60 Co, 59 "Nid Nian , . Their activities were estimated fro e followinmth g equation:
A. = CPTN.(a. + gl)| —— I (3) where
A¡ is the activity of the i-th activation product radionuclide (Bq), C is the conversion factor between fission rate and thermal power (3.1 x 1016 fissions-s'^MW"1), P is the mean thermal power of the reactor (MW),
T is the irradiation time (d),
59 62 3 Nj is the number of nuclei of the i-th target nuclideCoC,13 , "Ni, oNri (cnV), a¡ is the microscopic activation cross-section for the i-th target nuclide (cm2) relativa s i e measurg f neutroeo n spectrum hardness, resonance th s i I¡e integra i-te th h r targefo l t nuclide (cm2), radioactive th s i A. e decay constan e i-tth h f activatioo t n product radionuclide, 14C, 60Co, 59Nir o , 63Ni (d-1), and 235 1 macroscopie th Z s 23i jf c fission cross-sectior nfo U (cm" ).
23 In the calculations, the fuel burnup, in MW-d, is substituted for the product FT, and g is assumed to equal 0.2.
2.4.2. Submarine liquid metal reactors [7, 9]
2.4.2.1. Fission products actinidesand
Fission product and actinide activities for the nuclear submarine LMRs are based on their core histories and calculated neutron spectra. Calculation of the fission product and actinide inventories were accomplished with the AFPA [13] and CARE [14] codes, respectively. The latter code considers formatioe th decad individua3 9 nan f yo l actinide chain used BNAe san s th B [15] nuclear cross-section system.
In modelin lefe gth t board reactor wit damages hit d SNF assumes i fissio e t i ,th f o nd tha% 80 t products and actinides are in the RPV; the balance are located at the top of the SG.
2.4.2.2. Activation products
For the SS components of the nuclear submarine LMRs, the long-lived activation products of consequence are 60Co, 59Ni, and 63Ni. Their activities were estimated as follows:
21 A =
where
V is volume of the structural material considered (cm3), e concentratioth s i e i-tp,th h f targeo n t nuclide, "Co, 58e structura Nir 62th o Ni, n i , l material considered (cm"3), 21 summatioe th s i ^ neutron1 J 2 ove e th rn energy group approximation, j-i e microscopith s i j a¡ c activation cross-sectio e i-tth h f targeno t nuclid r neutron fo j-te th h f o s energy group (cm2), meae th s ni therma . W l neutron flux densit materiae th n yi r neutron fo lj-te th h f energso y group, (cm~2-s"'), and
all other terms are as previously defined. In these calculations, it was assumed that the mean reactor power during operatio approximatels n wa nomina e th f o l% powey10 tha d irradiatioe ran tth n tims ewa 10 times greater than the core burnup in effective-full-power hours. As such, the value of 3T used was assumed to be 10 times less than the nominal power neutron flux density in the material for neutrons j-t e thermae hth th corf energe d o th ean ln i shieldy . groupValue . sO . werf O ,s o e calculated i21-grouna p approximatio werd nan e those fro core mth e design analyses.
The use of Pb-Bi as a coolant and Eu in the CCRs results in the production of other long-lived radionuclides sucs ha Pb , Bi, Bi, Bi, Eud Euan , . Their activities were calculated using the approach described 205 207. 208 210m IJ2 IM
Typically, the distribution of LMR primary circuit activation products was approximately 30% in the SG, 40% in the RPV, and the balance throughout the remainder of the circuit and the volume compensator. However, in modeling the two LMRs, conservative assumptions were made. For the left board reactor where the accident occurred, 30% of the primary circuit activation products are
24 assumed to be in the SG and the balance are in the RPV. For the right board reactor, all of the primary circuit activation products are assumed to be in the RPV.
2.4.3. Icebreaker pressurized water reactor] s[8
2.4.3.1. Fission products actinidesand
Calculations of core cell burn-up in the N2 PWR were performed by the spectral code CETERA fissioe [16]Th . n produc actinidd an t e activities were estimated usin RECOe gth L [17] library data base, which was generated on the bases of the latest versions of the evaluated nuclear data files, ENDF/B-V, with corrections based on the results of critical experiments [18]. The criticality problem was solved for a realistic 3-D geometry model of a TFC by Monte-Carlo with RECOL and checked with MCNP [19 fresr ]fo h fuel load. One-group cross-sections were prepare burn-ur dfo p calculation f criticao l load f botso h fresspend an h t inpufued an lORIGEN-o t t 2 [20 r detaile]fo d radionuclide content calculations.
2.4.3.2. Activation products
All internal reactor constructions were made from type 1X18H9 RPVe th ; sT werSS e made from type 12X2MOA carbon steel. Almost all activation products associated with the icebreaker PWRs originate e internath n di l constructions made fro. mSS
Analysis of neutron activation shows that only four radionuclides are of consequence at ten or
60 59 more years after reactor shutdown. Co Thes"Nid Ni, U, C e an , ar .e Activitie f eaco s h were estimated from the following equation:
P =C A p + p -t- AK (5) ' m '*• « |lI
where
fractioe neutroe th th s i f no np flux capture maie th n di reactivity compensating lattice r unispe t
m fission; fractioe neutroe th th s i f no n , flup x capture additionae th n di l reactivity compensating latticer spe unit fission; AK is the fraction of the neutron flux captured in the shielding assemblage and lower plate per unit fission; Z is the macroscopic capture cross-section for thermal and intermediate neutrons in the steel of the
s shielding assemblage (cm"1), Zs+w is the macroscopic capture cross-section for thermal and intermediate neutrons in the steel and water of the shielding assemblage (cm"1), activatioe ratie th th f o s i n cross-sectio T f 13no C , 59Co, 5ltotae Ni th r 62 o lN,o t iactivatio n cross-sectio steee th lf no unde r irradiationd an ,
all other terms are as previously defined. Again, in the calculations, the fuel burnup, in MW-d, is substitute produce th . r dfo PT t
Text36. cont.p. on
25 £ TABLE HI. ESTIMATED 1994 RADIONUCLIDE INVENTORIES OF FISSION PRODUCTS, ACTIVATION PRODUCTS, AND ACTÎNÏDES ÎN THE MARINE REACTOR] 12 , 8 S, 7 DUMPE , KAR[3 E A TH ASE DN I
Factory Activity in 1994 number Fission products Activation products Actinides Total Becquerel Percent Becquerel Percent Becquerel Percent Becquerel Percent (Bq) (Bq) (Bq) (Bq) 901 7.2 x IO14 15 6.0 >t IO12 0.13 3.4 x IO12 0.073 7.3 x IO14 15 285 6.3 x IO14 13 1.3 >< 10" 0.27 8.1 x IO12 0.17 6.5 x IO14 14 254 - - 9.5 >« IO12 0.20 - - 9.5 x IO12 0.20
260 - - 5.1 >: IO12 0.11 - - 5.1 x IO12 0.11
OK- 150' 1.8 x 10" 39 2.3 >: IO14 5.0 8.3 x 10" 1.8 2.2 x 10" 46 421 2.9 x IO14 6.1 2.9 »< IO12 0.062 2.8 x IO12 0.061 2.9 x IO14 6.2 601 5.3 x IO14 11 3.0 >c IO14 6.5 3.6 x 10" 0.008 8.4 x IO14 18 12 538 - - 4.5 >: IO12 0.096 - - 4.5 x IO 0.096
Total 4.0 x 10" 86 5.7 >c IO14 12 9.7 x 10" 2.1 4.7 x 10" 100
e fissioTh n product, actinide twenty-seved an , n percen f activatioo t n product activities were discarde reinforcea n di d concret stainlesd ean s steel container. 1 TABL . ESTIMATEEIV D 1994 ACTIVIT LONG-LIVEF YO D RADIONUCLIDE LEFE TH T N SI BOARD REACTOR PRESSURE VESSEL WITHI REACTOE NTH R COMPARTMENT FROM NUCLEAR SUBMARINE FACTORY NUMBER 901 DUMPED IN THE ABROSIMOV FJORD [3, 12]
Activitn yi 1994 Radionuclide Becquerel (Bq) Percen) (% t Fission products 3H " 10 x 4 1. 0.038 8SKr 2.8 x 1012 0.78 "ST 8.5 x 10" 23 90-y " 10 x 5 8. 23 "Tc x 10' 6 °2. 0.0073 l25Sb 4.8 x 10' 0.0013 129j 7 10 x 6 2. 0.0000071 137Cs " 10 x 3 9. 26 13 I37gara x 10 0 9. 25 147Pm io'x 8 °9. 0.027 15ISm 2.1 x io12 0.58 Subtotal 3.6 x 1014 99 Activation products
uc x 10' 4 °1. 0.0038 «•Co 5.3 x 10" 0.15 59Ni " 10 1. 4x 0.038 63Ni 2.3 x io12 0.63 Subtotal 3.0 x IO12 0.82 Actinides 238Pu " 10 x 4 1. 0.039 239Pu 1.2 x 10" 0.032 240Pu 5.3 x io10 0.014 241Pu 1.2 x IO12 0.34 24lAm 1.7 x 10" 0.048 Subtotal 1.7 x IO12 0.47 Total U 10 x 6 3. 100
27 TABLE V. ESTIMATED 1994 ACTIVITY OF LONG-LIVED RADIONUCLIDES IN THE RIGHT BOARD REACTOR PRESSURE VESSEL WITHIN THE REACTOR COMPARTMENT FROM NUCLEAR SUBMARINE FACTORY NUMBER 901 DUMPED IN THE ABROSIMOV FJORD [3, 12]
Activitn yi 1994 Radionuclide Becquerel (Bq) Percent (%) Fission products 3H " 10 x 4 1. 0.038 85Kr x 10 8 122. 0.78 »Sr x 10 5 138. 23 90y 8.5 x 1013 23 "Tc x 10' 6 °2. 0.0073 125Sb 9 10 x 8 4. 0.0013 129j 7 10 x 6 2. 0.0000071 137Cs x 10 3 1J9. 26 13 l37Bara x 10 0 9. 25 '47Pm 9.8 x 10'° 0.027 151Sm x io 1 122. 0.58 Subtotal x io' 6 43. 99 Activation products 14C x io 4 101. 0.0038 «'Co " 10 x 3 5. 0.15 "Ni 1.4 x 10" 0.038 "Ni 2.3 x IO12 0.63 Subtotal 3.0 x io12 0.82 Actinides 238Pu " 10 x 4 1. 0.039 239Pu " 10 1. 2x 0.032 240Pu 5.3 x 10'° 0.014 12 24,pu x IO 2 1. 0.34 241Am " 10 x 7 1. 0.048 Subtotal 1.7 x IO12 0.47 Total iox 6 143. 100
28 TABLE VI. ESTIMATED 1994 ACTIVITY OF LONG-LIVED RADIONUCLIDES IN THE LEFT BOARD REACTOR PRESSURE VESSEL WITHIN THE REACTOR COMPARTMENT FROM NUCLEAR SUBMARINE FACTORY NUMBER 285 DUMPED IN THE ABROSIMOV FJORD [3, 12]
Activity in 1994 Radionuclide Becquerel (Bq) Percent (%) Fission products 3H 2 1 10 x 8 2. 0.44 85Kr 5.7 x 1012 0.88 wSr 1.5 x 1014 23 90y 4 1 10 x 5 1. 23 «Tc ° 10' x 4 4. 0.0067 125Sb 1.8 x 10'° 0.0028 I29j 4.2 x 107 0.0000065 l37Cs 1.6 x 1014 25 l37Bam 4 1 10 x 6 1. 24 l47Pm 3.7 x 10" 0.056 15lSm 3.6 x 1012 0.55 Subtotal 6.3 x 1014 98 Activation products HC ° 10' x 9 1. 0.030 «Co " 10 x 5 9. 0.15 59Ni 2 1 10 x 7 1. 0.26 63N¡ 2 1 10 x 8 2. 0.43 Subtotal 5.4 x 1012 0.84 Actinides
238pu 2.4 x 10" 0.037
239pu 5.1 x 10" 0.078
240pu " 10 x 3 2. 0.035 12 24,pu 6.4 x JO 0.99 24lAm 7.2 x 10" 0.11 Subtotal 8.1 x 1012 1.3 Total 6.5 x 1014 100
TABLE VII. ESTIMATED 1994 ACTIVITY OF LONG-LIVED ACTIVATION PRODUCTS IN THE RIGHT BOARD REACTOR PRESSURE VESSEL WITHI REACTOE N TH R COMPARTMENT FROM NUCLEAR SUBMARINE FACTORY NUMBER 285 DUMPED IN THE ABROSIMOV FJOR] D[3
Activity in 1994 Radionuclide Becquerel (Bq) Percent (%) Activation products MC 2.6 x 10'° 0.36 "Co 1.3 x 1012 17 59Ni 2 1 10 x 3 2. 31 63Ni 3.8 x 1012 51 Subtotal 2 1 10 x 3 7. 100
29 TABLE VIII. ESTIMATED 1994 ACTIVIT LONG-LIVEF YO D ACTIVATION PRODUCTN SI THE LEFT AND RIGHT BOARD REACTOR PRESSURE VESSELS WITHIN THE REACTOR COMPARTMENT FROM NUCLEAR SUBMARINE FACTORY NUMBER 254 DUMPED IN ABROSIMOE TH V FJOR] D[3
Activitn yi 1994 Radionuclide Becquerel (Bq) Percen) (% t Left board reactor MC 10'x 2 °2. 0.23 "Co 8.3 x 10" 8.8 59Ni " 10 x 9 1. 2.0 "Ni 3.1 x 1012 33 Subtotal x 10 2 124. 44 Right board reactor
i«c 2.8 x io'° 0.29 '"Co 1.1 x 1012 11 59Ni " 10 x 4 2. 2.5 "Ni x 10 0 124. 42 Subtotal x 10 3 125. 56 Total 9.5 x 1012 100
TABL . ESTIMATEEIX D 1994 ACTIVIT LONG-LIVEF YO D ACTIVATION PRODUCTE TH N SI LEFRIGHD AN T T BOARD REACTOR PRESSURE VESSELS WITHI E REACTONTH R COMPARTMENT FROM NUCLEAR SUBMARINE FACTORY NUMBER 260 DUMPED IN THE ABROSIMOV FJORD [3]
Activity in 1994 Radionuclide Becquerel (Bq) Percent (%) Left board reactor
I4C 1.3 x 10'° 0.25 "Co 4.6 x 10" 9.2 59Ni 1.1 x 10" 2.1 "Ni x 10 8 11. 2 35 Subtotal 2.4 x 1012 47 Right board reactor UC 10'x 4 °1. 0.27 '"Co " 10 x 3 5. 10 59Ni 1.2 x 10" 2.3 "Ni 10x 0 122. 40 Subtotal 10x 7 122. 53 Total 5.1 x 1012 100
30 TABLE X. ESTIMATED 1994 ACTIVITY OF LONG-LIVED RADIONUCLIDES FROM THE N2 REACTOR OF THE ICEBREAKER PLACED WITHIN THE CONCRETE AND STAINLESS STEEL SHELL CONTAINER DUMPE TSIVOLKE TH DN I A FJOR] 12 , D[8
Activity in 1994 Radionuclide Becquerel (Bq) Percent (%) Fission products 3H " 10 x 3 8. 0.042 85Kr 1.7 x IO13 0.83 "'Sr 4.4 x IO14 22 90y 4.4 x IO14 22 "Tc 1.3 x 10" 0.0068 125Sb 10'x 8 °5. 0.0029 129 1 2.1 x 10" 0.000011 l37Cs 4.8 x io14 24 4 l37Bam 4.6 x IO' 23 147Pm 1.4 x IO12 0.069 15lSm x io 1 131. 0.55 Subtotal x IO 8 151. 93 Activation products s I4C 10 x 3 4. 0.000000022 '"Co 1.1 x IO13 0.57 59Ni 4.9 x 10" 0.025 63Ni 5.0 x IO13 2.5 Subtotal 6.1 x IO13 3.1 Actinides 238Pu 1.0 x IO12 0.053 239Pu 5.0 x IO'2 0.25 12 240pu 2.3 x IO 0.11 24lPu 6.7 x IO13 3.4 24lAm x IO 1 2 7. 0.36 Subtotal 8.3 x io13 4.2 Total x IO 10 5 2. 100
TABLE XI. ESTIMATED 1994ACTIVITY OF LONG-LIVED ACTIVATION PRODUCTS IN THE THREE REACTOR PRESSURE VESSELS WITHI REACTOE NTH R COMPARTMENT FROE MTH ICEBREAKER DUMPE TSIVOLKE TH DN I A FJOR] 12 , D[8
Activity in 1994 Radionuclide Becquerel (Bq) Percent (%) Activation products 14C 1.2 x IO6 0.00000070 "Co x IO 11 3 3. 19 59Ni x IO 3 121. 0.79 63Ni x IO 3 141. 81 Subtotal x IO 7 141. 100
31 TABLE XII. ESTIMATED 1994 ACTIVIT LONG-LIVEF YO D RADIONUCLIDE METAE TH N SLI CONTAINER ENCASED SEPARATED RIGHT BOARD REACTOR PRESSURE VESSEL FROM NUCLEAR SUBMARINE FACTORY NUMBE DUMPE1 NOVAYE R42 TH N DI A ZEMLYA DEPRESSIO] 12 , N[3
Activity in 1994 Radionuclide Becquerel (Bq) Percen) (% t Fission products 3H 1.6 x 10" 0.055 "Kr 3.3 x 1012 1.1 x'Sr 7.4 x 1013 25 90y 7.4 x 1013 25 "Tc x 10 9 101. 0.0066 12SSb x 10 2 102. 0.0076 129i 1.9 x 107 0.0000065 137Cs 6.8 x 1013 23
I37gam 6.5 x 10° 22 147Pm 4.7 x 10" 0.16 l51Sm 1.7 x 1012 0.57 Subtotal x 10 9 142. 98 Activation products
I4C 1.2 x 10'° 0.0042 '"Co 9.8 x 10" 0.34 59Ni 1.0 x 10" 0.035 53Ni 1.8 x 1012 0.61 Subtotal x 10 9 122. 0.98 Actinides 238Pu 8.7 x 10'° 0.030 239Pu " 10 1. x 1 0.039 240Pu 10'x 9 °3. 0.013 12 24,pu 2.4 x 10 0.83 241Am 1.6 x 10" 0.056 Subtotal x 10 8 122. 0.97 Total x 10 9 142. 100
32 TABLE XIII. ESTIMATED 1994 ACTIVIT LONG-LIVEF YO D RADIONUCLIDE LEFE TH T N SI BOARD REACTOR PRESSURE VESSEL WITHIN THE REACTOR COMPARTMENT OF NUCLEAR SUBMARINE FACTORY NUMBER 601 DUMPED IN THE STEPOVOY FJORD [7]
