ML20041B554
| ML20041B554 | |
| Person / Time | |
|---|---|
| Site: | West Valley Demonstration Project |
| Issue date: | 01/31/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20041B549 | List: |
| References | |
| NUDOCS 8202240247 | |
| Download: ML20041B554 (64) | |
Text
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Nuclear Regulatory Staff Safety Evaluation Report on the Dormant West Valley Reprocessing Facility l
January 1982 Docket No. 50-201 Nuclear Fuel Services, Inc.
and l
l New York State Energy Research and i
Development Authority Western New York Nuclear Service Center West Valley, New York i
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8202240247 080128 PDR ADOCK 05000201
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s TABLE OF CONTENTS Page 1.
INTRODUCTION.................
1-1 2.
FUEL RECEIPT AND STORAGE...................
2-1 2.1 Present Status.....................
2-1 2.2 FRS Pool Water.....................
2-1 2.3 FRS Ventilation Syste.i...
2-2 2.4 Implications of Cask Drop in FRS............
2-2 2.5 Earthquake.......................
2-3 2.5.1 Spent Fuel Pooi.................
2-3 2.5.2 Fuel Storage Rocks and Canisters.........
2-3 2.6 Tornado.
2-4 2.7 Conclusion.....
2-4 3.
REPROCESSING P LANT......................
3-1 3.1 Present Status.....................
3-1 3.2 Severe Earthquake....................
3-2 3.3 Severe Tornado.....................
3-2 3.4 Use of the Plant.
3-3
- 3. 5 Conclusion...
3-4 4.
HIGH-LEVEL LIQUID WASTE STORAGE...............
4-1 5.
LOW-LEVEL LIQUID WASTE TREATMENT...............
5-1 6.
FACILITY BURIAL GROUND....................
6-1 7.
LOW-LEVEL SOLID WASTE BURIAL GROUND.............
7-1 8.
SUPPORTING SERVICES AND PASSIVITY..............
8-1 9.
MANAGEMENT ORGANIZATION...........
9-1 10.
TECHNICAL SPECIFICATIONS...................
10-1 11.
SUMMARY
AND CONCLUSIONS...................
11-1 12.
REFERENCES..........................
12-1 APPENDICES APPENDIX A--NRC STAFF COMMENT AND CONSEQUENCE ANALYSIS OF THE EFFECT OF EARTHQUAKE ON SPENT FUEL STORAGE P0OL AT WEST VALLEY,.......
A-1 ii
4 s
r CONTENTS (Continued)
Page APPENDIX B--NRC STAFF COMMENT AND CONSEQUENCE ANALYSIS OF THE EARTHQUAKE RISK FROM THE DORMANT NUCLEAR FUEL SERVICES, INC. REPROCESSING PLANT AT WEST VALLEY, N.Y..........................
B-1 APPENDIX C--NRC COMMENT AND CONSEQUENCE ANALYSIS OF THE EFFECT OF TORNADO ON DORMANT REPROCESSING PLANT AT WEST VALLEY, N.Y..........................
C-1 APPENDIX D--NRC STAFF COMMENT AND CONSEQUENCE ANALYSIS OF THE EARTHQUAKE RISK FROM THE NEUTRALIZED LIQUID WASTE TANKS AT THE WESTERN NEW YORK NUCLEAR SERVICE CENTER, WEST VALLEY, N.Y.
D-1 APPENDIX.E--NRC STAFF COMMENT AND CONSEQUENCE ANALYSIS OF THE EARTHQUAKE RISK FROM THE ACID LIQUID WASTE TANKS AT THE WESTERN NEW YORK NUCLEAR SERVICE CENTER, WEST VALLEY, N.Y......................
E-1 APPENDIX F--NRC EVALUATION OF THE SAFETY ASSOCIATED WITH THE DEFECT IN PAN 80-2 AT WEST VALLEY, N.Y.
F-1 APPENDIX G--GLOSSARY.......................
G-1 iii
11 s
e s
LIST OF FIGURES EiULe 1.
Organizational Chart of Nuclear Fuel Services, Inc.,
West Valley, N.Y...
10-2 iv
-. e 1
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I LIST OF TABLES fag!!
1.
Radiation Level in the Process Cells.............
3-1 2.
Tornado Parameters for West Valley Site...........
3-3 3.
Radiation Levels in Support Areas..............
3-5 4.
Technical Specifications No Longer Needed..........
10-1 4
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s SAFETY EVALUATION REPORT 1.
INTRODUCTION On September 22, 1976, Nuclear Fuel Services, Inc. (NFS) announced its intention to withdraw from commercial nuclear fuel reprocessing operations at the Western New York State Nuclear Service Center (the facility) in West Valley, N.Y.
Reprocessing activities had ceased there in early 1972 when NFS had considered possible expansion of capacity and modernization of the plant through a modifica-tion program.
On October 1, 1980, the West Valley Demonstration Project Act (P.L.96-368) was signed into law.
This Act authorizes the Department of Energy (DOE) to carry out a high-level nuclear waste management demonstration project at the facility.
The purpose of this report is to provide the evaluation of the currently reduced activities at the facility by Nuclear Regulatory Commission (NRC) staff.
Since the issuance of the previous safety evaluation in August 1977,1 the staff has further analyzed areas where its information was lacking.
Much of this updating centers on the completion of the staff's series of analyses of the effects of severe natural phenomena on major structures at the facility.
The staff considered all the information currently available and the inactive status of the facility.
Where it was able to draw conclusions based on the available information, it did to.
When precise information was not available, conservative " bounding" assumptions were made to estimate the consequences.
The report considers the management of radioactive wastes at the site, the storage of spent fuel, the surveillance of the reprocessing plant, and present administrative requirements.
Each of these topics is related to the present facility license, CSF-1.
This report replaces the staff's Interim Safety Evaluation I Report for the West Valley Reprocessing Plant, dated August 1977.
A safety evaluation was prepared for the issuance of the license in August 1965.2 These evaluations were based on the information then available.
Since the issuance of the 1977 safety evaluation, the staff has developed and updated the information, particu-larly with respect to the potential impact of severe natural phenomena.
This new information is considered in this assessment.
1-1
9 2.
FUEL RECEIPT AND STORAGE 2.1 Present Status The NFS fuel-storage pool contains approximately 165 MT of fuel (approximately 70% of the capacity).
About 5 MT of fuel was shipped to Battelle Northwest Laboratory in 1978.
Fuel was last received in 1975 and the licensee has no plans for receiving any more fuel.
The pool was emptied and cleaned in late 1971 after the termination of fuel reprocessing operations.
Since completion of cleaning the pool, the radio-activity level in the pool has averaged about 6 x 10 4 pCi/ml,* 97% of which comes from cesium-134 and cesium-137.3 The radiation level over the pool (measured approximately 1 ft above the pool surface) is 3-4 mrem /hr* with the higher readings occurring around the edges of the pool, probably because of
" bathtub ring" effect.
The fuel-receiving station (FRS) cranes, fuel-handling equipment, and pool water-cleaning and cooling systems are operational.
The pool water-cooling system has seldom been used because of the low fission product heat content of the fuel in storage.
The pool water temperature has been 83*F.3*
- 2. 2 FRS Pool Water The pool water-cooling system for the FRS consists of two coolers, each having a capacity of 6 x 108 Btu /hr.4 The actual heat load is much less, about 1 million Btu /hr, because the fuel has cooled for a few years.5 Water is circulated through each cooler by an electrically driven 1,200 gpm pump.
The coolers may be operated independently.
Because of heat transfer to the ground, forced cooling may not be necessary.
The coolers have not been used during the winter.
Even if the coolers are needed to meet water-temperature require-ments (not a Technical Specification), their loss would be more a nuisance than a real safety problem; that is, the high vapor pressure might over-humidify the FRS air and slight contamination could occur, but water losses can be easily made up so there would be no loss of shielding problem.
At any rate, because of the heat capacity of the water, 3 weeks or more would be needed for it to reach boiling.
A control system shuts off the pump when the pool level drops 6 inches.
If this interlock should fail, the pool level can drop only 18 inches more because the pump suction line will be uncovered.
Plant water is used as the secondary coolant with blowdown going to the sludge pond.
The blowdown is monitored by a radiation detector that picks up the presence of radioactivity if pool water should leak to the secondary side of the cooler.
The cooling system is not protected against severe natural phenomena, but, as indicated above, cooling-system loss is not a safety severe problem.
The Sandia Laboratories have considered the possibility of complete water loss for the considerably more severe case of recently di: charged reactor fuel in a
- Updated information provided during site visit to West Valley, N.Y.,
on March 25, 1981.
2-1
't draft report prepared for the NRC staff.G Figure B.10 of that report shows the variations in peak cladding temperatures at different assembly decay heat powers.
The decay heat rate is directly related to the cooling time since discharge.
The staff has compared the average decay heat generation of the long-cooled fuel currently stored at West Valley (about 1 kW per assembly) with the curve of the figure.
In the storage situation in the pool, 1 kW of decay power corresponds to about 100*C peak clad temperature.
The storage configuration at West Valley is sufficiently similar to the model configuration in the Sandia report considering the very low peak cladding temperature predicted, that the staff has concluded that there would be no danger of cladding failure caused by the thermal environmental conditions resulting from loss of pool water.
2.3 FRS Ventilation System The ventilation system for the fuel-receiving and storage building has been upgraded and is considered to be in good condition.
It incorporates a recir-culation system passing the air through high efficiency particulate air filters.
On February 12, 1976, in connection with an application for possible expansion of pool capacity, the staff requested information from Nuclear Fuel Services, Inc. (NFS) concerning the FRS off gas system.
The staff's concern was the possibility of increased occupational exposure from an improperly installed off gas duct.
Because NFS has not supplied the requested information and is no longer pursuing its application and the staff's review has terminated, the staff has no basis for a conclusion regarding the system.
Because of the relative inactivity in the FRS, very little. load is being handled by the off gas system.
Primary sources of off gas would come from the venting of shipping casks (none were received) and from the changing of filter and ion-exchange media.
Over a period of time, the radiation background to personnel could be increased from this system.
The staff considers this a minor safety question of employee exposure.
It recommends that the NFS health physics staff monitor this area and corrective action be taken if there is a measurable i
change in radiation level.
l 2.4 Implications of Cask Drop in FRS Accidents such as a cask or assembly drop or release of contaminated cask coolant during cask receipt and unloading of assemblies in the FRS area could occur.
Considering the case of a cask drop, NFS, in its 1973 Safety Analysis Report, noted that the pool is situated entirely within the impermeable silty till formation.7 NFS concluded that a dropped cask could puncture the unload-ing basin.
However, the glacial soil surrounding the basin would prevent a massive leak.
Lawrence Livermore Laboratory (LLL) analyzed, at the request of the NRC staff, the effect of a cask drop and confirmed the NFS conclusion that the basin floor could be punctured from a full drop.* Estimates indicate a low probability of such an occurrence.
The NRC staff has considered, in addition, the very infrequent cask-handling operations.
The staff notes that, i
although it considers that basic safety protection is provided under conditions
(
of infrequent transfers, a substantial increase in the use of the unloading basin may necessitate a reevaluation of the protection provided in the event l
of a cask drop.
The staff has, therefore, concluded that there is adequate 2-2 l
protection provided during the interim dormant period, but a change in the status of fuel storage should require a reassessment of the safety of cask handling.