Activity in 1994 Radionuclide Becquerel (Bq) Percent (%) Fission products 3H " 10 x 9 1. 5.3 85Kr 2.9 x 1012 0.79 ^Sr x 10 6 134. 13 90y x 10 6 134. 13 "Tc x 10' 2 °1. 0.0033 125Sb 2.4 x 10'° 0.0064 • 29J 2.6 x 107 0.0000071 137Cs 5.0 x 1013 14 137Bam x 10 8 134. 13 147Pm 3.7 x 109 0.0010 15lSm x io 5 121. 0.40 I55Eu 3.8 x 10'° 0.011 Subtotal x io' 1 42. 58 Activation products Control/compensation rods IS2Eu x IO 10 3 3. 8.3 1S4Eu 6.0 x IO12 1.4 Primary circuit medium "Co 4.2 x 10' 0.00011 59Ni 6 IO x 2 5. 0.0000014 "Ni 8 IO x 4 5. 0.00015 7 205pb IO x 7 6. 0.000018 207Bi 6.0 x IO9 0.0016 208Bi 2.2 x 10' 0.00060 9 2IOBjm 1.2 x io 0.00032 Core steel components »Co 9 IO x 3 6. 0.0017 S9Ni 2.9 x IO7 0.0000079 "Ni 9 io x 4 2. 0.00066 Core thermal shields "Co x IO 16 34. 13 59Ni 7.8 x 10" 0.21 "Ni 6.9 x IO13 19 Reactor pressure vessel "Co 2.5 x 10" 0.068 59Ni 1.8 x io9 0.00051 "Ni " JO x 1 1. 0.030 Subtotal " 10 1. 5x 42 Actinides 2}'Pu 9 io x 4 3. 0.00093 239pu " 10 x 4 1. 0.037 9 240pu IO x 6 2. 0.00071 9 24,pu 1.1 x io 0.00030 24lAm 4.6 x 10' 0.00013 Subtotal 1.4 x 10" 0.039 Total 3.7 x 10" 100
33 TABLE XIV. ESTIMATED 1994 ACTIVITY OF LONG-LIVED RADIONUCLIDES IN THE STEAM GENERATOR OF THE LEFT BOARD REACTOR WITHIN THE REACTOR COMPARTMEN NUCLEAF TO R SUBMARINE FACTORY NUMBE 1 DUMPERE 60 TH N DI STEPOVOY FJOR] D[7
Activity in 1994 Radionuclide Becquerel (Bq) Percent (%) Fission products 3H 4.8 x 1012 9.0 85Kr " 10 x 2 7. 1.4 '"Sr 1.2 x 1013 22 90y 1.2 x io13 22 "Tc ' 10 x 0 3. 0.0056 125Sb 9 JO x 9 5. 0.011 I29i 6 io x 5 6. 0.000012 l37Cs 1.3 x io13 24 l37Bam 1.2 x IO13 22 147Pm 9.2 x IO8 0.0017 151Sm 3.7 x 10" 0.69 IS5Eu ' 10 x 6 9. 0.018 Subtotal 5.3 x io13 100 Activation products Primary circuit medium 60Co 8 IO x 8 1. 0.00034 59Ni 6 IO x 2 2. 0.0000042 "Ni 2.3 x 10" 0.00043 20SPb 7 io x 9 2. 0.000054 207Bi ' 10 x 6 2. 0.0048 208B¡ 9.4 x 10' 0.0018 2l°Bim 8 IO x 0 5. 0.00094 Subtotal 9 io x 4 4. 0.0083 Actinides 238Pu » 10 x 5 8. 0.0016 23'Pu 3.4 x io10 0.063 240Pu 6.5 x 10« 0.0012
24,pu ' 10 x 8 2. 0.00052
2"Am 1.1 x 10" 0.00021 Subtotal iox 16 0 3. 0.067 Total iox 133 5. 100
34 TABL . ESTIMATEEXV D 1994 ACTIVIT LONG-LIVEF YO D RADIONUCLIDE RIGHE TH TSN I BOARD REACTOR PRESSURE VESSEL WITHIN THE REACTOR COMPARTMENT OF NUCLEAR SUBMARINE FACTORY NUMBER 601 DUMPED IN THE STEPOVOY FJORD [7]
Activity in 1994 Radionuclide Becquerel (Bq) Percen) (% t Fission products 3H 2.4 x io13 5.7 85Kr 3.6 x io12 0.86 '"Sr 5.8 x IO13 14 90y 5.8 x 10IJ 14 "Tc x 10' 5 °1. 0.0036 125Sb 2.9 x 10'° 0.0070 129, 7 IO x 2 3. 0.0000077 137Cs x io 13 36. 15 13 I37gam x io 0 6. 14 M7Pm 4.6 x 10' 0.0011 l5'Sm 1.8 x IO12 0.44 155Eu 4.8 x io10 0.011 Subtotal 2.7 x 10M 64 Activation products Control/compensation rods 152Eu 3.0 x IO13 7.2 IS4Eu 5.0 x IO12 1.2 Primary circuit medium <°Co 6.0 x IO8 0.00014 "Ni 7.5 x IO6 0.0000018 63N¡ 8 io x 7 7. 0.00018 20SPb 7 io x 6 9. 0.000023 207Bi " 10 x 5 8. 0.0020 208Bi 3.1 x IO9 0.00075 210Bim 1.7 x io9 0.00040 Core steel components '"Co 6.3 x 10" 0.0015 59Ni 7 io x 9 2. 0.0000069 63Ni 2.4 x IO9 0.00057 Core thermal shields "Co x IO 6 134. 11 59Ni " 10 x 8 7. 0.18 63Ni 6.9 x IO13 17 Reactor pressure vessel "Co " 10 x 5 2. 0.060 59Ni 1.8 x IO9 0.00044 63Ni 1.1 x 10" 0.026 Subtotal 1.5 x 10M 36 Actinides 238Pu 4.3 x 10' 0.0010 239Pu " 10 1. 7x 0.040 240Pu ' 10 x 3 3. 0.00078 24,pu 1.4 x 10' 0.00033 241Am 5.7 x 10s 0.00014 Subtotal 1.8 x 10" 0.043 Total H 10 x 2 4. 100
35 TABLE XVI. ESTIMATED 1994 ACTIVIT LONG-LIVEF YO D ACTIVATION PRODUCTN SI METAE TH L CONTAINER ENCASED SEPARATED LEFRIGHD TAN T BOARD REACTOR PRESSURE VESSELS FROM NUCLEAR SUBMARINE FACTORY NUMBER 538 DUMPED IN THE TECHENIYE FJORD [3]
Activity in 1994 Radionuclide Becquerel (Bq) Percent (%) Left board reactor
HC 1.3 >t 10' 0.028
MCo 5.3 >c 10" 12 59Ni 1.0 >< 10" 2.3 "Ni 1.8 >< IO12 39 Subtotal 2.4 >< IO12 54 Right board reactor
nc 1.1 x IO9 0.024 "Co 4.8 x 10" 11 "Ni 8.9 x 10'° 2.0 "Ni 1.5 x IO12 34 Subtotal 2.1 x IO12 46 Total 4.5 x IO12 100
2.5. DISPOSAL OPERATIONS
The following sections detail what is currently known about the disposal operations related to the seven nuclear submarine icebreakere th d san .
2.5.1. Submarine pressurized wate] r6 reactor, 5 , 4 , [3 s
With the exception of the right board RPV from submarine factory number 421 and the two PWRs from submarine factory number 538, all PWRs were dumped in their separated RCs. The SNF s removewa d fro lefe msubmarinf th t o boar V dRP e factory botnumbed an h 5 RPVr28 f submarinso e factory numbers 254, 260, and 538. The SNF remained in the right board RPV of submarine factory number 285, the right board RPV of submarine factory number 421, and both RPVs of submarine factory number 901.
Before disposal primare th , y circuit loop equipmend san l PWRal f o ts were washed, driedd an , sealed. However, indicatiothero n s ei n thaseale th t s were hermetic. Those RPVs containinF gSN were filled with Furfurol(F), a hardening compound based on furfural, prior to disposal. Before filling witeacV hhRP guidFurfurol(F)R CC e 0 tube3 e diametesth , m werm 0 e1 sealera breathed dan r hole drillewas d throug uppehthe r wal twoof l . Durin heated was breathe g fillingone , RPV r the ,hol ewas breathee on inle e d useth an t rs outletd e a hol uses th s ewa da . Onc procese eth completeds swa , each diametem 10m r breather hols cappeewa thicm d m witk 5 weldh2. .
The shallow waters of Abrosimov Fjord were used for four separate disposal operations. Separate fros mdRC submarine factory numbers 901 wer4 , 28525 ed dumpe,an 196n di t estimate5a d [1]m ,0 respectively2 d an , 4] , 1966n [l I . [4]separatem e m depth ,0 th 0 ,2 5 f frosC o dR m submarine factory s alsnumbeowa dumpe0 26 restimaten a t da [1]dm dept0 .2 f ho
36 time A th f tdisposal wers eo RC e e alloweth , flooo dt d thereby exposin gsignificana t portion externae th f o cavitiee lth surfacinternad d an san eacf V eo l hRP construction thosf so e RPVs without SNF to sea water. As such, sea water is assumed to have been within the left board RPV of submarine factory number 285botd an ,h 9 RPV2 d f submarinso an , 30 , e30 r factorfo 0 y26 numberd an 4 25 s years, respectively.
The right board RPV, with its SNF, was removed from the RC of submarine factory number 421, placed int osteea l collar-like support structure withi bargehula e f nth covered o l an , d with concrete. e concret thicke concretm Th m s abouTh .0 wa e 20 d telaye li betwee V r abov RP e oute ne th eth r innee th walV d r surfac an lbarge RP surfac th e les o f th en thicke o sf m s hul eo tha 0 wa l. n80
In 1972 barge th , e containin righe gth submarinf t o boar V dRP e factory dumpe s numbewa 1 dr42 in the Novaya Zemlya Depression at an estimated depth of 300 m [l, 4].
Both RPVs, their associated SGs theid an ,r associated primary circuit pumps, were removed from f submarino C R e eth factory placed numbean 8 d 53 intr osteea l collar-like support structure within the hull of a barge. The RPV lids and all penetrations into the lids were sealed by welding. No other protective barriers were provided.
The barge containing both RPVs, their associated SGs, and their associated primary circuit pumps f submarino e factor y e shallo s sunth numbe wa n ki w8 53 waterr f Techeniyso e Fjor n 198n a di t 8a estimated dept 35-4f ho externae [1]0m Th . l surfaces, cavities internad ,an l construction eacf V so hRP assumee ar havo dt e been expose watea se o drt sinc time edisposalth f eo perioa , abouf do years7 1 t .
2.5.2. Submarine liquid metal reactors [6, 7, 9]
The SNF remained in the two LMRs of submarine factory number 601. Before disposal, a number of actions were taken. The following is a summary of those actions:
) (1 solidified Pb-B s removewa i d fro l sectionm al undamagee th f so d right board reactor primary circuit except the RPV; (2) solidified Pb-Bi remained in all sections of the damaged left board reactor primary circuit; ) (3 control rods were permanently cores e fixeth n di , their drive mechanisms were removede th d ,an upper ends above the reactor lids were cut away; ) (4 channels containin CCRe gth s were filled with Furfurol(F regione th n i ) s abov Pb-Bie eth ; (5) channels containing the EPRs were filled with Furfurol(F) in their entirety and sealed by welding steel coveruppee th o srt surfac reactoe th f eo r lids; ) (6 cylindrical channel region thaV s texterna RP wer e th e o formet l d throug additioe o hth tw f no concentric cylindrical steel shells were filled with Furfurol(F); (7) SGs of the damaged left board reactor were filled with Furfurol(F) via the secondary circuit; thickm m 0 ellipti2 ) c (8 covers wereacf o volumee ehp th weldeto reactod e an th s d betweeo drli t n covere th lidsd thick m san maximua , m ,0 fille49 f mdo with bitumend an ; volumee th structuree th ) f so (9 s containin SGse gth , circulation pumps volumd an , e compensators, abovm th elliptim e elead-wate e e th 0 tan th levea f 24 kco o f t coverlo C R r e shieldsth werd ,an e filled with bitumen.
Overall f bitumeo 3 f Furfurol(F, o m som 3 0 nm wer25 e2 d e an r disposal) usefo preparo dC t R e . eth
In September 1981, over 13 years after the reactor accident, submarine factory number 601 was sunk in the shallow waters of Stepovoy Fjord at an estimated depth of 50 m [1]. At the time of her sinking, the hatches of the RC were open. As such, sea water has been in the compartment above the bitumen fille r ove yearsfo 4 r 1 r .
37 2.5.3. Icebreaker pressurized water reactors [4, 5, 8]
2.5.3.1. Reactor compartment
All SNF and the core barrel from the N2 reactor were removed from the three RPVs. Before disposal, the primary circuit loops and equipment were washed, dried, and sealed; and the ceiling of the RC was equipped with special pressure relief valves.
e icebreakerTh aboardC , R wits towee wa hth , d from Murmans Tsivolko t k ae th Fjor r fo d disposal operations Septemben O . wit, C 1967R 19 rh e thre,th e RPV s dumpe shalloe swa th n di w wate Tsivolkf ro a Fjorestimaten a t da directldm dept0 6 f yho fro icebreakee mth r throug bottoe hth m of the hull. The disposal site was approximately one kilometer from the site that was used for the cordamaged an e F barredSN reactor2 l N fro e mth .
2.5.3.2. Core barrel and spent nuclear fuel
consequenca s A coul accident V e TFC9 th disposee RP d21 f b e o 2 se N froth ,e f onlmdo th 4 y9 normaa n i f o l manner e remaininTh .core TFCth 5 ed 12 gbarre an s RPV2 l N fro e m,th hereafte r know Configuratios na , wernA e placed withi nreinforcea d concret shelS S d l containerean , hereafter known as Container B. Figure 9 shows a schematic cross-section of Configuration A. Container B consiste followine th f do g constructions:
primara ) (1 y inner stee insidm l5 line1. e diameterf o r thicm m k0 walls5 , 3.5d an ,height 5m ; secondara ) (2 y inner stee l regioe active line th th insid m n f i rneo 6 cor1. e diameterf eo m m 0 3 , thic heightm k4 walls2. ;d an , thicm m k0 lid10 ; ) (3 (4) an outer steel shell of 2.7 m outside diameter, 5 mm thick walls, and 3.55 m height; innen a r ) concret(5 e thicknessannulum m 5 f 545d so 51 an ;o -t thicm m k0 bottom10 a . ) (6
The voids within the cavity of Container B were filled with Furfurol(F), and the lid was secured by welding. Figur show0 1 e schematisa c cross-sectio f Configurationo withinA n Containe . OncB r e sealed, Container B was then moved to a temporary land storage facility constructed of concrete blocks. hale Afteon f yearsd r abouan e , Containeon t removes wa rB d fro temporare mth y storage facilitd yan place specialla n di y prepared caisson, hereafter know Containes na diamete m , aboar C r5 6. y da b r 12.5 m long steel pontoon. The walls and lid of Container C were constructed of type 8X18H10T SS of 12- and 18-mm thickness, respectively. The voids between Container B and the interior confines of Container C were filled with Furfurol(F), and the lid was secured by welding.
Like the icebreaker, the pontoon was towed from Murmansk to Tsivolka Fjord for disposal. During transit, a storm occurred in the region of the Kara Gate and the pontoon was temporarily lost due to rupture of the towing cable. The pontoon was subsequently found, secured to the towing vessel Lepse, towed an Tsivolko dt a Fjord Septemben O . , 196718 r pontooe ,th dumpes nwa d withie non kilometer of the site that would be used for the RC.
38 Stainless steel 08X18H10T
F/G . SchematicP . cross-section technicale oth f fuel channels cored an barrel from icebreakere th 2 N pressurized water reactor [4J.
39 FIG. 10. Schematic cross-section of the technical fuel channels and core barrel from the icebreaker N2 pressurized water reactor within Container B [4].
40 3. MODELLING STRATEGY
In order for the IASAP to provide an assessment of the radiological impact of the dumped marine reactors, source terms were require followine th r dfo g release scenarios:
) (1 Scenari "bese th : t oA estimate " discharge scenario. (2) Scenario B: the "plausible worst case" scenario is a situation where a disruption, e.g., collision or munitions explosion, causes a complete breach of the containment including the SNF from the icebreaker. This will be assumed to occur at the year 2050. (3) Scenario C: the "climate change" scenario refers to a major environmental disruption where global cooling followed by glaciation scours out the fjords. Subsequent warming would then release activity directly int Kare froa oth disrupted/crushee amSe th d reactor cores. This release will be assumed to occur at 3000 AD, one thousand years from now.
informatioe Th n presente Sectionn di (Method1 s3. Assumptions)d san (Mode2 3. , l Construction)d an , 3.3 (Corrosion Rates) are particularly germane to the development of the release Scenario A.