2.5 Earthquake Nuclear Fuel Services, Inc. originally recognized that seismic activity could be expected in the West Valley area and designed all structures according to the Uniform Building Code for Zone 3 (area of possible major damage).
How-ever, as discussed in Appendix A, the NRC staff later concluded that the UBC was not an adequate basis for evaluating the struct'sres and that new design should be based upan a method which considered tne dynamic effects which would result from peak horizontal acceleration of 0.2 gravity.
This method is similar to the method used to evaluate nuclear power facilities and is described in NRC Regulatory Guides 1.60 and 1.61.
2.5.1 Spent Fuel Pool Because, as discussed above, complete loss of water is not considered a health and safety hazard offsite, the principal concern would be the risk to employees.
To better determine the overall risk from a severe earthquake, the NRC staff requested that the Lawrence Livermore Laboratory (LLL) analyze the effect of an earthquake on the fuel-storage pool.
In its completed report, LLL indicated that the design strength of the pool floor would be exceeded.9 Its studies showed that structural distress could occur at accelerations of 0.16 gravity in the pool region in the upper east corner of the north well of the storage basin.
This distress could cause cracking that would result in a leak in this region.
LLL estimates that leakage, if it occurred, would occur above the soil and into the building enclosing the FRS pool.
As Dr. W. J. Hall of the Nathan M. Newmark Consulting Engineering Services stated, "I concur that it is likely that some degree of cracking should take place with an earthquake characterized by about 0.20 gravity peak horizontal ground acceleration, and very probably in the general area cited.
If the wall is cracked indeed some leakage may occur."S The NRC staff concluded in its analysis that some water seepage is possible following a severe earthquake (Appendix A).
- However, because of the surrounding silty till soil used as backfill material with excellent impermeability during construction and acting as a second contain-ment barrier, very little flow would be expected.
Simple water additions, already made to replace evaporation losses, would maintain present water levels in the fuel-storage pool.
2.5.2 Fuel Storage Rocks and Canisters Science Applications Inc. (SAI) was asked by the NRC staff to prepare a seismic-structural analysis of the existing racks in the FRS.
SAI completed its analysis to determine the threshold ground acceleration above which potential yielding, permanent deformation or collapse, and/or excessive deformation could occur to the storage rack structure and canisters.s Examination of the failure threshold levels show that the anchor bolts in the storage rack connection to the North wall will be the first item to reach yield stress.
This event occurs at about 0.18 g.
The results also show that the anchor bolt at the north wall connection might break at a threshold of about 0.21 g.
A probable scenario of events following the north wall bolt failure is that the anchor bolts connecting the rack columns to the pool floor will fail causing 2-3
the rack to deflect until it impacts the south wall.
This event could cause a few canisters to become dislodged from the rack and fall to the pool floor causing crushing of the canisters.
2.6 Tornado No offsite impact would be expected from the effect of a tornado on the spent fuel stored in the FRS, even if that unlikely event occurred.
Because of the protection afforded by the pool water, even an unlikely direct strike by a tornado would not result in a significant release of radioactivity.
The staff states in Section 3.34 that the frequency of a site strike of the largest credible tornado imaginable is less than once in 1 billion years.
The only tornado effect of interest for the FRS is the missiles it could generate.
In an almost identical review for the Midwest Fuel Recovery Plant (Docket 50-268),
the staff's analysis showed that there was no offsite risk from missile strikes of the pool.
First, the probability of a missile of potential harm striking the pool is very remote (and superimposed on the already low chance of a tornado strike).
Second, the fuel contains very little radioactive gas to be released even if all the fuel pins were somehow ruptured.
Third, the missile damage is selective; that is, only a canister or two of fuel would actually be a f fected.
Fourth, a critical excursion would be unlikely, even if a tornado struck, because the fuel would still be restrained because of close spacing.
Finally, the staff has calculated a maximum dose of only 1 mrem /hr at the surface of the pool,1 even if a critical excursion did occur.
2.7 Conclusion Based on the discussion above, the staff concludes that no undue risk to health and safety of the public or employees resulting from the continued storage of spent fuel at the West Valley site exists.
The staff considered the effects of natural phenomena and man-caused accidents on the stored fuel and pool in this evaluation.
2-4
4 3.
REPROCESSING PLANT 3.1 Present Status Following plant shutdown in 1972, Nuclear Fuel Services, Inc. started preparation for its proposed modifications.
Equipment was flushed out and decontamination began in the plant.
Extraction Cells II and III were decontaminated.
- However, this process was not completed in the plant as a whole.
Current radiation levels in the cells are updated and shown in Table 1.
Table 1 Radiation levels in the process cells mrem /hr mrem /hr Location (low)
(high)
Process Mechanical Cell 8x102 1.2x103 General Purpose Cell 5.0x105 1.8x108 Chemical Process Cell 1.4x104 2.4x104 Extraction Cell I 4.3x103 3.3x104 Extraction Cell II 10 90 Extraction Cell III
<1 12 Plutonium Product Cell
<0.5
<0.5 Uranium Product Cell 100 100 Source:
Updated information provided durint site visit to West Valley, N.Y. on Sept. 25, 1981.
I Most of the equipment remains intact in the cells, although some equipment such as the shear, has been partially disassembled.
Because the cells are i
still contaminated, they would require extensive decontamination before decommissioning.
While there are no process activities in the cells at present, the contained radioactivity could be released by the impact of natural phenomena such as earthquake and tornado.
Because of the terrain of the site, flooding is precluded.
Based on calculations from estimates given in a consultant's draft report on decommissioning,11 the Process Mechanical Cell contains 1,000-10,000 Ci of radioactivity, principally from strontium and cesium.
A substantial portion of this activity is probably fixed, that is, sufficiently held by walls and floor to resist becoming airborne during changes of ventilation floor patterns.