3.1. METHOD ASSUMPTIONSAND S
3.1.1. Initial assumptions
At the outset it was realized by the Group that a detailed and accurate prediction of radionuclide release rate s impossibleswa , taking into accoun smale th t l amoun f informatioo t n availabls it d ean accuracy and reliability. There were many factors that would be difficult, if not impossible, to model witdegrey han f confidenceeo s assumewa t I . d tharelease th t f radionuclideeo s fro dumpee mth d objects would be driven by sea water corrosion of the active material in the objects, but any further analysis quickly raised problems:
How doe watee sth r ente containmene th r t barriers? At what rate does corrosion occur? How does biofouling affect the corrosion rates? Is there enough oxygen inside the containment to support corrosion? Is there any flow through the containment barriers? How is corroded material removed from the containment? What fractio f materiao n l escapes fro containmene mth t whaa d t an trate ? fillee th ro d material w Ho s (Furfurol(F), concrete, bitumen) behav a water doew se n ei sHo ? their effectiveness degrade with time, following immersion in the sea? Does the build up of corroded material hamper further corrosion and removal of material? whao T t extencore th es i tdamagedoe w t affecho si integrite d th dtan containmente th f yo ?
IIASAn e aim e vieth th f "assessf o wo Po (t riske s th humao st n health... "examind "an e possible remedial actions....") it was decided to apply stated assumptions in order to obtain data that could be considere worse th e tb caseo dt , i.e., fastest rate f releaseso .
The assumptions used were:
l materiaAl particulaa f ) o l (1 r type (e.g., mild steel , Pb-BiSS , ) corrode fixea t da d base rate, modified by a correction factor. The choice of corrosion rate values is discussed in Section 3.3 (Corrosion Rates). correctioe Th ) n(2 factors use modifo dt bese yth t corrosion rates (BCRs), know factorsk s na , were dependent on the degree to which the containment barriers had been breached. They crudely model the slowing of the corrosion rate as the oxygen is used inside a volume with little contact restrictioe th d an opee waten a nth i nse o t r flow throug reactore hth .
41 l materiaAl s assume ) wa lreleasee (3 b environmene o dth t o dt s corrodedwa t s sooi a t s na . This avoids modellin e removagth l processes from corrosio e opeth no t nsea d immediatel,an y provide worse sth t case. (4) The filler materials were assigned a lifetime instead of a corrosion rate. At the time of dumping, the filler was assumed to be a perfect barrier to sea water; the filler material then degrades at a constant rate until, at the end of the assumed lifetime, the filler ceased to provide any kind of barrie wateo t r r ingres r radionuclidso e release. (5) The presence of fuel pin cladding was ignored as the extent of fuel pin damage was unknown.
validite Th reliabilitd yan thesf yo e assumption discusses si morn di e detai Section i l (Reliability1 n4. )
3.1.2. Route watesfor r ingress
e nexTh t modelline stagth n ei g procesascertaio t s wa smoste nth probable route r watefo s r ingrese activth o et s materia e dumpeth n i l d objects. This involved detailed discussion wite hth Russian member Groue th examinatio d f so pan structurae th f no l information provide Russiae th y db n Federation. Because of the high activities involved, initial studies were concentrated on those objects containing SNF. In all cases, particular points of weakness such as control rod channels were identified which would lead to water ingress to the SNF. The detailed strategies used for the submarine PWRs, submarine icebreakee LMRsth d an , r PWR coveree sar individuae th n di l Sections 3.2.1 (Submarine pressurized water reactors), 3.2.2 (Submarine liquid metal reactors) d 3.2.an , 3 (icebreaker pressurized water reactors), respectively.
3.1.3. Modelling methods
Once the route and timing of water ingress were identified, the process of modelling corrosion and releas f activeo e material could begin e tim t Th whicea . h water first made contact with each material was taken to be the start time for release. The BCR, modified by the relevant k factor, was then used to determine the rate at which material corroded and hence released to the environment. The model hence depende calculation d o rate factork t whicth e a e d th an sf h no materia s releasewa l y db corrosion. The following three sections illustrate how the effective corrosion rates were obtained to allow release rate calculations.
3.1.3.1. Calculation factorsk of
In most cases, the active material contained in the SGI was enclosed by an outer containment barrie rsubmarine sucth s ha e hul RPVr o l . Although breache allowind dan g ingres watera se f so , these containment barriers were assumed to provide a restriction in the oxygen transport into the reactor interior, slowing the true corrosion rate. This effect was modelled by introducing a factor kc, where:
are f breacao containmenn hi t c total surface are f containmenao t
In practice, the containment factor kc consists of the k factors derived for various RPV penetrations, e.g., breather holes, CCR tubes, primary coolant tubes, etc. The value of kc varies throug corrosioe hth n process exampler ;fo ,PWR e mosth f to s were dumpe separatea n di , whicdRC h was sealed with steel end plates. When these corrode and allow ingress, a step increase in the value of kj models the increase in size of the breach. The factor kc was scaled to give a value of unity once botplated hen s have fully corroded. Wher corrodine eth g materia s surroundei l mory db e thae non barrier factork e multipliee th , ar s d togethe givo rt cumulativea e factor.
42 Furfurol(F) has the effect of inhibiting the activity released from the SNF encased in it by the
multiplyine factoTh . k r g facto zers i k ro unti materiae th l dumpeds i l thed an ,n ramp unito st y over f
the fensuing 100 years:
0.0 It (t < 100 years) k = (7) f 1 (t > 100 years) where
t is the time from commencement of corrosion (a).
Thu effective sth e corrosion materia a ratr efo l insid Furfurol(Fea : )is filleV dRP
where
1 Xeff is the effective corrosion rate (mm-a" ), kc is the RPV containment k factor, e Furfurol(Fth s i kf factork ) , (mm-aR BC e 1)th . s i X
Concret uses froV encaseo wa dmt RP submarine eth e factory number 421. Again, with little informatio type th f concrete eo n no behavios it watera r o se n i rfactoa , r kconcrele, with values identical to that of k, was introduced to represent the effect of the concrete barrier. The RC of the dumped
submarine factorf y number 601 was partly filled with bitumen prior to dumping, and the model for release from this unit use dfactoa , agaib k r n wit hvalua e identica. f k o t l
3.1.3.2. Analysis of fuel pin geometry
Once sea water has entered the core region of RPV s with SNF, corrosion of the cylindrical fuel pins will occur, allowing releas f fissioeo n product actinidesd san release Th . e rate will depene th n do fuee th siz l f epino volume :th f fueeo l alloy release r yeadpe r wil givee b l y nb
where
volumreleasee F th SN s v(ti r yeaf eo )dpe r (m3-a"'), heighe fuee (m)n th th pi ls f i ,o t h effective th s Xi effe corrosion rate define s . (8)a , Eq , n di R is the initial fuel pin radius (m) t is the time from commencement of corrosion (a).
The activity of fission products and actinides released per year can be obtained from:
00) where
a(t) is the activity released per year (Bq-a"1) A(t) is the total activity of all the material, corroded and uncorroded, at time t (Bq) V is the total initial volume of the material (m3)
43 usind . (9)an Expressin activit, e gEq h , th d an term yn i R releas gf V so f fissioeo n productd san actinide showe : b be n o nt sca
2X (11)
wher l termeal previousle sar y defined.
3.1.3.3. Analysis of activation product release
The release rates of activated material from SGIs were calculated by applying effective corrosion rate simplifieo st d geometries representin structure gth question n i submarine th r Fo . e PWRd san icebreaker, the majority of the activation products came from the thermal shields and RPVs; these were modelled using plane geometry for simplicity.
LMRse case th I nth f bule ecompleo d th , kan x geometr thermae th f yo l shield ingresd san s routes required a more detailed analysis, and corrosion rates were modelled using a variety of circular corrosion geometries applie coree th ,o d t therma l shield SGsd san .
3.1.4. Calculation of activity release rates
Activity release rates were calculated usin gFORTRAa N code develope Royae th t da l Naval College Greenwich, running on a 486DX PC. For each component of the SGI, the actual corrosion rate (and hence activity release rate) was calculated using derived effective corrosion rates.
decae Th y fro originae mth l inventor isotopee eacr th yfo f ho s usecalculates dwa timestepr dfo s intfuturee oth . Onl Pu-Ae yth m decay schem s considereewa r daughtedfo r product accumulation.
The remaining thickness of each component at the timesteps was calculated using the effective corrosion rates, allowin face th t r thagfo t these rates will chang s levelea f containmenso t disappear. exampler Fo protectioe ,th greatlnV provideRP ye decreaseth y db corrosioe sth n contents e rateth f so . thermaS S e Th l shield instancr sfo e will generally outlas RPVse tth . Whe ncomponena corrodes ha t d through, it was removed from the calculation and adjustments made in effective corrosion rates from that time onwards.
effective Th e activity release rate eacn si h year were then calculated thicknese Th . f materiaso l which would corrode away in that year, as a fraction of the original thickness, multiplied by the activity still remaining, gives the release rate to the ocean.
It was assumed that the distribution of activity in activated components was linear. For the thicker steel components, the RPV wall for instance, the distribution will not be linear but information t availablno s againd wa ean assumptioe th , n pessimistie tendth o st earle th yn ci years.
Since model timescale sordee werth f hundredf ro eo thousandr so f yearsso radionuclidy an , e in the core inventories with a half-life of a year or less was ignored. Any release rate less than 1 Bq-a'1 was also eliminated from the output.
date writtes Th awa ASCIo nt I file graphinr fo s g using Microsoft EXCEL distributior fo d an , n to the IASAP Modelling Group. Files for the yearly release rates of fission products, activation products and the actinides, and totals for each group were prepared.
44 3.2. MODEL CONSTRUCTION
l SGIsal r , reactod Fo corrosioan e outeV th r f componentrRP o n surfac e th f o e s begins immediately free afte floodo ar t e rs dumping, RC albeie th s t a ,initiall sloa t ywa rate owine th o gt limited accese ope a wateth se na hulo t svi r l penetrations e firsTh t . releas s therefori e e almost
immediate, with activation products release rate f corrosioth eo t da carboe th f no n RPVe steeth f ;o l this release is dominated bCy60 o and so rapidly decays over the first twenty five years after dumping.
After that, the activity release pattern is primarily dependent on the way in which water enters the RPV containment and corrodes the SNF and active material inside.
3.2.1. Submarine pressurized water reactors
submarine th r Fo e PWRs fastese th , t ingress routes int core oth e are: breatheo tw e th r a holesvi describes a ) , (1 Section di n 2.5.1, intwalle oth f certai so guidR nCC e e typicath tubes r lFo RPV. , ther diameteem m wer 0 1 e r hole witV s hRP use e o filt dth l Furfurol(F), cappe thicm m dk 5 witweld2. ha . Pitting corrosion attacking these joints will offer the first access to the interior of the reactor and the SNF. These tubes will open at about 50 years from tim f dumpingeo . (2) via the primary circuit inlet and outlet tubes, which had been cropped and welded shut with a 10 mm thick steel plate cap; and ) (3 ingress thic intm controe m ok th 0 channelsd stee1 ro le l th capsa vi , . Pitting corrosion would enable ingress through both these route year0 s6 s after tim f dumpingeo . MucV h laterRP e th , itself will corrode away, leaving behind the SNF, the SS cladding and the thermal shields. By then however, there is little left of the SNF and the major component of the release comes from activatee th d shields.
FoU-Ae th r 1 alloy fuel assemblages fastea , r rat f releaseo s give morwa e som o nt th f eeo soluble and mobile radionuclides of the fission product inventory, typically 20% of the total activity.
3.2.2. Submarine liquid metal reactors
firse Th t barrie watea se o rt ingres bitumee th s si n which cover whole sth e SGI. This bitumen is assumed to degrade linearly in 100 years from full to zero effectiveness, modelled by the bitumen factor kb = 0.01t (t ^ 100 years). The fuel, and hence the fission products and actinides, is surrounded solidifief o abovm belom d t leas a e0 dan y wPb-Bt30 b i coolant multipld an , e . layerUsinSS f e so g th most pessimistic estimates of corrosion rates and thicknesses, the minimum time for water ingress to the fuel via corrosion directly from above or below is over 40 000 years. This timescale is much longer than that of ingress through the EPR tubes and ECTs; hence, this method will be assumed to be the primary means of release and is studied in more detail below.
Barrier breakdow LMRe th s considere f no swa thren di e main areas.
steame Th generators:) (1 por e containG th S t fue e remainine th s lth f froo lefe m% th t gboar20 d reacto totae th lf ro inventor (i.e.% submarine 10 , th f yo efore SGI th damagef m o n i ) d fuel pins, and both SGs contain activation products in the Pb-Bi coolant. (2) The thermal shields: the majority of the activation products are contained in the thermal shields around each core. These will corrode via the ECTs and from the outside. (3) The reactor cores: ninety per cent of the total fission product and actinide inventory is contained coreso tw e ,i nwhicth h will corrode initiall channelsR y EP fro e lated mth an , r fro outside mth e thermae coree oth th f s sa l shields corrode away. Activation products will als releasee ob t da the same rate from the Pb-Bi coolant and structural material in each core.
45 3.2.2.1. Corrosion of coolant and damaged fuel in the left board steam generator
The left board SG contains 10% of the total fission product and actinide inventory. Once the bitume Furfurol(Fd nan ) have degraded thed watea an n attacse ,n reacG rexterioca S e e ke th h th th f ro damaged fuel pins inside, allowing releas fissiof eo n product actinides d watea san se e r wilTh . l also attac Pb-Be kth , releasini coolanSG e th gn i tactivatio n products , showin diagraSG A .e e th g th f mo weld site shows SNFd i san , Figurn ni . 11 e
The SG is covered by bitumen and the secondary side filled with Furfurol(F). The primary side of the SG is filled with Pb-Bi coolant. The Furfurol(F) and bitumen both degrade in 100 years, allowing pitting corrosio weldsecondary/primare e th th f t nso a y header connectio commenceo nt e Th . fastest ingress route int regioe primare oth th f no y header assumes wa througe witb F o dt hSN h weld , the Furfurol(Fe A nth secondare th n i ) y header througd an , primare hth y header tube walls beloe wth header plate. The total time for water to reach the SNF by this method is 185 years. If the maximum IASAP corrosion rates (i.e., BCR + 50%) are used, the time drops to about 115 years.
Furfurol(F) Welds
Tube plate Damaged fuel pins Secondary system coolant channels (show parn i t only)
Pb-Bi coolant
FIG. 11. Cross-section through the left board steam generator of the liquid metal reactors of submarine factory number 601, showing damagedthe fuel pins, Pb-Bi coolant primarythe in circuit, and Furfurol(F) in the secondary header and channels.
The fuel pins are assumed to be lying on top of the denser coolant, and their whole surface area is expose availabld dwatea an se r r efo corrosion . Thus fuecorrosioe n th ,pi l appliee nb ratn eca o dt calculat activite eth , whicy releasSG he wilth n ei l ratfalF zero yeare0 t l SN froo 50 e sm th afte r first contact with water, i.e., at D + 685 years.
The activation products in the coolant will also begin release as the water breaches the tubes. In this case, the release rate will rise linearly as the surface area of corroding coolant increases with distance fro tube e coolanl mth Al . t will corrod118+ 5D y yearseb .
3.2.2.2. Corrosion of thermal shields
majorite activatioe Th th f yo n product containee sar thermae th n di l shields aroun coree dth . These shield penetrate e ECT4 sar 2 e sth whicy d b fille e har d with Furfurol(F wild an )l become open to sea water as the Furfurol(F) degrades. The ECTs provide the primary site for corrosion of the
46 shields. Corrosion of the shields also occurs radially inward once the bitumen and Furfurol(F) barriers have degradeexterioV RP intd allowe"stead e e oan rth dan th watema o t d se envelope.n o r " This 30 mm space between two external SS layers, filled with Furfurol (F) at time of dumping, is assumed prone b o wateo t et r ingres Furfurol(Fe th s sa ) must have been injecte suitabla a dvi e opening; once the Furfurol(F) degrades, this would allow water to penetrate.
Water ingress to the ECTs occurs once the elliptical cap and bitumen fail to be effective barriers, years0 10 Pb-Bd .an + S iCorrosioD S layert assume e a whice b the n th e n ca b snf hca no o d t commence corrosioe th ; treate e nlayere b ratth n homogeneoua thaf s desa ca o f o t f SS:Pb-Bo x smi i in the ratio 36:3, since most of the SS layers are 36 mm wide and all the Pb-Bi channels are 3 mm wide (see Figure 8). Assuming BCRs for SS and Pb-Bi of 0.02 mm-a'1 and 0.001 mnva'1, respectively, meae th homogeneoue n th rat r e fo thu n takee ca s b x 0.01s na smi 9 rnm-a"1. Corrosion will then expand froECT4 2 m s e show a seacth f Figurn ho ni . Corrosio12 e n will continu releaso et e activation products as the circles expand, until they meet after 3100 years (D + 3200 years).
Centrf eo ECT
Steam and water envelopes corroded away
FIG. Formation12. annulasan of from emergencythe cooling liquidtubesthe in metal reactorsof submarine factory number 601 at about 3200 years after dumping.
The release rate will now become complex as the circles merge; an approximation can be made
by simplifying the geometry to form an annuîus as shown in Figure 13. Corrosion and activation product release can then continue inward and outward at 0.019 mm-a" 1 until the BeO reflector or the corrosion attacking inward from the edge of the structure are reached.
During the formation and expansion of the annulus, corrosion will also be occurring at the outside edge of the thermal shields. The outer edge of the RPV will be corroded by sea water penetrating into the "steam envelope." This space has an inner edge at 900.5 mm from the core center, and corrosion will begin at that edge at approximately D +100 years, once the bitumen and Furfurol(F) have degraded.
corrosioe Th n fro steae mth m envelope will mee corrosioe tth n fro ECTe mth s after 1600 years, so the outer thermal shields will have corroded away by D + 4800 years, as shown in Figure 14.
47 Annulus
Corrosion continues inward meeo t t annulus
FIG. 13. Formation of an annulus from the emergency cooling tubes and corrosion from the outside of the thermal shields in the liquid metal reactors of submarine factory number 601.
Corrosion continues inwaro dt meet reflector
Outer thermal shields have almost disappeared outer FIG.Lossthe 14. of thermal shields continuingand corrosion inner ofthe shieldsliquidthe of metal reactors of submarine factory number 601 at about 4800 years after dumping.
48 e inneTh r thermal shields will continu corrodo et e unti reflectoe th l s reachedi 938r+ D 0 t a
years. The BeO reflector is assumed to corrode at 0.01 mm-a" 1 and so will corrode away after 6500 years (D + 15 880 years). Inside the reflector is a 10 mm thick SS layer, which will corrode away in years)0 watee 38 50 6 Th 01 . r year + wil D ls( then reac outside coree hth th f .eo
The fractional release rate of material from the thermal shields is shown in Figure 15.