3-1
O
- 3.2 Severe Earthquake Earthquake occurrence at the site is infrequent, and even if one did occur, the building structure itself would remain standing after the earthquake and any likely winds would not remove any significant radioactivity from the cells of the structure.12 A separate, earlier analysis by the NRC staff had concluded that a free field acceleration of about 0.2 gravity at the surface would be a suitable design-basis value for new construction at the West Valley site.
This value is related to the largest earthquake that could be expected to occur at the site.
Both the Lawrence Livermore Laboratory (LLL)12 and the Los Alamos Scientific Laboratory (LASL)12 have analyzed the effect of an carthquake on the reprocessing l
plant.
The results of their detailed seismic evaluation studies of the NFS facility are described in Appendix B along with NRC staff comments based on an analysis of the LLL and LASL reports by an NRC consultant.12 In Appendix A, the analyses indicate that under the present dormant condition, the cells containing substantial quantities of radioactivity (see Table 1) would not release radioactivity at seismic accelerations up to 0.2 gravity.
For example, onset of failure of the General Purpose Cell is predicted to occur at about 0.1 gravity, but because it is embedded, the radioactivity would not be dispersed into the earth's atmosphere.
The Chemical Process Cell and the Process Mechanical Cell would be expected to withstand 0.15 gravity or more before the onset of failure.
Based on Figure 1 in Appendix B, an earthquake of this severity is not likely to occur once in a thousand years.
In any event, the release and dispersal of the radioactivity would still be unlikely, taking into account the lack of a dispersal mechanism within tho building, i.e.,
temperature or pressure gradients and the fact that " failure" as defined in the analysis, is the value at which stresses ' exceed elastic limits such that structural materials could yield or cracks could appear, but it is still unlikely that the confinement of radioactivity would be lost.
3.3 Severe Tornado Tornado occurrence is very infrequent at the site.
However, critical weaknesses l
in the building structure that might be affected by the impacts of a possible i
severe tornado strika were identified and the appropriate corrective actions l
were taken by NFS as noted below.
l The staff's initial estimate 1 of the recurrence interval for a tornado of the design-basis size proposed by the NRC staff in 1974 was 10 million years.
A recent estimate (Appendix C) places the recurrence interval for that size tornado in the West Valley area at about 1 billion years or more.
The recurrence interval for a tornado in that site region with maximum windspeeds of 200 mi an hour is estimated at 2 million years.
It is for this reason that the staff considers a tornado strike with windspeeds greater than 200 mi/hr unlikely.
However, the parametric analysis by Oak Ridge Natienal Laboratory (ORNL),13 as discussed in the following paragraph, considered windspeeds up to and including 300 mi/hr as shown in Table 2.
3-2
[
Table 2 Tornado parameters for West Valley site Maximum Radius Total Pressure tangential of tornado pressure drop drop rate velocity (mph)
(m)
(psi)
(psi /sec) 100 56 0.18 0.03 (4.9 in. H 0)
(0.9 in. H 0/sec) 2 2
200 108 0.63 0.12 (17.6 in. H 0)
(3.4 in. H 0/sec) 2 2
300 157
- 1. 3 0.26 (36.3 in. H 0)
(7.2 in. H 0/sec) 2 2
The principal concern for tornado impact in a heavy reinforced concrete structure of the kind at West Valley is the effect of a tornado on the ventilation system.
On our request the Oak Ridge National Laboratory has analyzed the effect of any credible tornado on the fuel reprocessing plant at West Valley.13 The analysts used computer code developed at the Los Alamos National Laboratory (LANL) expressly for this type of analysis.14 Conclusions from this analysis are that for certain of the pathways and tornado windspeeds examined, the tornado depressurization would not be sufficient to overcome the induced draft from exhaust blowers and acts to pull air into and through the cells.
In these cases, no air could be drawn from the cells to the atmosphere through unfiltered pathways, and no potential for release of radioactive materials exists.
Analysis of other pertinent cases indicated that tornado-induced forces could draw unfiltered air from the cells.
The pathways with the greatest potential for release of unfiltered air were identified as seven unsealed manipulator ports in the walls of the Process Mechanical Cell.
The analysis also indi-cated several other pathways with minor release potential.
The results of the analysis are discussed in Appendix C along with NRC staff comments.
Separate analysis by an NRC consultant 15 of the credible impacts of a severe tornado strike at the West Valley site indicated:
(1) the building wails would withstand the effect of a tornado and (2) there was a need for additional protection against tornado generated missiles along the north wall of the head end l
ventilation system enclosure.
These deficiencies were brought to the attention of the licensee and appropriate corrective actions were taken.
Therefore, the staff concludes that the health and safety of the public are adequately safe-guarded against the effects of any credible tornado that could strike the site.
I 3.4 Use of the Plant The plant is not being used for any processing at the present.
All of the solvent used for extraction has been removed and disposed.
Table 3 shows the radiation levels in areas accessible to personnel.3 8ecause work is limited to surveillance activities, occupational exposure has steadily decreased over 3-3
recent years.
If the plant were to be reused for any purpose, there would probably have to be additional decontamination of these areas.
4 3.5 Conclusion i
Based on the above discussion, the staff concludes that there is little risk to the health and safety of the public from the dormant reprocessing plant.
The plant was designed and constructed for active reprocessing operations and 1
is now in standby condition.