Fractional release rate (lO-'/year)
2.0
Expanding circles Circles merge and from ECTs annulus forms
1.0 Shields externao t l ECTs disappear Outer edge of reflector reached
10 time (103 year)
FIG. 15. Fractional release rate versus time for activation products from the corrosion of thermal shields via the emergency cooling tubes (ECTs) and steam envelope of the liquid metal reactors of submarine factory number 601.
3.2.2.3. Corrosion of core
The elliptical shield that covers the RPV lid is 20 mm thick; pitting corrosion of the SS at 0.4 mm-a" 1 will be modified by the bitumen factor k = 0.01t for the first 100 years, giving a mean
corrosion rate of 0.2 mm-a'1 and so penetration of theb welds of the elliptical shield will occur after 100 tubeR EP s wild yearsan l Furfurol(F e havR Th .CC e usefuceasee a th e b n i )ld o t barrie thiy rb s stage, so water will be able to enter the tubes and attack the tube walls. The CCR tubes are filled with solidifie e cord th Pb-B eapproximatelo f t o (se p u p iabov eto m Figurm e eth 0 , wit e 6) 50 y hth remainde tube th ef rfilleo d with Furfurol(F) tubeR completele EP sar e Th . y filled with Furfurol(F), hence, onc Furfurol(Fe eth degradeds ha ) , water wiln l tubeca penetratR d EP bottos e an e th th o emt f o attac tube kth ethick wallm m , 5 scoreallowine wall e 0. nex th e Th .o s t ar g penetratio occuo nt r after 25 years. Hence, 125 years after dumping (D + 125 years), the water has reached the core via the EPR tubes.
Once tubes wateR corrosioreachee s EP ,th rha e corth e a dth envi will sprea concentricallt dou y from thre e eacth f eho ingress sites int Pb-Bie oth shows a ,nearese Figurn Th ni . t 16 efue l pine sar distancaboum m 6 t channelsd e ro fro e mth ; corrosion will progres Pb-Be th t sa i corrosion rate until
49 the fuel is reached, so activity release from the core for the first 600 years (D + 725 years) is solely activation products from the Pb-Bi. After 600 years, the expanding circles of corrosion will reach the first fuel pins and fission products and actinides will begin to be released. The solidified Pb-Bi is assumed to have a corrosion rate of 0.01 mm-a"1; the U/Be/BeO fuel is assumed to have a corrosion rat f 0.00eo 1 mm-a"steee th ld structura1an ; l materia rata l f 0.0eo 2 mm-a"1. Treatin fuelee gth d area homogeneoua coolant s % a 36 meae d th ,an nS corrosioS s mixturfuel% % 10 54 , f neo rats ei
(0.01 x 0.36 + 0.001 x 0.54 + 0.02 x o.l) = 0.006 mm-a"1
BeO reflector
Fuel pin matrix and Pb-Bl coolant _____ ~v^^S^SS^^¿^^^>í^^^^^f" \ ^^^•••^^^ Expanding circle corrosiof o n m Emergency protectio d nro
Q Control or compensation rod
CirclesFIG.16. corrosionof expanding from emergencythe protection channels rod liquid the of metal reactors of submarine factory number 601.
corrosioe Th n throug thermae hth l shield years0 38 s 6 wil,1 l+ reac core D edge th y heth b f eo by which tim expandine eth g circle channelR s EP fro e mth s will havfrom m e 8 corrode9 o t t dou the EPRs; at this stage the situation will be as shown in Figure 17. At this point, the corrosion from the outside begins to release fission products and actinides. The corrosion from the outside will meet corrosiouR p EP wit e hth n after 767 years 0 geometre 000 yearsTh 4 . 2 + , y i.e. D the t ,a n becomes comple expandine th s xa g outside circleth d esan corrosion merge arrangemene :th shows i t Figurn i e 18. For simplicity, it is assumed that the material release rate then remains constant until all core material has corroded away, so the core will corrode away in 10 600 years, i.e., at D + 34 600 years. release Th e rat f materiaeo l fro core m th shows e i Figurn ni . e19
50 Emergency protectiod nro Expanding circle of corrosion Control or compensation rod
FIG. 17. Circles of corrosion from the emergency protection rod channels of the liquid metal reactors of submarine factory number 601 at about 16 400 years after dumping. The stainless steel thermal reflectorBeO shieldsthe outsideand corethe have corroded away, allowing -water attackto the outside core.the of
FIG.Circles18. corrosionof centered emergencythe on protection channelsrod merging with corrosion from the core exterior of the liquid metal reactors of submarine factory number 601 at about years000 24 after dumping.
51 Fractional release EPR circles rate (10'5/year) and outside corrosion merge
4
3 Corrosion Circles begint sa expand outside of core from EPRs
10 20 30 time (103 year)
FractionalFIG.19. release rate of fission products actinidesand from corrosion emergencythe via protection (EPR)rod channels liquidcoresthe the of of metal reactors submarineof factory number 601.
3.2.2.4. Corrosion of control rods
made ar f EuBeo R s indicate A 6CC materiaa , d an d R learlier witEP e h th ,cerami c properties. Little informatio availabls ni e about corrosion rate f EuBso watern i 6assumes i t I . d here tha rate tth e is similar to that of other ceramics such as BeO, and a rate of 0.01 mm-a"' is assumed. Water enters th yearse5 channel12 ;+ henceD t sa , corrosio rode th sf no wil l commencd an t thaR ea tEP date.e Th CCR have radii of 10 mm and 8.5 mm, respectively, and the CCRs are surrounded by approximately 1 mm of Pb-Bi coolant within the channel. Both types of rod channel were filled with Furfurol(F) prio dumpingo rt CCRe th alse t sar o bu , surrounde solidifiey db d Pb-Bi fillin channee gth heigha o t l t f approximatelo abovm m coree e0 th y,50 which will prevent water reachin rode gth s from abover fo at least 50 000 years. Hence, only the EPRs will be attacked by sea water within the timescales of interest ,years 5 beginninEPRe 12 wilradiuo Th s m .+ slm d D havcorrod0 san t 1 ga ea e away after 1000 years (D + 1125 years). The resulting release rates of activated Eu will be modified by the change in size (i.e., surface area) of the rod; this mechanism will be identical to the modification to fuecorrosioe n th pi l n rate.
3.2.3. Icebreaker pressurized water reactors
Although the geometry of the structure containing the SNF and the core barrel from the N2 reacto complicateds i r , modellin release gth e rate relativels si y simpl ebees [21]ha nt I assume. d that no activit releases yi d fro whole mth e unitContainef o d , corrodeuntili s e ha lth rC d off, followey db the lid of Container B, using pitting corrosion rates applicable to the material of each container, acting weldsd li e onth . Once seawater penetrates Containe five thermaS th S e , B r l shields 2 withiN e nth PWR core barrel are assumed to corrode away on both sides. The results are discussed in more detail in Sectio 2 (Sensitivity)4. n n notca et her e thabu e ,containeron eth t se yea opeth rn i n2305 , approximately 340 years after dumping.
52 The damaged fuel from the N2 PWR is assumed to comprise UO2 pellets, 5% enriched, 4.5 mm diameter, 10 g-cm"3 density, carried in Zr-Nb alloy cladding. As soon as the fuel is in contact with sea water immediatn a , fissioe th f o ne releasproduc% 20 actinidd f eo an t e inventor assumes yi d froe mth fuel grain boundaries. The remaining fission products and the actinides are released through dissolutio fuef no l grain pessimistia t sa 10"* c0 7 rat3 g-cm'^d" f eo 1 [22] fuer Fo .l pinm m s wit5 4. ha diameter and 10 g-cm"3 density, this equates to a constant dissolution rate of 0.0011 mm-a"'. Once expose seawatero dt fuee th , l will take some 2250 year corrodo st e away.
e same RPVth th , r es modellinFo RC dumpe e th n i d g philosophy whic s appliee hwa th o dt submarine PWRs was used (Section 3.2.1). The same breather holes, control rods and primary pipework caps were welded onto the entry points to the RPV.
differencee Th s were:
(1) The RC was sealed with a steel plate over the upper hatch. The welds were attacked by pitting corrosion and yielded in 60 years, allowing access for RPV corrosion to begin. s thesA e ) earl(2 y reactor boltedd ha s , s assumeinsteawa f weldem do dm 1 df o flanges p ga a , l primaral heaV d betweean RP dy e pipewore jointth th n f o s k couplings. This allowed immediate, if limited entry for water into the RPV, to attack the internal components. Furfurol(Fo N t int RPVse pu oth ) s (3 .wa )
The result was that first release of activation products would be 60 years after dumping, with rates increasin yearg5 s later pipeworV , wheRP e nth k caps fall off.
3.3. CORROSION RATES
The modelling strategy relies heavily upon corrosion rates of containment and barrier materials in the dumped SGIs. Such materials include the metals used in their original fabrication, and fillers applied prior to dumping to inhibit metallic corrosion and provide additional barriers to activity release into the sea. Strictly speaking, therefore, in this context the term corrosion encompasses all forms of material degradation this i t si mechanisd an , m that liberates radionuclides throug breakdowe hth f no radioactive materials and of the barriers encasing them. This process is considered under the heading f eaco h material type late thin i r s section.
Factors affecting the lifetime of materials exposed to sea water include: dissolved oxygen, temperature , salinitypH , , sulphates, pressure marind an , e life (biofouling) . o [23HeiseSo ]d an r provid edetailea d revieinfluence th f wo thesf eo e factor ranga n materialsf o se o , with data derived from analysis of dump sites in the Atlantic and Pacific. However, little data is available on the behavio f materialo r relativeln i s y shallow Arctic waters derivo T . e corrosio nIASAPe rateth r sfo , comprehensive review availablf so e source references were mad supporn ei thif to s work, with notable contributions from Carter [23]o [22 Heised So ]an .d an r
In all cases, realistic but conservative values of material corrosion rates were determined; their values are summarized in Table XVII. The rates lie towards the high end of recorded data in order to err towards the pessimistic side of the spectrum. For modelling purposes, corrosion rates established in this way are designated BCRs and filler lifetimes. Where there is a paucity of data, e.g., in the case of the Pb-Bi eutectic, nuclear fuels, and certain filler materials, best use has been made of available informatio judgementd nan t shoulI . notede db , however, that containmen factorsk t describes a , n di Sectio (Mode2 n3. l Construction), effectively inhibit BCR restrictiny sb g pathways through barrier materials.
To account for the uncertainties in corrosion rates employed in the models, values generally 50% eithe BCRe r th sidapplie e f sensitivite sear o th n di y analyses describe Section di (Resultn4 d san Analyses), which also has an assessment of the reliability of the data used. The ranges in the parameters studie sensitivite th n di y analyse alse sar o include Tabln di e XVII.
53 3.3.1. Metals
3.3.7.7. Steels
Not surprisingly, much greater data is available on the performance of steels in sea water than any other material. The range of values, however, is wide, and related to the factors cited earlier. Applyin principlR BC e gth e detailed above, realisti conservativt cbu e corrosion rates were determined for long term immersio a watese n ri n unde o distinctw r t mechanisms: bulk corrosio d pittinnan g corrosion, with the latter applied under conditions of high stress in the heat affected zones of welded areas. As can be seen from Table XVII, for simplicity two types of steel were assumed: SS for specific SGI components and the icebreaker SNF containment (Containers B and C), and mild steel (low alloy and or carbon steels) for RPVs and submarine structure. To account for the possible accelerating effect of biofouling, corrosion rates were doubled for steels with outside surfaces fully expose seae th .o dt
3.3.1.2. Europium hexaboride
This materia CCRe uses i th lLMREPRe d n di th san f so s installe submarinn di e factory number 601. The Group could find no published data on the corrosion behavior of this material, and assumptions had to be made for modelling purposes. Although pure Eu is a reactive metal, in ceramic B form it is expected to be stable and relatively impervious to corrosive attack. The BCR of
0.06 1 mnva" s selecte e assumptiowa 1th n do highey s unlikelnan thawa e t rb i t thao yt n e thisTh . importance of this parameter in the LMR model relates to the criticality issue; and its value is shown in Sectio (Reacto2 n5. r Criticality e insensitivb o t )criticalite th n i e y analysis thereford f an ,o t no e significance with regard to accuracy.
3.3.1.3. Lead-bismuth eutectic
The Pb-Bi eutectic is used as a coolant in the LMRs installed in submarine factory number 601. In the absence of information on the performance of this material in sea water, the model assumes the
same corrosion rate as Pb. Heiser and Soo [23] report uniform corrosion rates of 0.008 to
1 f 0.0o 1b P mm-a"r fo 0.01 therefor s i R 3 BC t mm-a"unreasonablea eno d an , 1 . Ther clearls ei y some uncertainty ove performance rth Pb-Bwatee a th se f n eri o eutectic compared with pur . HoweverePb , until further data becomes availabl formere th consideres i n eo t i , drepresentativa e th n i e e ratus o et model at this stage.
3.3.2. Reactor fuels
3.3.2.1. Uranium-aluminum pressurized water reactor fuel
The sole source of information on the corrosion of U-A1 alloy fuel used in PWR SGIs is that
of Yefimov, et al. [24], which reports the results obtained from3 immersion in laboratory sea water of spent fuel fro icebreakee mth r Siberia. hige vien th I hf w o corrosio nmm-a"3 0. rat f eo 1 established sample foth r e fuel, this valu soluble modee uses th ei th r n di fo l e fission products fraction onlyA . rate an order or magnitude lower, 0.03 mm-a"1, is applied to the insoluble fraction. Due to the unknown extent of damage to fuel cladding, its effect as a barrier is ignored in the model.
3.3.2.2. Uranium-dioxide icebreaker pressurized water reactor fuel
Not much is known of the behavior of UO2 fuel pellets in sea water. A review of available data by Carter [22] enabled an estimated grain dissolution rate of 30 x 10"7 g-cm"2-d"' to be determined for 3 UO2 fuel pellets. For fuel pins of 4.5 mm diameter, of regular geometry and 10 g-cm" density, this equate dissolutioa o st n rat f 0.001eo 1 mm-a' same Th 1.e reference summarizes wor Shoesmity kb h and Sunder [25 whic]n i e releasth h f activito e y from oxide a wastfue n i l e repository vauls i t considered. This suggests:
54 ) (1 rapid releas f volatileeo s fro claddinfuee d mth an l g gap, (2) leaching of fission products from fuel grain boundaries, and ) (3 slow releas f radionuclideeo s fro fuee mth l matri resula s xa f fue o t l grain dissolution.
However, because of the difficulties and lack of information in estimating the rates of radionuclide release from grain boundaries modee th , l assume instantaneoun sa s release fro fuee mth l surfacd ean grain boundaries onc watea ese r comes into contact conservativwita fuele s hth A . e estimate, this si totae assumeth lf o fissio accoun% o dt n20 r producfo t actinidd an t e inventor fuele th .n y i Similarly, since the extent of damage to the Zr-Nb alloy cladding around the fuel is unknown, the model assumes that cladding is ineffective as a barrier to sea water and radionuclide release.
3.3.2.3. Uranium-beryllium liquid metal reactor fuel
This fuel is assumed to be in a ceramic form, manufactured from U-Be alloy sintered into BeO, and therefore relatively imperviou corrosivo st e attack. Again absence th behavioe n i , th f dat eo r afo r of this material in sea water, the Group consensus was to assume a similar order of dissolution rate to 1 that used for UO2 fuel grains. Accordingly, a rate of 0.001 mm-a" was modelled. As for other reactor fuels, the barrier afforded by fuel cladding is ignored because of the unknown extent of damage prior to disposal.
3.3.3. Fillers
3.3.3.1. Organic materials
Little informatio s availabli n e abou e lonth t g term behavio e filleth f r o rsubstance s when immersed in sea water and under irradiation. Bitumen is known to become brittle and crack below room temperature s Furfurol(FA . a patente s i ) d material, information abou s compositioit t d an n behaviot availablno Groupe s th lonwa rd o et an ,g term case y behavioan ,n i difficul , is r predicto t t . mixtura knows I i tfollowine e b th o f nt eo g constituents: epoxy resin, aminé type solidifier, mineral filler, shale distillat furfuryd ean l alcohol [26] effectivn A . e lifetimyear0 50 quotes i se f o eth n di Whit r thie fo Boos ] material[1 k .
In the absence of reliable data on the performance of Furfurol(F) in such environmental conditions, a conservative lifetime of 100 years in the radiation environment is assumed in the model and supportethis swa preliminara y db y evaluation [27]. Hencecalculatioe th r source fo , th f no e term, assumes iwa t dtime thath dumpingf et o a t fillere th , s were fully effectiv barriers ea wateo st r ingress and radionuclide release at the time of disposal, but quickly began to degrade through shrinkage, embrittlemen crackind an t becomd gan e ineffectiv yea0 e 10 rafte e lifetimeth r .
3.3.3.2. Concrete
Concret s usefroV encasewa o mdt RP submarine eth e factory number 421knows i t I . n that concrete is, in almost all conditions, porous to water. However, with little information on the type of concret behavios it waterr a eo se n i rsimilaa , r effective lifetim othee thath o et f ro t filler assumes i s d modelse inth .
55 TABLE XVII. CORROSION RATES AND FILLER LIFETIMES USED FOR IASAP STEAM GENERATION INSTALLATION MODELS.
Material „ . . Best corrosion rates (mm a" ) Sensitivity ranges (mm a'1 )
Stainless steel Bulk 0.020' 0.01-0.1 Pitting 0.50 0.25-0.75 Mild steel Bulk 0.0751 0.038-0.11 Pitting 0.166 0.08-0.33 U-A1 alloy 0.032 0.015-0.045
UO2 0.0011 0.0001-0.01 U-Be ceramic3 0.001 0.0001-0.01 Pb-Bi coolant 0.01 0.001-0.1 Boronated graphite 0.01 0.005-0.05 Europium hexaboride 0.01 0.001-0.1 Bitumen4 100 year lifetime5 50-500 Furfurol(F)4 yea0 10 r lifetime 25-500 Concrete4 100 year lifetime 50-500 Biofouling factor 21 1-3 steelsr Fo 1 , bulk corrosion rate outen so r surfaces were increase accouno t facto a 2 y biofoulingr f db o rfo t . 1 For fission products with low solubility and actinides; rate used for soluble fraction of fission products is 10 times the slow rate. ' Uranium beryllium allo berylliun yi m oxide ceramic matrix. ' Filler materials were give lifetimna preferencn i e corrosioa o et n rate. ' Lifetim perioe th s di e after whic fillee longehth o n r r provide physicasa l barrier. 6 The ranges are not necessarily those given in references but chosen for sensitivity analyses purposes.