1 1
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Table 3 Radiation levels in support areas Location mrem /hr Below grade Crane room (GPC) 100-150 GPC operating aisle 0.5 Ventilation duct
- 1. 0 First floor Scrap removal room 20 Equipment decontamination room 30-200 Off gas blower room 0.3-1.0 x 103 Acid recovery pump room 0.5-1.0 x 103 South stairwell (failed line) 5-10 Liquid waste cell 0.5-1.0 x 103 Lower warm aisle 20-30 Ram equipment room 30 Manipulator repair room 2 x 103 Mechanical operating aisle 0.5 Uranium loadout area 0.5 Head-end ventilation duct 20-50 Second floor Upper warm aisle 50 Flush Penetration 5
Acid leak 5
Ventilation washroom 1-2 x 103 Ventilation duct 120 Crane room (PMC) 0.3-1.2 x 103 Equipment decontamination room 30-200 Third floor Ventilationduct(nearSpectrographic Laboratory) 150 Ventilation washroom effect 10 Chemical crane room 50-100 Solvent storage tank (empty) 100 Process' mechanical cell-door well 50-150 Fourth floor Control room 2.5 Stack area 10 l
Process mechanical cell-door well 100-200 l
Chemical process cell-door well 10-20 Chemical process cell roof 2.5 Fifth floor Pulse equipment aisle.
20 Cell ventilation duct (XCR) 50 Source:
Reference 3 3-5
i t
4.
HIGH-LEVEL RADI0 ACTIVE LIOUID WASTE STORAGE High-level radioactive liquid waste from plant operation from 1966 through 1971 is currently stored in two tanks at the site.
The larger volume, about 560,000 gal, is contained in a carbon steel tank that is enclosed within a reinforced concrete vault.
This liquid waste was neutralized with excess sodium hydroxide before being transferred to the storage tank.
A smaller volume, about 12,000 gal, is acidic high-level waste containing thorium and is stored in a stainless steel tank.
Each tank has a spare; the carbon steel spare is in a separate vault, the stainless spare is contained in the same vault as the tank in use.
In the Interim Safety Evaluation I report dated August 1977, the staff concluded that:
(1) No leaks have been detected from the tanks since they were placed in opera *.; ion about 15 years ago.
(2) An inspection of corrosion coupons in 1976 indicated that general corrosion rates are much less than those allowed by design.
(Corrosion rate is about 1/10 design allowable.)
(3) Temperatures in the tank are below boiling and are dropping (<185 F).
As a result, general corrosion is expected to decrease and stress corrosion cracking becomes more unlikely.
(4) The occurrence of stress corrosion cracking is difficult to predict, but all welds in the tank were stress-relieved to prevent this type of crack-ing.
The pH and nitrate concentrations are at levels which experience indicates are favorable for preventing stress corrosion cracking.
Furthermore, if a leak were to occur, it would have little effect on the public because:
(1) The tanks are encased in underground concrete vaults to contain any possible leakage.
(2) Experience at Department of Energy sites indicates that leaks which develop in tanks containing neutralized wastes tend to be self-healing.
Salt crystals form at the crack and seal off or slow down leakage providing time for corrective action to be taken.
However, in their design and operation 00E tanks use a dehumidifed annular space.
The effect of the humid air in the West Valley annular space is an indeterminant factor on the effect of the self-healing nature of the salt cake.
(3) Tank liquid waste contents can be pumped to adjacent spare tanks if a leak occurs.
Several improvements in the equipment and procedures for emergency waste transfer were made in 1979 by NFS.
(4) Soil surrounding the tank and vaults is notably impenetrable to water flow and seepage.
Thus, transport of radioactivity offsite and consequent 4-1
exposure of people off the 3,300-acre site from a leak through the multiple containment barriers is not considered credible.
In the staff's present evaluation, it considers the effects of natural phenomena and other occurrences in the high-level liquid waste storage.
In this regard, the staff has received a report from its consultants at the Oak Ridge National Laboratory (ORNL), dated February 7, 1977.17 ORNL reviewed information received from NFS concerning these effects and performed its own independent analyses.
The staff agrees with the ORNL conclusion that there would be virtually no impact offsite from tank failure caused by a catastrophic event such as a tornado or earthquake.
General flooding of the site is highly unlikely.
A constant head of water surrounding the vaults is maintained by a feedwater, drainage, and pumping system beneath the tanks.
For this reason, local flooding, if it does occur, will not be a problem.
Although the staff expects no offsite impact from waste-tank failure, regardless of cause, some difficulties would be presented by the recovery from failure, particularly structural failure of the tank-vault system.
In order to gain additional insights into the risks involved, NRC staff asked its consultants at Lawrence Livermore Laboratory to investigate the effect of an earthquake on the waste tank-vault systems.
Their reports, based on equivalent static analyses of both the neutralized liquid waste tanks 18 and the acid liquid waste tanks 19 were completed and are discussed in Appendices D and E, respectively, along with the NRC staff's own analyses.18'19 The " floating" incident discussed below was considered in these seismic analyses.
The large waste tanks and vaults " floated" during construction in April 1965.
The incident is described in a report by a Bechtel consultant, Louis S. Bernstein, who recommended corrective action, primarily grouting underneath the vault slab to remove bending stresses on the slab.20 The corrections were imple-mented in 1965 and accepted by NFS on January 24, 1966 through its consultants, Nussbaumer, Clarke and Velzy, Inc.
Since the operating license was issued on April 19, 1966, the Atomic Energy Commission (now NRC) regulatory staff accepted this correction.
In a subsequent safety investigation of the storage of the high-level radio-active waste in tank 8D-2 in 1978, the pan under the high-level radioactive waste tank 80-2 was detected to be defective.
The pan is provided beneath the tank to hold waste if a small leak should occur.
The pan has a capacity of approximately 30,000 gallons, while the tank has an operational maximum capacity of 600,000 gallons.
The pan was designed to act as an interim collection chamber or barrier for small leaks.
The safety implications of the 80-2 pan l
defect are described in Appendix F.
This evaluation does not consider the ultimate disposition of the waste.
Congress enacted P.L.96-368, the West Valley Demonstration Project Act on October 1, 1980, authorizing DOE to establish a project to demonstrate solidification techniques for high-level radioactive waste.
The DOE's rosponsibilities under the demonstration project include solidification of the liquid high-level radioactive waste; development of containers for disposal of the solidifed waste; transportation of the solidified waste to a Federal repository as soon as it is feasible; disposal of low-level and transuranic radioactive waste generated by the solidification program; and decontamination l
and decommissioning of all facilities associated with the storage and solidifica-tion of the high-level radioactive liquid waste.
4-2
9 5.
LOW-LEVEL RADI0 ACTIVE LIOUID WASTE TREATMENT The West Valley plant was designed to release liquid effluents containing small quantities of radioactivity during operation.
These releases were controlled by the use of settling lagoons, a weir, and further dilution in onsite streams.
Ali of these releases have been monitored.
On January 14, 1971, NFS submitted its safety analysis 21 of a low-level radio-l active waste treatment plant (LLWTP) proposed for reducing radioactive content of the liquid effluents.
The LLWTP plant began operation in May 1971.
The plant is conveniently located adjacent to the settling lagoons and separated from the reprocessing building.
Suspect water is fed to;the LLWTP plant from one lagoon and discharged to another.
In September 1972, NFS suspended releases of liquids in order to scrape and remove silt and settled radioactivity from the bottom of the discharge lagoon (No. 3).
Although liquids can still be released from the reprocessing plant for treatment, one principal use of the low-level radioactive waste treatment system, which has continued to the present, is the treatment of pump discharg7 from the low-level radioactive waste burial ground licensed by New York State.
Because of the low levels of radioactivity involved in this plant, the staff does not consider its potential loss because of an accident to be of signifi-Operation of the plant is considered by the staff to be necessary and cance.
beneficial for reducing the radioactive concentration in plant liquid effluents to levels as low as practicable.
1 5-1
4 6.
FACILITY BURIAL GROUND Based on its previous reviews, the staff concludes that no problems have occurred with the operation of the high-level' radioactive solid waste burial ground since operation of the West Valley plant began in 1966.
However, as noted in the later discussion in this section on the NRC licensed burial area, the NRC staff is presently funding some confirmatory research on the site f
geohydrology to improve the understanding of the containment capability of the burial grounds.
6.1 Present Status The high-level radioactive solid waste burial ground at the NFS West Valley site is operated under the provisions of the license for the Fuel Reprocessing -
Plant, CSF-1.
All of the buried high-level radioactive waste originated in the West Valley operations and consisted of process waste such as fuel hulls and contaminated equipment.
The site area for high-level radioactive burial is about 7.2 acres and is located about 1,400 feet southeast of the process building.
It is completely enclosed within the plant security fence.
6.2 NRC licensed burial area Leached hulls and other high-level radioactive solid waste that can be packaged in 30 gallon steel drums, are buried in holes aoout 32 inches by 78 inches and up to 50 feet deep.
These disposal holes are dug by using a crane and " clamshell shovel." Earth removed from the hole is inspected to make sure that burial will occur completely within the silty till.
After completion of the hole, the surface area is mounded, and drainage trenches are provided to minimize drainage if surface water into the disposal hole.
Solid-waste drums are normally brought to the high-level radioactive burial ground in groups of three.
j The three drums are placed in the hole one at a time by a crane and a lifting device that engage and release the drum automatically.
When the three l
drums are in place, a predetermined amount of backfill is added to the disposal hole to cover the drums.
This procedure is continued with added drums and backfill until the hole is filled nearly to the top.
Larger holes or trenches are prepared for equipment or material that cannot conveniently be fitted into the deep burial holes.
- his includes large process equipment and fuel-assembly components, tanks filled with sorbed solvent, and used ventilation filters.
The material buried in the high-level radioactive burial ground is secure from natural phenomena.
The local terrain prevents flooding at the burial site and reprocessing plant, and an earthquake having an acceleration of 0.2 gravity would not affect the material confined in the burial ground.22 Since the beginning of operation of the NFS plant, about 150,000 ft of material 3
containing about 500,000 Ci of radioactivity have been buried in the high-level radioactive burial ground.
This waste is buried in a thick layer of silty till.
The surface of the burial area is graded to minimize erosion and drain-age away from the disposal area.
Concrete monuments or cairns mark the location 6-1
of aurial sites, and the location and general type of burials are recorded on a chart of the site.
Duplicate copies of the burial records are maintained at separate locations to minimize the possibility of loss by fire.
The bulk of the radioactivity buried as high-level radioactive solid waste is induced radioactivity and residual fission product contamination associated with the cladding hulls.
This induced radioactivity and any residual fuel material can be released only by dissolution of the highly insoluble hulls to the groundwater.
Dissolution rates may be expected to be very low in view of the refractory nature of the zirconium alloy or stainless steel hulls and of the fact that hot nit.ric acid had been used to dissolve the fuel from the hulls.
A number of New Production Reactor (NPR) fuel assemblies (AEC-owned fuel furnished as part of the guaranteed base load) were received by NFS in poor condition.
NFS in consultation with AEC decided not to process this fuel because of problems related to its leaky condition.23 About one ton of this j
fuel was returned to the Hanford Works, and about one-half ton was encased in l
cement within 30 gal drums and buried in the high-level radioactive waste burial ground.
Release'of radioactivity from the high-level waste burial ground was evaluated-by NFS (p. X-3-20 of Ref. 4).
Mechanisms that were considered included (1) retention of the radioactivity in the burial hole, (2) retention of radio-activity in the soil near the disposal hole or trench, (3) dilution of radio-activity-bearing water by infiltrating rainwater filtering down through the soil column, and (4) radioactive decay during travel through the media.
The NFS study indicates that an escape rate of 10 7 fraction /yr from the trench is highly conservative.
Passage through 4 ft of silty till results in a fractional release of 6.4 x 10 8 to 6.4 x 10 8 and through 100 ft of surficial till, an additional 2.5 x 10.s to 5.0 x 10 7 According to NFS, the overall offsite releases including dilutions from onsite streams and by Cattaraugus Creek, result in stream concentrations many orders of magnitude below allowable ccncentrations (10 CFR Part 20).