3.4. RELEASE SCENARIOS
Throughout this section, calculated event roundee sar five th e o dyeat r date.
3.4.1. Submarine pressurized water reactor releases
The overall Scenario A and C release rates for all dumped submarine PWR SGIs are shown in Figures 20 to 25. For example, Figure 20 shows the release of fission products, actinides and activation products, togethe totam r su lwit release hth e rate, from unifro1 time m42 t th dumpinf eo g in 1972 unti yeae th l r 3710 steel e whe th e las f nth ,o t wit associates hit d activation productss ha , corroded away.
encases wa Uni1 concretn di 42 tdiscusse s a d ean d earlier slowla , y degrading lifetim0 10 f eo year s assumeswa thir dfo s containment barrier similaA . r lifetim assumes Furfurol(Fe ewa th r dfo ) encapsulatin concrete SNFe th gth s A . e barrier becomes mor mord ean e porous, activation products are released from the outside of the RPV. Then the breather hole into the interior of the RPV is corroded open in the year 2005, allowing fuel and interior SS corrosion to begin. The other RPV penetrations and barriers begin to open up in the year 2035, shown by the peak in release rates for the fission products, 370 GBq-a"1 and actinides, 0.2 GBq-a"1. Coupled with the continuing steel corrosion, the total peak release rate is 370 GBq-a"1.
56 Date (year) 1800 2300 2800 3300 3800 1.00E+03 Fission Products Total 1.00E+02
g 1.00E+01 O
S 1.00E+00 to .3! tu 1.00E-01
1.00E-02
1.00E-03 Fission Products Actinides 1.00E-04
FIG. 20. Novaya Zemlya Depression, unit 421, total activity release for Scenario A.
After this e fissioth , n product release rate falls away quickl e morth s ea ymobil e atome ar s released, unti yeae th l r 2500. After that rate follows th ,ei lese sth s soluble atomfissioe th n si n product inventory. The actinide release follows this slow degradation rate; as time goes on, it is seen however, that the overall rate of activity release is dominated by the activation products. Between the years 2500 and 3360 release th , GBq-a'2 ordee 0. th e f f ro rato 1s .ei
In the year 3360, the RPV has finally disintegrated, replacing the slow effective rate of corrosion
interioe fo structure th r th fastee f th o r y r eb shorBCRa d tan , pea fission ki activatiod nan n product release is seen, totalling 2.5 GBq-a". 1 After that, the fuel rapidly corrodes and is corroded away by the year 3385.
internae Th claddin V thermae th RP l d gan l shield finalle sar y corroded yeae awath r y 3710yb ; in this last period, the release rates are dominated by the long-lived activation products at 1.6 GBq-a"1.
Figure 21 shows the same situation for unit 421, but for the overall Scenario C release. All original processes detailed above occur until disintegration of the unit in the year 3000. Then, a total releases i q GB d than 0 inti Kare o 63 tfa o th yearaSe , from what remain unite th f . o s
sho3 2 same Figurewd th an e scenario2 s2 unite th Abrosimo n i sr sfo v Fjord. These diagrams summatioa e ar releasee th f no unitsse froth l m, al wit withoud han t SNF takd an ,e into account their different dumping dates. The units were dumped in their RCs, which represents an outer containment barrier againd an ,unite th , s havwitF interion hSN ea r protectio f Furfurol(F)no .
In overale Figurth , el22 releas Scenarir efo peae th k, oA releas e comeyeae th rn si 2040 , with totaa f 270o l 0 GBq-a"1. After that fissioe th , n product contributio release th o nt e decreasee th d san activation products begin to dominate the total release. By the year 2350, the rate is on the order of 3.3 GBq-a"1 until the year 2690, when the RPVs have corroded away, allowing the fuel release peak to be seen again. The fuel has corroded away by the year 2740, and the remaining steels by the year 3075.
Figur show3 e2 overale sth l releas Scenarir efo Abrosimor fo oC v Fjord. This shows thae th t fuecorrodes ha l d awa onld activatioe yan yth n products GBq0 88 t ,a , remai releasee b o nt d inte oth yeae th rn i 3000Kar a aSe . Comparin r Abrosimofo 3 g2 Figured van Fjor1 2 s de th show w ho s concrete barrier of unit 421 has extended the life of the fuel material into the glacier scenario.
57 Date (year) 1800 2300 2300 3300 3600 1.00E+03 Fission Products Total 1.00E+02
g 1.00E+01 ü
OOE+01 $ 0 (0 "55 E t 1.00E-01.
1.00E-02
1.00E-03 - Fission Products Actinides 1.00E-04 FIG. 21. Novaya Zemlya Depression, unit 421, total activity release for Scenario C.
Date (year) 1800 2000 2200 2400 2600 2800 3000 3200 3400 1.00E+04
1.00E+03 Activation Products Total
(0 1.00E+02 . cr ffl Ü 1.00E+01 ta0> (0 9) v 1.OOE+00 (£. & 1.00E-01
Î 1 OOE-02 Fission Products 1.00E-03
1.00E-04
FIG. 22. Abrosimov Fjord, units 254, 260, 285 and 901, total activity release for Scenario A.
Date (year) 1800 2000 2200 2400 2600 2800 3000 3200 3400 1.00E+04
1.00E+03 Activation Products Total
'S- 1.00E+02 ffl O 1.00E+01
1.OOE+00
1.00E-01
1.OOE-02 Fission Products 1.00E-03
1.00E-04
AbrosimovFIG.23. Fjord, units 901,254,and total260,285 activity release Scenariofor C.
58 Figures 24 and 25 show the situation for unit 538 in Techeniye Fjord. Here, the unit was dumped without SNF in 1988 and only the activation products in the steels corrode away, releasing their activity. Starting at 0.8 GBq-a'1, the release rate falls to 0.06 GBq-a"1 at the year 2655, when the RPV finally corrodes away. Afterwards, only the cladding and the thermal shields are left, and by the year 3075, the havo yto e corroded away finae Th l. releas eGBq-a"4 rat0. s ei 1. Figur show5 e2 e sth peak in the year 3000 due to the glacier disintegration of the remaining material.
3.4.2. Submarine liquid metal reactor releases
Overall Scenari releasoA e rate showe sar Figurn ni . Froe26 dumpine mth 1981n gi , initially, no active material is expected to appear due to the hull and bitumen barriers. When it does start, the initial fission product and actinide release is less than 0.0001 GBq-a"1 at the year 2105 when the fission product and actinide inventory in the corroded left board SG starts to appear. Corrosion of the outer surface activatethe of s d RPV thasby t time contributes abou GBq-a"8 t 1.
yeae th ry 2180B , fission product actinided san s fro damagee mth undamaged dan d reactor cores joi release nth e tota e streath ld releasman ordee GBq-a"5 yeae th ef th f o r rrato y 3000s B ei 1. e th , rise in release rate of the SNF and thermal shields of the reactors, caused by the expanding circles of corrosion exceed release decayo t th e d fale sedu th an l, rate rises unti yeae th l r 5200 whe ECTe nth s merge to form an annulus. The release rate then varies as the shields external to the ECTs corrode away by the year 6800 and the left board SG loses all its SNF by the year 7500. By this stage, the release rate has fallen to 0.07 GBq-a"1.
e releasTh e rate remains steady unti yeae th l r 11400 whe thermae nth l shields corrode away. Release rates then continue at 0.004 GBq-a"1 from the SNF and the Pb-Bi, as the reflector is attacked by corrosion smalA . l rise occur yeae th rn si reflecto1840 e th s 0a r corrode watese awath s d i r yan able to attack the outside of the core. Release then continues at a rate of 0.007 GBq-a"1 until all the SNF has corroded away in the year 36580.
e overalTh l releas r Scenaris show i efo yeae e pear uni1 th Figurfo n rTh i n 60 tki o C . e27 ordee th f 210300f o ro 0s 0i GBq. Not change eth time-scalen ei s betwee. n27 Figured an 6 s2
Despit predictioe eth n fro modee mth l thafirse th t t release wilt occuno l r until 21se earlth t n yi century sampla , f seaflooeo r sedimen Join e s taketh wa ty nb Norwegian-Russia n Expert Groun pi 1993 showin isotopeu gE s l52 154 d Ehighed Euan u an r than expected level 137f s o sampl e Csth [28]s A e. closs sit ewa submarin o et e factory numbe suggesteds r 601wa report e i ,th n i t that this activity might have originated from the LMRs.
Although this is a possibility, the first predicted pathway to the damaged SNF is via the welded body of the SG. To open this path requires the corrosion of two SS welds of 20 mm thickness and the degradation of the Furfurol(F) inside the SG and the bitumen surrounding the whole SGI. However wilF lSN have th , emixtura f fissioeo n product actinidesd san , including 155 'd 37ECsuan . structurV channelR RP e CC th e d f so an R EP e th n i As i large, t witCs ro bu h n , sourc Eu f eo itself. However, this pathway is not predicted to leak until the end of the 21st century.
The mode l observatione resultth d san s conflict; this migh becausee b t :
(1) The RC was contaminated by 132Eu, IS4Eu, and '"Cs before disposal. sedimene th n i s tC comed an u s E fro e mTh anothe ) (2 r source. (3) The bitumen and Furfurol(F) barriers have been breached and corrosion of the SG, reactor cap, and CCR cap welds has been much faster then anticipated; the SGI is in fact leaking.
59 Date (year) 1800 2000 2200 2400 2600 2800 3000 3200 3400 1 OOE+01
Activation Products only
-5 IT 100E+00
OJ 0)
OOE-01 > 1.
1.00E-02 FIG. 24. Techeniye Fjord, unit 538, total activity release for Scenario A, activation products only.
Date (year) 1800 2000 2200 2400 2600 2800 3000 3200 3400 1.00E+02
Activation Products only •. 1 OOE+01
O
1 ooE+oo
1 OOE-01
1 OOE-02
FIG. 25. Techeniye Fjord, unit 538, total activity release for Scenario C, activation products only.
Date (year) 0 5000 10000 15000 20000 25000 30000 35000 40000 1 OOE+02
1.00E+01 Total eg 100E+00 O" 0 ( 1.OOE-01 0> 1 0OE-02
1.00E-03
!> 1.00E-04
1.00E-05
1.00E-06 Fission Products Actinides
1 OOE-07
FIG. 26. Stepovoy Fjord, unit 601, total activity release for Scenario A.
60 If (1) and (3) were the case, it is thought that the other isotopes from the LMR would have been observed in the sediment sample. It is suggested that the observations be explained by contamination coming from another source in Stepovoy Fjord.
3.4.3. Icebreaker pressurized water reactor releases
Total release rate fissior sfo n products, actinides activatiod ,an n products into Tsivolka Fjore dar shown in Figures 28 to 33. As the two dumped components are resting close to each other in the Fjord, the release rates shown are from both units combined.
3.4.3.1. Fission product and actiniae release
Figure 28 shows the overall release for Scenario A. Figures 29 and 30 break this down to illustrate the fission product and the actinide releases separately.
sho0 3 Figurepea e wd th kan generate9 s2 containere th s pontoode a th n sfinallo e nar y breached in the year 2305. Five hundred GBq of fission products and 1600 GBq of actinides are immediately released to the Kara Sea from the cracks and porosity of the damaged fuel. In the following year, the rat f releaseo e revert calculatee th o st d corrosion oxide th rat f eeo fuel; fission produc actinidd an t e release rates are 1.7 GBq-a"1 and 5.7 GBq-a"1', respectively. The fuel slowly corrodes away and the activity of the fuel itself decreases; in the year 3300, the release rates for fission products and actinides are 0.05 GBq-a'1 and 2.7 GBq-a'1, respectively. The fuel is finally corroded away by the year 4570.
3.4.3.2. Activation product release
activatioe Th n product release rate show Figurn ni mors i 1 e3 complicate various da factorsk s come into play on the two dumped units. Again using Scenario A, the hull surrounding the RPV compartmen firss i t t breache yeae th rn di 2030 , commencing outer corrosiowalV RP l n release, with a small contribution from material inside the RPVs through gaps with are assumed to exist in the bolted flanges release Th . e rate GBq-a"start3 2. t s a fivt 1bu e, years late hule rth l fails completeld yan
the release rate jumps to 23 GBq-a". Caps on the main coolant pipes, control rod tubes, and other
penetration yeae s th fairn i 2055l , increasin release gth eGBq-a"1 rat1 4 o et . Unti yeae th l r 2300e th , release rate falls, largely due to decay of 1 the 60Co component of the total activation activity.
yeae Ath t r 2305 containmene pontooe th , th n Containere o breaches th ni B f d o t d an dan sA the thermal shields of reactor N2 are exposed to corrosion at the full BCR. This shows as an increase
overale inth l releas eGBq-a"2 rat2 o et 1. These shield corrodee sar d yea e awath e r y th y266b d 5an rate drops to 3.6 GBq-a", 1 as the only material left is in the RPV assemblies.
The RPVs within the RC corrode away at the year 2700 and their cladding by the year 2795. This expose remainine sth g materia thermae l th lefn i t le shieldth d an corrosioo st R fule BC th l t na
release rate jumpGBq-a"0 6. rate o st eTh . falls away graduall finae year0 th 25 r l se y fo unti th l al l activated material i1 s corroded away by the year 3050. From Section 3.4.3.1 (Fission product and actinide release), there is still an appreciable amount of fuel material left to corrode away after the activated steels have corroded away.
3.4.3.3. Icebreaker total release rates
As indicated earlier, Figure 28 shows the overall release for Scenario A. The initial release is dominated by the activation product material from the RC, with the peak between the years 2035 and
GBq-a"orde2 e 2 205th f n o r5 o . Release rates drop unti yeae th l r 2305 when ContainerB d an sA
1 holding the damaged SNF open and the fission product and actinide peak of 2100 GBq-a" 1 dominates the release in that year.
61 Date (year) 1800 2000 2200 2400 2600 2800 3000 3200 3400 100E*04
100E»03 iooe»oz
1 OOE»01 iooe»oo 1006-01
100E-02
100E-CO
1 OOE-04
100E-05 Actmides
100E-08
100E-07 Stepovoy27. FIG. Fjord, unit 601, total activity release Scenariofor C.
Date (year) 1800 2300 2800 3300 3800 4300 4800 1 OOE+04
1 OOE+03 Activation Products OOE+01 . •5 2 Total m 8 1.00E+01 (O Sí 1 OOE+00
è" 1 OOE-01
< 1 OOE-02 Fission Products 1 OOE-03
1| OOE-0VUU.-W4^ 1 mujjL i in- —— II-I-ILI i i i m^.i...... Ei^aj,,...... ,,,,,, -,...... mjj.ti i j r .iPH'T Tsivolka28. FIG. Fjord, icebreaker reactor compartment fueland container, total activity releasefor ScenarioA.
Date (year) 1800 2300 2800 3300 3800 4300 4800 1 OOE+04
1.00E+03
•g. 1.OOE+02 m Fission Products 1.00E+01 %m S. 1.OOE+00 oQ:) & 1.OOE-01
< 10OE-02
1 OOE-03
1.OOE-04 FIG. 29. Tsivolka Fjord, icebreaker reactor compartment and fuel box, fission product release for ScenarioA.
62 The decrease in the total release rate at the year 3050, when the last of the thermal shields corrodes away markes i ,rate th e s GBq-a"4 dropda 3. o st 1. This leaves onl fuee continuo yth t l o et yeae corrodth ro t 4570 n eo .
Figur show2 e3 overale sth l releas r Scenarie fo munition A . oB s acciden r otheo t r mishas pi postulated in Tsivolka Fjord, near the dump site in the year 2050. This accident is assumed to breach Containers B and C on the pontoon, and also break off the RPV lids in the RC. The contents of all unite thee th sar n expose BCRso dt , wit containmeno hn t barriers.
e initiaTh l corrosion rat s i observede , thee accidenth n t produce a releass e peaf o k 110 000 GBq-a"1. Due to the increased corrosion rates, the activated components are now corroded away by the year 2565 (as opposed to the year 3050 for Scenario A) and the fuel corrodes away by the year 4320 (as opposed to the year 4570 for Scenario A).
Figure 33 shows the overall release for Scenario C. The situation up to the time of glacial disintegratio yeae unite th th r n f si 300n o assumes 0i tha e r Scenarib fo to d t , the oremainine A nth g active materia totaa ( l f 300o l 0 GBq releases i ) environmene th o dt than i t t year.
3.4.4. Kar totaa aSe l release rates
Figure 34 shows the Scenario A total radionuclide release rates from all the dumped PWR SGIs dumped int Kare oth a Sea exclude.s i Uni 1 60 t d from this diagram becaus muce th f heo extended time frame for its corrosion. The 3000 GBq-a"1 peak from the release of PWR fuel between the years 2030 and 2050 210e th , 0 GBq-a"1 peak fro fuee mth l containe yeae th re sharn i r2305th d p an fal, l when thermal shields corrode away and cease to contribute to the total release rate are noticeable.
Figur show5 e3 same sth e releas e Kar totath a r aSe fo el Scenari . HereoC release yeae ,th th rn ei 3000 totals 4500 GBq-a"1.
4. RESULTS AND ANALYSES 4.1. RELIABILITY
The main sources of errors in the predicted release rates are as follows:
) (1 informatio steae th mn no generating installation structure materialsd san ; ) (2 radionuclide inventory; ) (3 value f BCRsso d an ; (4) degree of pessimism used in the models.