The staff is considering the value of reanalyzing the health, safety, and environmental impact of the burial of these wastes over a long period.
The staff has asked the New York State Geological Survey to assist in this investigation.
The NRC's Office of Research is presently funding contracts by the USGS and NYSGS to study the geology, geomorphology, surface, and groundwater hydrology of the West Valley site.
The objective of the USGS work which will be continued through August 1983, is to assist the NYSGS in defining the extent of shallow gravel deposits and the locations and thickness of units underlying the gravel at the
" North Plateau" study area, define the groundwater flow regime through the shallow gravel and the underlying units of the " North Plateau" study area, define groundwater' flow in the till and lacustrine units at the NRC-licensed burial area, and define stream flow leaving both study areas.
The NYSGS work is to provide an understanding of the adequacy of the present containment and
(
the probable life span of the waste burial sites at West Valley as natural erosion processes work through time.
The potential for movement of radionuclides offsite through both the surface and groundwater systems will also be evaluated (see previous USGS contract work).
The work will specifically define the geological and hydrologic framework and mechanisms at the West Valley site and will evaluate the potential pathways for offsite radionuclide migration.
6-2
7.
LOW-LEVEL RADI0 ACTIVE SOLID WASTE BURIAL GROUND The low-level radioactive waste burial ground at the West Valley site has been operated by Nuclear Fuel Services, Inc, under license from the state of New York.
New York is an Agreement State; that is, it has assumed authority for regulating activities involving radioactive materials such as the low-level i
radioactive waste burial ground.
In recent years, a series of site investi-gations at West Valley centered on the hydrological and geological behavior of the soil and rock strata with respect to buried radioactive waste, The work has been sponsored by the Environmental Protection Agency, New York State, U.S. Geological Survey, and NRC.
The principal concern has been the potential movement of the buried waste.
As discussed above in Section 6, NRC is cur-i rently funding an ongoing research program by the Geological Survey of the State of New York to better understand the capability of the glacially formed soils to contain the waste matorial.
l l
[
7-1
8.
SUPPORTING SERVICE AND PASSIVITY The plant operations for which utility services are essential are the high-level radioactive waste storage, spent fuel storage, and process building safekeeping.
Electricity and natural gas are the two primary sources of energy supplied from offsite.
Diesel fuel oil (40,000 gal) is stored onsite as an alternate.
Electricity is used to drive ventilation fan motors and air compressors.
Natural gas or fuel oil supply energy to one of two boilers to make steam.
Water is supplied from two onsite lakes.
The pumps for this water supply are powered by a separate electric supply.
Complete electrical failure is relatively unlikely Decause there are two independent offsite supplies of electricity and an emergency generator, sup-plied by the onsite diesel fuel oil.
If electricity were completely lost, the main building ventilation fans and waste tank off gas blowers would shut off.
Because there are no process activities, the loss of building ventilation fans for the short time required to return to normal power or to repair the generator will nct cause any problems.
Without the off gas blowers, the waste tanks will lose their low vacuum, thereby increasing the potential for some contamination around access risers.
Pressure relief would still be available through the vent system, but the condenser fans would be shut off.
Natural convectior, alone will provide some cooling and water could be sprayed on the heat-transfer pipes as a temporizing measure.
Steam and electricity are used to compress air.
One 300-hp compressor is driven by an electric motor, a second 300-hp compressor by steam turbine.
Air is used in the waste tanks to c rculate the waste to prevent " hot spot" boiling i
l and smooth out the heat distribution.
Compressed ai.- is also used for liquid-level and specific gravity determination (bubblers) and to operate pneumatic instruments and controls, iacluding the boilers.
Steam is also used for heating and liquid transfers.
The sump eductors identified in Technical Specification 6.2 are steam jets.
Even though the vessels are no longer in use, NFS continues to test the eductors because this testing is required by the Technical Specification.
At any rate, if cell decontamination is planned, these eductors, and the motive steam, will be needed.
A steam eductor also is necessary for removing liquids from the high-level waste vault sumps.
Liquid collecting in the sump may be either intrusion water or water from radioactive waste leakage or transfer low-level liquid waste (LLLW) in the plant.
No direct radiological safety reason exists for heating the process building and, therefore, no steam is needed for this purpose.
The staff cannot deter-mine, from the type of information it receives, what effect loss of heat would have on the process building (that is, whether or not pipes can be drained or will freeze and burst).
The only solution to be concerned about in the pro-cess building is the depleted uranium stored in the uranium product cell (UPC) in two product tanks.
The UPC could be heated separately to keep this solution from freezing.
1 8-1 I
The staff concludes that electricity, steam, water, and air are needed at the site, but at a much reduced rate of use.
NFS has kept this service equipment repaired and operable.
The supporting services are more than sufficient to meet current needs.
N t
I 8-2 J
- w g-----
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9.
MANAGEMENT ORGANIZATION The Technical Specifications for West Valley require a staff which is competent to operate a reprocessing plant.
These staff positions are General Manager, Assistant General Manager, Production Manager, Health and Safety Director, and Tec'-ical Services Manager.
In addition, the Technical Specifications require a plant safety committee consisting of all of the above positions as a minimum.
In February 1981, the staff received information from NFS concerning its current organization.* The organizational chart for West Valley is shown in Figure 1.
All of the positions required by the Technical Specifications are filled by individuals with many years' experience at the site.
Because of the passive nature of the operation, a large staff is not required.
However, the staff must be competent and able to exercise the good judgment necessary if adverse circumstances should arise.
In addition to the onsite staff, a supervisor of the Health and Safety Department at the Erwin, Tennessee plant has visited the West Valley site several times to audit the health and safety program.22 Technical Specifi-cation 7.1.6 requires periodic audit of the operation by competent NFS or other personnel not directly responsible for the operations.