4.1.1. Informatio steae th mn no generating installation structure materiald san s
reliabilite Th informatioe th f yo submarine th n no icebreaked ean I structureSG r materiald san s centers on the reactor cores, the thermal shields, the RPVs, and their associated support structures. Core details for the submarine LMRs is the most comprehensive, with information on the fuel rod materials ,materialR dimensionsEP d locationsd an an s d pitch R d overalan , an CC ,, l materials distribution icebreakee th r Fo . r PWRs informatioe ,th essentialls ni y limite fuematerialse d th ro l o dt , dimensions, and pitch and the configuration of the control rods (CCRs) and scram rods (EPRs) within the cores botn I . h cases core th , e detail substantialle sar y more reliable tha core nth e detaile th r sfo submarine PWRs, where assumptions were made. Informatio thermae th r nfo l RPVe shieldth d s san is reasonably complete with regard to the materials, dimensions, and locations of one with respect to another. The most detail is associated with the submarine LMRs. Information on the support structure reactoe th r sfo r cores, thermal shields t bee RPVd no ns an , providesha significany an o dt t degree. As such, it is the area where the largest number of assumptions have been made and where the data is the least reliable. With respect to future concerns of potential reactor criticality and remedial actions, the lack of details on the support structures may impact future decisions.
63 Date (year) 1800 2300 2800 3300 3800 4300 4800 1.00E+04
1.00E+03
-2. 1.00E+02 m o 1.00E+01 Sä «J .3! 1 .OOE+00 4) CU & 1.00E-01 Actinides < 1.00E-02
1.00E-03
1.00E-04 FIG. 30. Tsivolka Fjord, icebreaker reactor compartment and fuel box, actiniae release for Scenario A.
Date (year) 1800 2300 2800 3300 3800 4300 4800 1.00E+04
1.00E+03
-5. 1.00E+02 CD O 1.00E+01 Q) m CO £ 1.00E+00 cteu .^ 1.00E-01 Activation Products OOE-01 < 2
1.00E-03
1 OOE-04 FIG.Tsivolka31. Fjord, icebreaker reactor compartment and'fuel container, activation product release for Scenario A.
Date (year) 1800 2300 2800 3300 3800 4300 4800 1.00E+06
1.00E+05
1.00E+04 . 5 Activation Products OJ 1.00E+03 . CD Total 0> 1.00E+02 £ S. 1.00E+01 Actinides 0) K . 1.OOE+0g 0
'Í5 1.00E-01
1.OOE-02 Fission Products 1.00E-03
1.OOE-04 FIG.Tsivolka32. Fjord, icebreaker reactor compartment fueland container, total activity releasefor Scenario B.
64 Date (year) 1800 2300 2800 3300 3800 4300 4800
1. OOE+03 4 Total
-5 1.00E+02 mCT 0 ^.^ 1.00E+01 p— 0) ^^Ü^~Ü lg n T 1.00E+00 V u•*-— —— 'S -. Activatio— ~- n Products û£ /v .£• 1.00E-01 •^— —^ ^ •^ Actinides à L < 1.00E-02 Fission Products 1.00E-03
1.00E-04
FIG.Tsivolka33. Fjord, icebreaker reactor compartment fueland container, total activity releasefor ScenarioC.
Date (year) 1800 2300 2800 3300 3BOO 4300 4800 1.00E+05
1.00E+04 Total: all activity I 1 OOE+03 o 1.00E+02 n £ 1.00E+01
¿f 1.00E+00
1.00E-01
1.00E-02
1.00E-03 FIG.Kara34. Sea, unitsall except 601, total activity release Scenariofor A.
Date (year) 1800 2000 2200 2400 2600 2800 3000 3200 3400 3600 1.00E+05
1.00E+04
1.OOE+03
1.00E+02 cöe>
1.00E+01
1.00E+00 Total activitl :al y
100E-01 KaraFIG.35. Sea, unitsall except 601, total activity release Scenariofor C.
65 4.1.2. Radionuclide inventory
e reliabilitTh e radionuclidth f yo e inventory center detaile reactoe th th n f so o r cores, their associated reactor operating histories core th ed modelan , scalculationse useth n di reliabilite Th . f yo the core details have been described in Section 4.1.1 (Information on the steam generating installation structures and materials). Suffice it to say that the submarine LMR and icebreaker core details are the most reliable. Reactor operating histories obtained from RRCKI [3, 8] and IPPE [7] records for the submarines and icebreaker, when compared with those from Russian Navy records [10], were found to differ by no more than a factor of 2 and on average yielded a comparative ratio of 1.0 ± 37%. As such, the reactor operating histories used in the Group calculations were considered reliable. Core models used in the Group calculations have been discussed in Section 2.4 (Radionuclide Inventories). Models used for the submarine LMRs and icebreaker PWRs represent the actual configurations; models used for the submarine PWRs are assumed the same as that of the icebreaker. In addition, when compared to an independent estimate of the radionuclide inventories prepared by SIAE [10], the Group inventories were determined to represent the best estimate for the IASAP. With respect to potential radionuclide release, the submarine LMR and icebreaker radionuclide inventories are deemed the most reliable, while the submarine PWR radionuclide inventories are deemed the least reliable.
4.1.3. Value bese th tf corrosioso n rates
reliabilite Th value e BCRe th th f f yo so use dependens di t upo type nth materiaf eo l considered. Degradation rates are best estimates based on currently available data, some of which is scant for material sPb-Be sucth s ha i eutecti fuelU thein d i s can r various allo oxidd yan e ceramic formse Th . different type f materiasub-dividee so b y ma l s followsda :
(1) Structural materials. Extensive literature searches have been undertaken for common structural materials such as SSs and mild steels, and the BCRs employed are believed to be on the conservative side of the range of values examined [23, 29]. Soo [30] reviewed values determined by the IASAP, and considered them reasonable for the bulk and pitting corrosion rates of SS. Sufficient data exists on mild steels (low alloy and carbon) to justify their use and reliability. (2) Pb-Bi coolant (LMR only). In the absence of data on Pb-Bi, the best available alternative was the known corrosion performance of lead in sea water. However, there is a possibility that the behavior of the Pb-Bi eutectic may be different, and some doubt must therefore exist on the reliabilit thif yo s data. Give significance nth thif eo s material with regar likels it o dt y inhibition of fuel corrosion in the core, and indeed with regard to its thermal expansivity and the possible presenc coref e voideo th n ,i s further informatio thin no s materia essentials i l . (3) Fissile fuel. Research data on the corrosion of U fissile fuels in sea water is very limited. The best available data has been used for the dissolution rate of the icebreaker oxide fuel. (4) Filler materials: Furfurol(F), bitumen, ana concrete. Alexandrov, et al. [27] confirmed the lifetime of Furfurol(F) as 100 years, in a project conducted during the period of the study. Carter [22] helpe assigninn di g value concretr sfo bitumed year0 ean 10 wells st a withoun t a ,bu t better information as to their makeup, these must remain as best estimates.
4.1.4. Degree of pessimism used in the models
In orde produco rt e pessimisti realistit cbu c value radionuclidf so e release, several assumptions were made: material Al ) l (1 corrode immediatels di y availabl environmente th o et regardes i d an , releases da d for the purposes of the IASAP models. Making this assumption negates the need for analyses of solubility and transport mechanisms through the containment, and immediately demonstrates thaIASAe th t P release rates mus pessimistice b t practicen I . coursef o , , corrodee mucth f ho d material will be both heavy and insoluble and will remain within the containment, reducing the true release rates. This beefacs ha tn discusse contexe th n d i reactof to r criticality, whers wa t ei
66 assume d insolubls corrodee thath wa f t F o would abou deSN % an bottoe d80 te th fal th o t f ml o RPV. Retentio f corrodeno d material could further reduc release eth epresencs it rat s ea y ema slo prevenr wo floe watef th tw o r throug reactore hth , retardin preventinr go g further corrosion, and preventing further release of material. ) (2 BCRs use r releasdfo e rate calculations were chose givo nt e fast release rates, adding further pessimis modelse th mo t . Also, BCR materialI s applieSG e th so d t wer e constant, whereas some studies indicate that true corrosion rates will decrease with timcorrosioa s ea n resistant layer forms on the surface of material under attack. ) (3 Fue claddinn pi l bees gha n ignore IASAe th n di P release rate models. Inclusio claddinge th f no , even when the fuel has been substantially damaged, will retard or reduce the release rates. (4) The filler materials were assigned a lifetime and rate of degradation, at the end of which they cease barriere b o dt corrosioo st transporr no f corrodeo t d material. Ther littls ei e information available on the long-term behavior of the filler materials in sea water and in a radiation environment fillef i ; r lifetime longere sar release ,th e firshundre e ratew th fe t n si d years wile b l less. In reality, the filler material will not simply disappear at the end of its lifetime, and may still retard or prevent release of material. 4.2. SENSITIVITY Byverits y nature IASAthe , P release rate mode is lsensitiv chosethe to en corrosioor n degradatio ne stee th rate r l fo componentse barrie th r fo r materialsd an s . Althoug extensivn a h e literature study was undertaken and advice sought from materials specialists from the contributing countries, in order to provide the best estimates, nevertheless, steel corrosion rates in the ocean can hav widea e sprea f valuedo barried san r effectivenes t welno ls i defineds . usinn ru ratese rangge a mode e examplb n th Th f A n .eo ca l er mor migho var o e t f eo e y on b t the parameters, e.g., usin gmaximua pittinS mS g corrosio yea0 n 50 rvalue a Furfurol(F d an , ) lifetime, opposes a IASAe th o dt P recommended lifetim years0 illustrato 10 T f . eo e n this modee ru th ,s wa l using three set corrosiof so n rate barried san r lifetimes ratee lifetimer Th s.o s were chose uppee th t nra and lowe rpublishee endth f so r advisedo d values (see Table XVII) compared an , IASAe th o dt P recommended values. Figure 36 shows the result of one such run, using the Abrosimov Fjord units and the release of Scenario A. Using the maximum rates, all the units have corroded away by the year 2370. With the IASA ratesR time PBC ,th extende s ei yeae th rslowese o d3075t th r Fo .t rates unite ,th s survive until about the year 5000. As a general principle, there is a factor of 3 between release rates for the highest degradation rates and the IASAP rates and a factor of 30 between the IASAP rates and the slowest e maximu th e ratespea r Th GBq-a"0 fo k . e 00 mIASA th e 2 pea 1 rater th 1, fo ks Pi s rates i s 2700 GBq-a'1, and the peak for the slowest rates is 90 GBq-a'1.
Date (year) 1800 2300 2800 3300 3800 4300 4800 1.00E+05 , Maximum Degradation 1.00E+04 Rates IASAP Degradation Rates
1.00E-01 Minimum Degradation Rates 1.00E-02 FIG. 36. Sensitivity analysis, Abrosimov Fjord: Scenario A total release rates with maximum, minimum and IASAP rate values applied.
67 5. REACTOR CRITICALITY
radionuclide Th e release rates predicte corrosioe th y db n model too accouno changkn y an f o te in corrosion rate r fissioso n product inventorie possiblo t e sdu e criticality. However, since very little beefued e th ha l n f coresdecides e o useth wa n t di i , investigat o dt possibilite eth criticalitf yo y being achieve corrosios da n progressed reactoa f .I r core could achieve criticality, this could potentially have affected the predicted radionuclide release rates in two ways: e energTh y ) release(1 fissioy db n would caus increasn ea temperaturn ei coree th n ,ei leadino gt accelerated corrosion rates and hence release rates; (2) Criticality over a long period would produce a fresh inventory of fission products, some of which would have short half live t presen presene sno th n i t t inventories.
Three scenarios were reviewed to investigate the possibility of any of the reactor cores achieving criticaa l state:
(1) Corrosio a larg f no e proportio controe th materiad f n o ro s substantiall l ha F l beforSN e yth e corroded away. This could hav same eth e effec corn o t e reactivit s controya withdrawad ro l l durin greactoa causd r an star , e criticalityup t . (2) The corrosion process or certain forms of attempted remedial action cause a structural change within the core, such as the SNF falling to the bottom of the RPV or control rods being displaced, resulting in some or all of the SNF and core material reaching criticality. ) (3 Ingres f wateo s r intcore oth e causin n increasga neutron i e n moderation sufficien causo t t e criticality.
vera s yA rough estimate marina , cor R usefus it e ePW f wil o l d llif reacen e e onchth e about 40% of the fuel has been used, after which point it must be refuelled. For a core with a fuel load of 235f o Ug ,k thi0 5 sremaining g equatek 0 3 o st f lif.o Thi ed fuesen l load give indication sa w ho f no much fuel is required for a critical assembly of a lattice of fuel pins in water. Hence, for all core types, followine inth g analysi massa f fue so f les o l swilg considerethak e b l0 n3 unable b o d t achievo et e criticality, while a mass of more than that may achieve it.
Analysis of the core burn-ups, assuming 1.25 g 235U per GW-d burn-up rate, showed that none
of the submarine PWRs with SNF had used more than about 3.5 kg Uof; hence a typical dumped SGI fuellecorR ePW dcoul assumee b d contaio dt f fissilo ne th g abou k ef o 7 material 4 t % 94 r o , 235 original (star f lifeo t ) fuel loade righ Th n submarin.i t boarR LM d e factorys numbewa 1 60 r undamaged and so could be assumed to contain around 89 kg of 235U; the left board core lost 20% of its accidente resulloaa th s assumes da f wa o t o s ,contai o . dt kg 1 n7
During normal operation, the reactors would have been operated with EPRs fully raised to allow safe emergency shut down of the reactor and some the CCRs at a height which maintains critical condition. This operatin heighd gro t must have been abl muc s vara o et y possibls yhb a alloo et r wfo poison changes; the optimum rod height for the control rods would have been about half core height, withdrawn% 50 r o practicn I . CCRe ehavth y sma e been operate groupn di different sa t height alteo st r neutroe th n flux profile acros cor e compensatd sth ean fuer efo l burnu lond pan g term poison changes. No data was available on submarine PWR rod layout, but the LMRs had 3 EPRs and 10 CCRs.A typical operating configuratio LMRe th f no s would have allowed criticality wit CCRn hte t arounsa d withdrawa % EPR3 50 e th s fulld an ly withdrawn totae . Thith lf d o smas ro equate % f o s 38 o t s removedabsorbe corefollowin62% the the or ,in rIn . g analysis assumeis it , d tha a cort e could possibly becom absorbee th f eo criticar% materia62 f i l l removede werb o et .
5.1. CONTROL ROD CORROSION
All the corrosion models studied by the IASAP use the control rod channel as one of the primary ingress corewatea e routeth se r ;o r t thu sfo controe sth l rod subjece sar corrosioo t watey nb soos ra n
68 as water ingress occurs. For this scenario to allow criticality, the effective corrosion rates of the SNF anrode dth s mus suce materiad b t ro h f corrodn thao t ca leasla t% 62 et away befor fuee eth l mass becomes too low to allow criticality, i.e. below 30 kg. The rods and SNF are assumed to be subject same th eo t containmen filled an tfactors k r BCRcomparisoF o s , SN s d controe wilan th e b f ld no ro l used in the following analysis. However it should be noted that the time values obtained will be incorrect effective th s a , e corrosion rat ordero e tw wil f magnitude so b l e slowe owinR r BC tha ge nth containmene th o t factorsk t .
In the case of the submarine PWRs, all controls rods were assumed to be constructed of SS, enriched with neutron absorbing material. The BCR of the U-A1 alloy is 0.03 mm a". 1 If, as discussed above assumes i t i , d(i.eg k tha 7 . 94%4 t time th f disposals leff fuee)t o o a twa l f fueo ,lg thek 0 n3 wil lefe b l t whe fuee nth l pins radiu . Thidecreases sha mm s 4 wil o t ld takm fro yearm 3 e3 m5 t sa the BCR of U-A1 alloy (or 1100 years taking into account the k factors). As discussed above, during normal operatio totae materiad th nlro f o abou % l coul38 t insertee d b core e th n d;i henc criticalitr efo y occuro t contro e morr , o th abou materiaf d e% o ro l t62 l must corrode withi years3 n 3 contro e .Th l rods corrodo havt allo o t % o eradiua ws 62 e, requireabouf so mm 0 1 t maximusa m contro radiud ro l s after
33 years of 6.2 mm and hence a minimum corrosion rate of 3.8/33 = 0.11 mm a"1, which is not much more tha e maximunth m accepted IASA corrosioS PS a"m . 1 m However n1 0. rat f eo , usine gth minimum IASAP value for the U-A1 corrosion rate of 0.015 mm a"1, the assumed minimum critical wilg lefk e b l 0 t mas corrosio3 S afte S f years 4 so e 4 r Th .n ratd ero requiree th f o allo o dt % w62 materia corrodo t l year4 4 n thes ei si n 3.9/4 whic1 0.0a" 4= withis hi m 8 m IASAe nth P rangS S f eo corrosion rates. This example suggests that criticality by corrosion of control rods alone is possible
if the ranges of values of the IASAP corrosion rates are accepted as realistic. e icebreakeTh usef 0.001 R o fue a'UO d.m R PW l r 1 However 1m wit BC ha , ContaineC r
containe totae th l20.r f 2do o fue onlg 6reacto2 % lk N froy 60 e mth r whic assumes hi o to d e herb o et low an amount for a critical assembly to form.
reactorR LM highld e ha s Th y enriche e dumpe th burnup w l lo dal df d fue O cores.an l , they contain by far the greatest amount of SNF. However, the ten CCRs were encased in Pb-Bi for 500 mm abov coree ePb-Bth e channel;th R i CC fillwated e onlsth n an rca y corrod frot ei m above. Hence eth water would hav corrodo et e through approximatel Pb-Bf o m i m befor 0 y50 e attackin EuBe e gth th n 6i channelsR CC , whic r Pb-Bihfo woulR . UsinBC d e e IASA0 yeartakth gth t 00 least a e a s 0 P5 t assumptions EuB3 e th ,6 EPRs will begi corrodo nt e afteyear5 12 r s when water penetrate elliptie sth c shield, bitumen and EPR channels, and will have corroded completely in around 1000 years. However, corrosion of the 3 EPRs alone is insufficient to cause criticality, as the 10 CCRs will still be fully inserted in the core. As discussed in 3.2.2.3, the IASAP corrosion model of the LMRs predicts that water will corrode outwards from the EPR channels, removing SNF and Pb-Bi coolant. Eventually the water will reach the CCRs, which are about 160 mm to 173 mm from the EPRs. This will occur at yeararoun0 00 s7 d1 afte r dumping, after whic rode hth s will corrod aboun ei t 1000 years amoune .Th t of fuel remaining after 18 000 years can be approximated by taking the ratio of the area of the expanding circles to the total area of the core, as shown in Fig. 18; this ratio is around 0.5, i.e. the corroded fuel represents about half of the total initial amount. Hence at the stage when the CCR material corrodes away, about righhale fuee th lefs fth i tn l i tboar d lefte reactorth ,n i abou d % ,an t40 since approximately 20% of the SNF was transported to the steam generator in the accident. This equates to around 44 kg of fuel in the right board core, and 35 kg in the left. Hence the IASAP corrosion model suggests tha yeart aroun0 a t 00 s8 dafte1 r dumpin remainine gth g control rods could corrode away, leaving sufficient 235U to form a critical assembly.
5.2. STRUCTURAL CHANGES
5.2.1. Corrosio stainlese th f no s steel supporting structure
Alteration corn si e structure cause corrosioy db n are theiy b , r very nature, comple difficuld xan t to quantify. An example of a scenario which may lead to criticality would be disintegration of the
69 Furfurol(F) filler afte assumes it r d lifetim subsequend an e t structurcorrosioS S e th f eno holdine gth SNF, causing the fuel pins to fall into the bottom of the RPV. If sufficient U were left in the fuel pins, and the configuration of the fallen pins formed a suitable arrangement in a small enough volume, a critical assembly could result e possibilitTh . f thio y s would also e depenamounth d n o dan t arrangement of other materials such as water and corroded control rod material, and would be extremely difficult to model accurately. The rest of this section attempts to determine whether enough U allocoulo t possibilite V wdth RP arrivbottoe e th th criticalityf n ei yf o m o doet t attempbu , sno o t estimate the probability of the U forming a critical assembly. There is little information available about the configuration and thickness of the SS supporting structures; an estimated thickness of 3 mm will
be used in this analysis. Again an assumption is made that 47 kg Uof 235 remains in the core. In order to leave 30 kg in the pin provido st criticaea l mass pine th , s must only havf SNFo g .e k losThi7 1 t s will occue th f i r fuel pin radius has decreased to 4 mm when the SS structure corrodes away.
However, this does not take into account the SNF already corroded. Although the IASAP models assume that all corrosion products are released to the environment, and so provide the most pessimistic release rates, in practice a large fraction of the SNF is likely to be insoluble and remain in the RPV, collecting at the bottom as loose material. The IASAP corrosion model assumed 20% of the fission products in the alloy fuel were more soluble, leaving 80% effectively insoluble. If 80% of the U is assumed to collect in this manner, then an additional amount is available to provide the critical mass. If the IASAP submarine PWR Greenwich FORTRAN model is run with the SS corrosion rate set at the maximum value, and the U-A1 alloy rate set at the minimum, the 3 mm thick SS structure corrodes awa abouy yb yeae th t r 2700 they , radiufuelosine n B . n th pi decreases l mm gsha 2 3. o t d m from m5 60% of the mass of the fuel pin. This could still provide well over 30 kg of SNF at the bottom of the RPV, as shown in the following calculation:
(1) 60% of the mass of the fuel pins has corroded - 80% of this is insoluble, so 60% * 80% = 48% is lying loose on the bottom of the RPV; (2) 40% remains in the fuel pins, which fall to the bottom, giving a total of 88% of the mass of the fuel pins at the bottom, and; stare th f lif o tf eo fue% l 83 loa= s di % 88 x % 94 t o burns , no s up i t F SN e th f o % 94 r o g k 7 4 ) (3 bottome th lefn o t , which give massa . f 41.kg so 5
It should be noted that even without the failure of the SS supporting structure the corroded U coul bottoe de core th th fal o f t m,l o wher t coulei d collec compaca n i t t mas se les b whic sy hma affecte presence th y b d neutrof eo n absorber scontroe sucth s ha l rods. Hence, assumin corrodee gth d ablUs i falo et l pas remainine th t g core structur remnante th Furfurol(F)e d th ean f so possibilite th , y criticaa f o l mass being formed from corrode cannoF d SN rulee b t d out.
This analysis doe t takno se into accoun claddiny e SNFan t th , n go whic h will increase eth probability of enough U remaining in two ways:
claddine th (1) g will slo corrosioe wth more o SNFs th , f en o wil l remai nstructurS S onc e eth e corrodes; (2) the structural integrity of the fuel pins may depend on the cladding: if the cladding is corroded, the SNF in the form of pellets or small pins may fall to the bottom. The cladding is unlikely to be more than 1 mm thick, so will corrode away before the SS structure, and hence could allow bottoe th fal o t o mt l F soonerSN e th , where morremainsF eSN .
The presence of neutron absorbing material will increase the critical mass. In the worst case, however, such materia corrody ma l e rapidl ceasd yan affec o et time core th th etn e i estimate e th r dfo rod corrodeo st .
70 5.2.2. External corrosion
Corrosion to the external parts of the RPV that weakens its structure has the potential to cause the whole assembly to collapse or topple. The consequences of such an occurrence are difficult to predict. In the worst case, the control rods could be displaced out of the SNF and a critical mass achieved, possibly fast enoug causo ht e prompt criticalit situatioa y( n wher increase eth reactivitn ei y is sudden enough to produce an uncontrolled rise in the fission rate and a rapid, possibly explosive,
releas f energy)eo . Submarin I supportineSG huld an l g structure heave sar y gauge mildt a steel d an , an assumed thickness of 25 mm and BCR of 0.08 mm a" (takin1 g into account the effect of biofouling and corrosion from both sides) could fail after 300 years. The probability of the RPV toppling and the control rods coming out of the core will depend on the configuration of the other structures within the reactor compartment and core.
Any proposed remedial actions which involve linin wil movinr V go l RP hav e gtakth o e t e into accoun possibilite th t e resultinth f yo g structural change causing criticality mose th : t catastrophic scenari turninV oRP woule g th ove e db r during transpor controe shorth o t d elan rods fallinf o t gou the core under gravity.
5.3. INGRESS OF WATER
e submarin e th casI th nf o e e PWR s initialli d icebreaker F an s SN y e surroundeth , d with Furfurol(F), which can be assumed to degrade in 100 years, allowing water to enter the core and surroun fuee dth l pins. This will increas reactivite eth coree th f ;y o however norman i , l operatioe nth control rods will have been designed to produce a safe shut down condition, and with all rods present there is no risk of criticality. Combined with control rod corrosion and compaction of SNF under its own weigh botto e RPVe th , th t f a tm o however presence th , f wateeo r will increas probabilite eth f yo criticality.
LMRe Th s were never designe operato dt e wit wateha r moderator reactore :th operation si n were intermediate (i.e., unlike in a PWR, the neutrons did not have to be fully slowed down to thermal energie perpetuato st chaie eth n reaction), wit actinneutroe e hB th s ga n moderator Pb-Be th f I .i were to be replaced by water, a substantial increase in reactivity would result as the higher slowing down power (a measure of the efficiency with which a material slows down neutrons) of the water would produc mucea h larger thermal neutron fluhencd xan e greater fissio nhighle ratth n ei y enriche. dU
To allow water ingress, however, the Pb-Bi must corrode. This was assumed to occur at the I ASAP BCR of 0.01 mm a"1, ten times the rate of the SNF corrosion, so corrosion would remove the Pb-Bi preferentiall replacd yan t witei h water above Th . e discussio f controo 1 corrosiod 5. ro l n ni n conclude e lefb d t y e fuethaoncth ma l t ef arouno wate s % corrodeha r50 d d awa l controyal d ro l material after about 18 000 years. There are other factors which this analysis did not take into account: preferentiae th ) (1 l corrosio Pb-Bie th f moro no s , e fuel pins wil estimatelefe % b l t 50 tha e nth d above; presence th ) f wateeo (2 r will further increas reactivitye eth , leadin leso gt s fuel being requirer dfo criticality than was necessary during normal operation, and; Pb-Be th i ) corrosio(3 n fastesa rats eha t IASA a"1, m Pm whic valu1 0. hf e o woul d reduc time eth e reaco t CCRe hth arouno st d 2700 years arounr o , yeae dth r 5000.
Henc possibls i t ei e that owincombinatioe th o gt f corrosiono CCRse th presencf e no th , f eo water amound an , f fueo t l remaining, tha s sooa t s abouna e yeath t r 1300 e reacto0th r coren i s submarine factory numbe coul1 60 r d slowly achieve criticality.
The IASAP radionuclide release model also assumes that because the Pb-Bi mixture has a negative coefficien thermaf o t l expansivity volume ,th decreaset Pb-Bf eno o s remainiha d dan closn si e structuresS S contacd an F , tpreventin SN wit e hth g water from seeping throug gap y comind han s an g
71 in contact with the SNF. Prior to dumping there appeared to be no gaps [9]. However, the LMRs were dumpe coln di d Arctic water furthed an , r thermal contractio havy nma e occurred Pb-Be th f I .i volume s reducedha , making water ingres n thii s s manner possible, each n fuecoulpi l d eventualle b y surrounded wit hlayea f wateo r r which could caus elargea r increas reactivityn ei . This would then allow other method f reactivitso y increas occuro et :
) (1 corrosio Pb-Be th f nio layer aroun CCRe dth s leadin earlieo gt r corrosio CCRse th f no ; (2) corrosion of the SS supporting structures causing the SNF to fall into a smaller volume at the bottoRPVe th f mo , and; ) (3 corrosiocladdinS S fuee e th th l f f rodgo no s could fueresule th l n pelleti t s being releasede th f .I Pb-B s corrodeha i d sufficiently, they coul bottoe dRPVe th th fal o f t mlo , possibly forminga critical mass.
Suffic sayo t thermae t i e,th l expansio r contractiono Pb-Be larga th s f eha nio impace th n o t potentia r criticalityfo l time th t whic ea d an ,t occurs hi .
5.4. CONSEQUENCE CRITICALITF SO Y
Each of the three scenarios alone is unlikely to cause a critical assembly to be formed; however, t mutuallno the e yar y exclusive exampler fo , as , , contro corrosiod ro l causes ni watey db r ingresse .Th combinatio moror e two way nof whic in s h reactivit dumpethe corey of increasdSGI scan e hence make possibilite sth y less remote, particularl reactore th r yfo submarinf so e factory numbers 4211 ,90 and 601, which have the most SNF remaining and the higher enrichments.
If criticality is possible and occurs through slow corrosion of the control rods or water ingress, or a combination of the two, the approach to criticality will be extremely slow and the possibility of prompt criticalit kiny f explosioan do d yan r structurano l damag entirele b n eca y ruled out. Instead, the onset of criticality will cause a slow rise in fission rate, and an increase in SNF temperature. The conditions in the core are likely to be such that the heat generated is easily dissipated, particularly as the rate of reactivity addition is so slow, and the temperature is unlikely to rise significantly. This may cause a slight rise in corrosion rates and increased flow through the RPV. Since the cores will be water moderated, the rise in temperature will probably cause a reduction in reactivity, and there is a possibility that self regulation could occur corrosios A . n continues criticae th , l stat s likele i b o yt short lived compared wit totae hth l lifetim corrosionf eo , with further structural corrosio losd f nan so leadinF SN reductioa o gt reactivitn ni eventuad yan l subcriticality. Such behavio unlikels i r havo yt e a significant effect on total release rates.
A structural change resultin prompn gi t criticalit extremeln a s yi y remote possibility t cannobu , t be ruled out. If it occurred due to corrosion of the structure resulting in toppling and control rod displacement, the large release of energy that follows would cause the structure to further disintegrate and radionuclide release would be accelerated, possibly causing the total remaining inventory to be release environmente th o dt occurret i f attempteo I .t e ddu d remedial action which involved movement of the RPV, the consequences could be much greater.
A prompt criticality with core disassembl r distanfa e th t futurn yi e would involve very little radioactivity compared to the present radionuclide inventory in these cores. By the year 2700, nearly all of the current fission product inventory in the cores would have decayed. Also, the amount of fission products produce prompa n di t critical excursio relativels ni y small r exampleFo . amoune th , t of '"Cs generate" fissio10 a n n di criticalit y excursion (abou SL- e same th th ts 1 ea acciden e th n i t USA) would be 0.044 GBq [31].
5.5. FURTHER WORK
The above discussion has highlighted several areas which require further work to quantify the ris criticalitf ko y occurring. Areastudiee b o st d should include:
72 (1) detailed information on the structure and configuration of the cores, including the dimensions and supportinS S welde e siteth th f n si o g structures; (2) detailed modelling of the corrosion processes in the core, applying better estimates of corrosion rates to the core materials; ) (3 criticality calculation variour sfo s possible configuration SNFe th f , o sdroppes oncha e t ei th o dt bottom oftheRPV; (4) the behavior of the Pb-Bi eutectic thermal expansivity over the temperature range 125°C to 0°C; and ) (5 analysi change th f sneutroo n ei n spectru reactivitd man wates ya r enter core sth f submarin eo e factory number 601, when introduced as a consequence of bulk corrosion of the Pb-Bi, or if item ) abov(4 e suggest Pb-Be sth shruns ha iallowed kan d wate enteo t r r prio bulo rt k corrosion.
. REMEDIA6 L CONSIDERATIONS
6.1. REMEDIAL MEASURES
Before any remedial measures are taken, it is assumed that a site survey will be undertaken to establish the structural condition of the SNF containers and evidence of radionuclide leakage. In view of their relatively low levels of activity, this section does not address the issue of remedial measures for non-SNF radioactive material. Also attempo n , mads i t asseso et neede remediar sth fo l measures, as it is beyond the remit of the Group. However, an estimate is provided below of the physical condition of the SNF containers to assist in any assessment of necessary or feasible actions. It is up to other expert groups to assess whether the environmental impact, technologies, and cost considerations render such actions necessary and or feasible.
Potential remedial measures might include:
(1) reinforcemen existine th f o t g containment barriers aroun SNFe dr th o , (2) recovery of the dumped SNF for land storage.
6.1.1. Reinforcement of the existing containment barriers
To inhibit any monitored leakage from the SNF containers or enhance current containment arrangements, it may be possible to pump fillers inside existing barriers. Additional barriers may also be erected around the SNF containers, such as capping materials applied to potential leakage paths. Factor consideree b o st d include:
(1) The risk of a disturbance to the SNF containers that may damage existing barriers, perhaps inducing leakag weakenind ean g containment structures. (2) The risk of a disturbance causing a change in the orientation of the SNF in the RPVs sufficient to displace the CCRs and initiate criticality, to the extent that prompt criticality may occur. The resulting power excursion from prompt criticality could cause significan d immediatan t e radionuclide release to the environment and prejudice the safety of personnel at the scene. Even if slow criticality occurs, the increase in temperature may accelerate the corrosion rates and cause an early breakdow containmene th f no t barriers thif I . s occursealen i F ds SN witcontainers e hth , they may be ruptured by an accompanying rise in pressure, leading to premature breakdown of the barriers to radionuclide release. disturbanca rise f ko Th ) (3 e corrosione causinth RPVo e t th e g n i sdisplacemendu - F SN e th f o t weakened or damaged condition of the fuel supporting structures. In this case, the possibility criticaa f o l mass being forme bottoe RPVth n a t da f m ,o beneat CCRse hth , canno rulee b t d out. This could result in either slow or prompt criticality, similar to the situation discussed in the previous paragrap displacemenr hfo CCRse th f o t .
73 6.1.2. Recover dumpee th f yo d spent nuclear fuer lanfo ld storage
In the event that a decision is made to recover the dumped SNF to a land disposal site, the factor consideree b o t s rathee dar r more wide ranging:
( 1 ) The risk that, through a deterioration in their condition, the SNF containment barriers fail during recovery operations, releasing radionuclides eithescene th durint r era o g subsequent recoverd yan transportatio lane th d o nt disposa l site. (2) The risk that a disturbance to the fuel supporting structures or a change in the orientation of the SNF induces criticality, as described in items (2) and (3) of Section 6.1.1 (Reinforcement of existing containment barriers). (3) The nuclear safety hazard associated with the handling and transportation of the SNF to its final disposal site, including the radiation dose to personnel involved throughout the process. comparisoA ) potentiae (4 th f nlans o it dln i hazardisposaenvironmen e F th o SN dt e l th sit f eo t against retention at its sea disposal site.
6.2. STRUCTURAL INTEGRITY OF THE SPENT NUCLEAR FUEL CONTAINERS
Table XVIII provide estimatn sa structurae th f eo containersF l integritSN presene e th th f t yo ,a t time (1996) and in the year 2015, and implications for recovery purposes. This represents a theoretical assessment only should an , d therefor t suggeseno t remedial action without confirmatio actuae th f no l condition of the SNF containers through site survey.
7. CONCLUSIONS reactore th r theid Fo san r associate) (1 dumpeF Kare dSN th an di Seaisotopee th , s makine th p gu radionuclide inventor theid yan r activities were calculate e informatiobasie th th f so n do n o n quantity, enrichment, and burnup of the nuclear fuel. These inventories were grouped into three categories: fission products, activation products, and the actinides.
The total activity in 1994 was calculated to be 4.7 x 1015 Bq. Fission products made up 86% f thio s total, activation product cam% s2 12%ed fro an actinides,e mth majoe Th . r component of this total inventory comes from the container containing the SNF from the icebreaker N2 reactor.
It should be remembered that the inventory finally established by the Group for all these units was considerably lower than that originally published in the White Book. The White Book totae fissior th fo ls activatio d a nan q B " n10 quote productx dumpe8 e d8 th n si d objectse Th . Group established thae originath t l inventor e time long-live th f dumpinth o et l a y al r dfo g . Bq " radionuclide10 x 7 3 s swa
) (2 Using best estimate corrosioe th f I materialsso SG n e rateth f , so barrie r lifetimes, information structuree onth methodd an containmend F san sSN use e dumpingr th d fo f to computea , r model developeds wa . Calculations were releas e madth r radioactivitf efo eo y int Kare oth a Sea, using several scenarios.
Releas e radionuclideth el al rate r fo s s were calculated from object eacn i s n ha fjords A . example, usin bese gth t estimate scenari summind oan contributioe gth nfjorde forth l msal with PWRs, it was shown that a release rate peak of 3000 GBq-a" occurs around the year 2040 when the PWR containments are partially breached, and there is anothe1 r peak of 2100 GBq in the year 2305, when the icebreaker SNF container corrodes open. However, for a large part of the time, release rates lie between 20 GBq-a"1 and 2 GBq-a"1. Some very low levels of activation product releases migh expectede b t decada , int o nexe s or th e o t century fro corrosioe mth outee th f rno RPVse wallth f so .
74 TABLE XVIII. ESTIMATES OF THE CONDITION OF DISPOSED SPENT NUCLEAR FUEL CONTAINERS FOR REMEDIAL ACTION CONSIDERATIONS
Factory Condition numbed an r Disposal Disposal site Containment structures ——— Conclusions reactors date 1996 2015
901 1965 Abrosimov Fjor . Reactoa d r compartment (RC )- hul l a. 25% corroded - sound a. 40% -weakened RC bulkheads now (19%) believed 2PWR b. Mild steel bulkheads b. Unsound . Unsounb d ineffective as containment barrier, c. Reactor pressure vessel (RPV) c. Little corroded - sound c. 5% corroded - sound possible b hule t us l no o et y Ima t for recovery purposesV RP . remains intact. 285 1965 Abrosimo hul- vC l FjorR . a d a. 25% corroded - sound a. 40% -weakened RC bulkhead (1996w sno ) believed 1 PWR1 b. Mild steel bulkheads b. Unsound . Unsounb d ineffective as containment barrier, c. RPV c. Little corroded - sound corrode% c5 . dsoun- d possible b hul e t us lno o et y Ima t for recovery purposesV RP . remains intact. OK-150 1967 Tsivolka Fjord . Pontooa n a. Unsound2 . Unsounda 2 Pontoon may not be used for b. Outer container - steel box corrode% b30 . dsoun- d corrode% b45 . dsoun- d recovery purposes. Containers with welded dli remain intact. . Innec r containe steer- l clad c. 20% corroded - sound corrode% c35 . dweakene- d concrete box with welded lid 421 1972 Novaya Zemlya a. Barge a. Unsound2 a. Unsound2 usee b r t dfo no Barg y ema 1 PWR Depressio . Concretb en enclosure bulkheads degrade% b25 . d b. 45% degraded recovery purposes. Effectiveness c. RPV c. Little corroded - sound corrode% c5 . dsoun- d of concrete corrosion barrier likely severele b o t y degradedV RP . remains intact. 601 1981 Stepovoy Fjord a. Submarine hull a. 10% corroded - sound corrode% a25 . dsoun- d remaiV RP nRd Cintactan . 2LMR b. Bitumen inside reactor b. 15% degraded - sound b. 35% degraded - cracked Effectivenes f bitumeo s n barrier compartment considered severely degraded. c. RPVs and steam generators c. Sound c. Sound Otherwise no structural constraints on remedial actions.
1 A defueled PWR is also included in this RC. 1 Judged unsound on the basis that the pontoon and barge are known not to have been built for this specific purpose and are believed to have been old and at the end of serviceable life. Assumptions: a. The concrete and bitumen have effective lifetimes of 100 years, with a linear decrease in effectiveness over this period. b. Best corrosion rates (BCRs) Source uses th a , n di e Term Working Group models. c. A survey of the dump sites will be necessary to assess the actual condition of the discarded objects prior to any decisions on remedial actions. LMRse Th , with their very slow corrosion rate largd san e mas solidifief so d Pb-Bi coolant, were treated differentl separatelyd an y . Using this second models showwa t i n, tharelease th t e rate peak GBq-a"8 yea e t sa th remainin e rr mosy th 210b 1 fo f t o t 5bu g long lifetime, release rates stay around 7 x 10'3 GBq-a'1.
thir Fo s best estimate scenario calculatione ,th s were extended unti componentl al l corroded sha d away; for the PWR units, this was by the year 4570, for the submarine LMRs, the reactors will tak disintegrateo t e0 unti yea60 e th 6 l r3 .
With another postulated scenario, glacial action in the year 3000, it was calculated that 4500 GBq would be released to the Kara Sea simultaneously from all the PWRs and their SNF, and 210 froq LMRe m0th GB than si t year.
The release rate data, for all the radionuclides, from all the units in the Kara Sea was then submitted to the Environmental Modelling and Radiological Assessment Working Group of the IASAP for dispersion and dose calculation modelling. errorse Th , uncertainties) (3 conservatisd ,an modee th mn i l were discussed, nothing amongst others, the assumption made that all activity is released for immediate dispersal to the sea. This makes the estimated release rates perhaps overly pessimistic, as much of the corroded material will slump to the bottom of containment structures, or be buried in surrounding sediments and therefor t passeno e d intfjorde oth r circulatio fo s beyondnd intan Kare a oth aSe .
Corrosion rates were best estimates froe literaturmth t stilbu el leave some uncertainties, especially the rate assigned to the solidified Pb-Bi of the LMRs.
Any future studies should attemp actuae obtaio t us d l ncorrosioan n rate onsitd san e observations f barrieo r material effectiveness, from sample actuae th f so l objects themselves, providing such investigation does not breach the containment barriers.
(4) The risks of a criticality incident in the dumped fuelled reactors following corrosion of the components or remedial action was considered. The possibility of such an incident was shown vere b yo t lowrule e t coulb t bu ,d dno out . This shoul borne db minn ei f remediadi l actios ni contemplated.
(5) With regard to the availability of design information for the submarine and icebreaker SGIs, a great dea stils i l l unknown abou suppore tth t structure reactoe th r sfo r cores, thermal shieldsd an , RPVs. While sufficient informatio knows ni n froabouF icebreakee mdisposae SN th tth e th f o l r mako t s a e this concerna les f so lace thif th ,k o s informatio mosts i n significan criticalitr fo t y studies and remedial action evaluation of the submarine RPVs that contained SNF.
Sufficient informatio bees nha n provide submarine icebreaked th an r dfo coreR R eLM o rsPW t develop preliminary core models; however, informatio submarine th n coreno R bees seha PW n limiteo s requiro t s da e several assumptions with respec theio tt r compositio layoutd nan . Less is know abou U-Ae th t 1 allo ysubmarine fuelth f o s e PWRs tha knows ni n abou icebreakee th t r fuel and its configuration within the RPV. Thus, inventories and release rates associated with the submarine PWRs, and specifically those containing SNF, have the greatest relative uncertainty and would likely require further and more detailed evaluation should remedial actions be considered.
(6) Future work should also include regular onsite investigation of the integrity of the objects, looking for any leaks which have opened earlier than anticipated by the model. Firstly, the conditio welde th f nso which seal important leakage paths shoul investigatede db . These would PWRse includth f d o icebreakean ,e p eth to welde p containeF th ca n srSN R o weldd CC rli e th , the state of the concrete capping over unit 421. Hatches into RCs might provide access for small
76 remotely operated vehicles to look for radionuclides leaking from the SNF and for activation products corroding from the internal steels of the PWRs.
(7) Of course these disposal methods for damaged reactors would not be advocated now. From the available information however, this study has shown, that the containment methods employed by the dumping teams should be effective. Assuming that the Group has modelled the barrier strategy correctly and there is no disturbance to the objects, release rates will be very low, for most of the time these structures remain corroding away on the Kara Sea floor.
77 ABBREVIATIONS
BCR best corrosion rate R CC contro r compensatioo l n rods T EC emergency cooling tubes EPR emergency protection rods IASAP International Arctic Seas Assessment Project IPPE Institute of Physics and Power Engineering, Russian Federation R LM liquid metal reactor PWR pressurized water reactor RC reactor compartment RPV reactor pressure vessel RRCKI Russian Research Center "Kurchatov Institute" SG steam generator I SG steam generating installation SIAE State Institut r Applieefo d Ecology, Russian Federation SS stainless steel F SN spent nuclear fuel C TF technical fuel channel
78 DEFINITIONS control or compensation rods A movable part of the reactor in the form of a rod with high neutron absorption cross-section, which itself affects reactivity an uses d i r reacto dfo r control. coolant Either water or Pb-Bi in this report, that is used to remove heat fro e nucleamth r fuel. criticality e conditioTh f fissilo n e material, e.g..nuclear fuel, whea n fission chain reaction takes place. Subcritical means thae th t fission rate decreases. Supercritical means that the fission rate increases. emergency cooling tubes Tubes penetratin througV RP he th gwhic e hb watey ma r (LMR only) passe removo dt e post-shutdown decay heat fro reactoe mth r core. emergency protection rods CCRs reserve emergencr dfo y retur subcriticaa o nt l condition. prompt criticality uncontrollable Th e instantaneous increas fissioe th n ei n ratf eo a fissile material. reactivity parametee Th r givin deviatioe gth n from criticalit nucleaa f yo r chain reacting medium. reactor compartment space Th e withi nshia r submarinpo e tha dedicates i t e th o dt components of the SGI. reactor pressure vessel A cylindrical steel vessel containing the nuclear fuel, CCRs, EPRs ECTd an , s (LMR only) which penetrate througV RP e hth hea r liddo . slowing down power The effectiveness a substance has to slow down (moderate) fast neutrons. spent nuclear fuel Irradiated t fueintendeno l r furthereactorsn i dfo e us r . steam generating installation A collective term for the entire reactor plant contained within the RC. steam generator A vessel in which the heat of fission is transferred from the coolant to water and subsequently converted into steam for propulsion and electric power. technical fuel channel An assembl specifia f yo c numbe f fuero lreacto e pinth n si f ro the icebreaker Lenin.
oft BLÂ&K . f-. «^ftMgHagg
79 REFERENCES
[I] Facts and Problems Related to Radioactive Waste Disposal in Seas Adjacent to the Territory e Russiath f o n Federation , e PresidenOffic th e Russia f th o ef o t n Federation ("White Book-93"), Moscow (1993), 20-41. [2] SJOEBLOM, K.L., LINSLEY, G., "IAEA programmes relevant to the radioactive waste dumped in the Arctic Seas: Part 1. International Arctic Seas Assessment Project (IASAP)," in Environmental Radioactivit Arctie th Antarcti d n yi can . Strandc(P . HolmE , , Eds.), Scientific Committee of the International Conference on Environmental Radioactivity in the Arctic and Antarctic, 0steras, Norway (1993), 89-92. [3] SIVINTSEV, Y., Study of Nuclides Composition and Characteristics of Fuel in Dumped Submarine Reactor Atomid san c Icebreaker "Lenin": Par - Nuclea 2 t r Submarines, Russian Research Center "Kurchatov Institute," Repor . 31/7281No t , Moscow (1994). [4] SIVINTSEV, Y., Description of Shielding Barriers at the Nuclear Reactors Dumped in the Arctic, IAEA-IASAP-ReportNo Russia, .8 n Research Center "Kurchatov Institute," Repor. tNo 31/7358, Moscow (1995). [5] SIVINTSEV , PersonaY. , l communication, Russian Research Center "Kurchatov Institute," Moscow (1995). [6] SIVINTSEV, Y., Personal communication, Russian Research Center "Kurchatov Institute," Moscow (1995). [7] YEFIMOV, E.I, Radionuclides Composition, Characteristics of Shielding Barriers, and Analyse Weaf so k PointDumpee th f so d Reactor Submarinf so 601eN , State Scientific Center of the Russian Federation, Institute of Physics and Power Engineering, Obninsk (1994). [8] SIVINTSEV, Y., Study of Nuclides Composition and Characteristics of Fuel in Dumped Submarine Reactors and Atomic Icebreaker "Lenin": Part I - Atomic Icebreaker, Russian Research Center "Kurchatov Institute," Moscow (1993). [9] YEFIMOV, E.I, Personal communication, State Scientific Center of the Russian Federation, Institut f Physiceo Powed san r Engineering, Obninsk (1995). [10] RUBTSOV, P., RUZHANSKY, P., Radiation Characteristic Estimation of Irradiated Nuclear Fuel in Reactors of Nuclear Submarine and Icebreaker Lenin Dumped near Novaya Zemlya Archipelago, State Institut r Applieefo d Ecology, Ministr f Environmentayo l Protectiod nan Natural Resources of the Russian Federation, Moscow (1995). [II] SIVINTSEV, Y., Personal communication, Russian Research Center "Kurchatov Institute," Moscow (1995). [12] SIVINTSEV , PersonaY. , l communication, Russian Research Center "Kurchatov Institute," Moscow (1995). [13] KOLOBASHKIN, V.M., et al., Radiation Characteristics of Irradiated Nuclear Fuel, Moscow (1983 Russian)n )(i . [14] KOCHETKOV, A.L., TSIKUNOV, A.C. t e al.e ,Cod r CalculatioTh , fo e f Isotopo n e Composition of Fast Reactor Fuel, internal report No. 2654, State Scientific Center of the Russian Federation, Institute of Physics and Power Engineering, Obninsk (1980). [15] ABAGYAN, L.R., NIKOLAEV, M.N.t al.e , Group Cross-sectio Reactor nfo Shieldind ran g Calculation, Moscow (1981) (in Russian). [ 16] BELOUSOV, N.N., et al., "The use of the CETERA Code in reactivity effects analysis," Proc. VII All-Union Semina Probleme th n o r f Reactoso r Physics, Moscow (1991). [17] KEVROLEV, V.V., RECOL Continuous-Energy Monte-Carlo Cod Neutror efo n Transport, Preprint RRC KI, IAE-5621/5 (1987). [18] WILLIAMS, M.L., "Analysis of Thermal Reactor Benchmarks with Design Codes Based on ENDF/B-V Data," Nucí. Techn. 71 (1985), 386-401. [19] BRIESMEISTER, J.F., MCNP - A General Monte-Carlo Code for Neutron Transport, LA-7396-M-Rev.2 Alamos Lo , s National Laboratory Alamoss Lo , (1987)M N , . [20] GROFF, A.G., ORIGEN-2: A Versatile Computer Code for Calculating the Nuclide Composition and Characteristics of Nuclear Materials, Nucí. Techn. 62 (1983), 335-352.
81 [21] TIMMS, S., LYNN, N., MOUNT, M., SIVINTSEV, Y., WARDEN, J., Modelling the Release to the Environment in the Kara Sea from Radioactive Waste in the Dumped Reactor Compartment of the Icebreaker LENIN - Version II, Marine Pollution Bulletin, in press. [22] CARTER, J.H., Corrosio LENIf no N Reactor Compartment Materials, Departmen Nucleaf o t r Scienc Technologyd ean , Royal Naval College, Greenwich, London, Technical Memorandum RNC/NS/TM431 (1994). [23] HEISER, J.H. and SOO, P., Corrosion of Barrier Materials in Seawater Environments, Environmental and Waste Technology Center, Department of Advanced Technology, Brookhaven National Laboratory, Upton (1995)Y N , . [24] YEFIMOV, E.I. t al.e , , Radionuclides Release fro Fuee mth l Composition inta WateroSe . Assessmen f Nuclideo t s Radiotoxicit Dumpen yi d Reactor f Nucleaso r Submarine 601, State Scientific Centee Russiath f o r n Federation, Institut f Physico e Powed an s r Engineering, Obninsk (1995). [25] SHOESMITH, D.W. and SUNDER, S., An Electrochemistry Based Model for the Dissolution f UOo 2, AECL-10488, Atomic Energ f Canadyo a Limited, Whiteshell Laboratories, Pinewa, Manitoba, Canada (1991). [26] ALEXANDROV, V.P. and SIVINTSEV, Y., Composition and Properties of Compound Furfurol(F), Russian Research Center "Kurchatov Institute," Report No. 31/7522, Moscow (1995). [27] ALEXANDROV, V.P. and SIVINTSEV, Y., Application of the Compound "Furfurol" to the Isolatio f Spenno t Nuclear Fuel, Russian Research Center "Kurchatov Institute," Repor. No t 31/7705, Moscow (1996). [28] JOINT NORWEGIAN-RUSSIAN EXPERT GROU R INVESTIGATIOFO P F O N RADIOACTIVE CONTAMINATION IN THE NORTHERN SEAS, Dumping of Radioactive Wast d Radioactivan e e Contaminatio e Karth an i nSea , Norwegian Radiation Protection Authority, 0steras, Norway (1996). [29] BARTH, C.H., SHELDON, R.B., Corrosion Rate Structuraf so l Material Oceae th n nso Floor, KAPL-4701, Knolls Atomic Power Laboratory (1989). [30] SOO , Brookhave,P. n National Laboratory Memorandum: Revie Corrosiof wo n Rates Specified by the International Arctic Seas Assessment Program (IASAP), Brookhaven National Laboratory, Upton (1995)Y N , . [31] US ATOMIC ENERGY COMMISSION, DIVISION OF OPERATIONAL SAFETY, Operational Accidents and Radiation Exposure Experience within the United States Atomic Energy Commission, WASH-1192 AtomiS U , c Energy Commission, Washington (1975)C D , .
82 CONTRIBUTOR DRAFTINO ST REVIED GAN W
Dyer, R. US Environmental Protection Agency, United States of America
Gussgard, K. Norwegian Radiation Protection Authority, Norway
Lynn, N. Royal Naval College, United Kingdom
Mount, M. Lawrence Livermore National Laboratory, United State f Americso a
Sivintsev, Y. Russian Research Center "Kurchatov Institute", Russian Federation
Sjoeblom, K-L. International Atomic Energy Agency
Timms, S. Royal Naval College, United Kingdom
Warden, J. Royal Naval College, United Kingdom
Yefimov. E , Institut f Physiceo Powed san r Engineering, Russian Federation
Consultants Meetings
Vienna, Austria: 10-14 January 1994 Vienna, Austria: 14-18 November 1994 Greenwich, United Kingdom: 20-24 March 1995 Vienna, Austria: 18-22 September 1995 Vienna, Austria Februar9 5- : y 1996
83