Based on the above discussion, the staff concludes that the NFS management is capable of safely conducting its present activities.
A Updated information provided during site visit to West Valley, N.Y.,
on March 25, 1981.
9-1
Figure 1.
Organizational chart of Nuclear Fuel Services, Inc., West Valley, N.Y.
General Manager (West Valley) l Secretary III i
l 1
Administrative Security
?
Service Manager Plant Manager Supervisor N
Administrative Aide l
I Accounting Data Control Guard Shift Supervisor Clerk Supervisor l
l Contract Buyer 1 Guards Health & Safety Technical Services Operations Quality Assuranct Manager Manager Manager Supervisor
~
IIEALTil & SAFETY TECilNICAL SERVICES ilealth & Sa fety
' echnical Services T
Manager Mana ger
\\
llealth & Safety Supervisor Plant Assistance Emission S ecialist & Stds, flechanical P
Supervisor llealth & Safety Supervisor ng neer p
Specialist -
Environmental Process Control llealth & Sa fety "E
Specialist -
Industrial Sa fety OPERATIONS Ileal th & Sa fety Operations Specialist Manager Radiological Assist. Operations Training Maintenance Maintenance Manager Coordina tor Engineer 3eneral Foreman Shi f t Maintenance F reman Supervisors
10.
TECHNICAL SPECIFICATIONS The staff has revised license CSF-1 including its Technical Specifications.
Because Nuclear Fuel Services, Inc. no longer plans to reprocess, many Technical Specifications no longer are necessary and are listed in Table 4.
Table 4 Technical Specifications no longer needed at West Valley, N.Y.
No.
Title 4.4 Dissolver charging 4.5 Feed solution concentration 4.6 Fissionable isotope concentration in solvent extraction 4.7 Extractant concentration 4.9 Plutonium ion exchange operation 4.0 Plutonium solution storage 4.11 Rework solution concentration 4.15 Evaporator steam pressure 6.1 Boron raschig rings 6.6 Dissolver dilution air
- 6. 7 Boric acid 6.10 Poisoned dissolver baskets 6.11 Solvent analysis 7.2 Category 10 fuels Based on its review of these Technical Specifications and consideration of their application to activities with much greater potential for release of radioactivity, NRC staff concludes that these Technical Specifications are adequate for the present dormant activities.
-l e
4 10-1
11.
SUMMARY
AND CONCLU% IONS In the previous sections, the staff has considered and evaluated the radiological safety of the present activities at the West Valley, N.Y. site.
These activities include the storage of spent fuel, the management of radioactive wastes, and the ongoing surveillance of the dormant reprocessing plant.
Nuclear Fuel Services, Inc. maintains a staff of about 50 people at the site to conduct these limited activities.
NRC staff has considered the impact of the failure of structures, systems, and components important for the health and safety of the public and employees.
Failures can be causea by human error or severe natural phenomena such as an earthquake or tornado.,
The staff concludes that because of the passive nature of the present activities in a plant designed for operation with a much greater risk potential, no undue risk to the health and safety of the public or to the employees exists.
11-1
~,
12.
REFERENCES l
1.
Nuclear Regulatory Commission, " Interim Safety Evaluation I, Docket No. 50-201, Nuclear Fuel Services, Inc. and New York Energy Research and Development Authority, Western New York Nuclear Services Center, West Valley, N.Y.," August 1977.
2.
Safety Evaluation by the Irradiated Fuels Branch, Division of Materials Licensing in the Matter of Nuclear Fuel Services, Inc. and New York State Atomic and Space Development Authority, Docket No. 50-201.
1965.
3.
Letter from R. E. Brooksbank, ORNL, to R. M. Bernero, Oak Ridge National Laboratory, dated November 9, 1976.
4.
Letter from J. R. Clark, Nuclear Fuel Services, Inc., to S. H. Smiley, NRC, dated May 17, 1972.
5.
Letter from J. R. Clark, Nuclear Fuel Services, Inc., to Richard E.
Cunningham, NRC.
Subject:
Letter from J. R. Clark, Nuclear Fuel Ser-vices, Inc., to Mr. Richard E. Cunningham, U.S. Nuclear Regulatory Commission, dated August 4,1975, dated January 30, 1976.
6.
Letter from David J. McCloskey, Sandia Laboratories, Albuquerque, NM, to James E. Slider, NRC, dated March 24, 1977.
7.
Nuclear Fuel Services, Inc., " Safety Analysis Report, NFS Reprocessing Plant, West Valley, N.Y.," Vol. III, p. X-3-3, Docket No. 50-201, 1973.
8.
Johnson, Neil E. and Walls, James E.
" Seismic Resistance Capacity Evaluation of Spent Fuel Storage Racks and Fuel at West Valley, New York," NUREG-CR-2236, Science Applications, Inc.
9.
Seismic-Structural Analyses on Fuel Receipt and Storage Pool, West Valley, New York Reprocessing Facility:
NRC letter forwarding reports to co-licensees, L. C. Rouse, NRC a.
to R. W. Deuster, NFS and J. Larocca, NYSERDA, dated May 14, 1979.
b.
Letter from Dr. W. J. Hall, Nathan M. Newmark Consultant Services, to Dr. A. T. Clark, NRC, dated March 7, 1979, providing review and comment on laboratory analysis listed below.
c.
NRC staff comment on the Effect of Earthquake on the Spent Fuel Storage Pool at West Valley.
d.
LLL report, " Structural Analyses of the Fuel Receiving Station Pool at the Nuclear Fuel Service Reprocessing Plant, West Valley, New York," (UCRL-52575) dated Mt/ 5,1978.
10.
Memorandum from C. R. Marotta, NRC, to J. R. Miller, NRC,
Subject:
Fuel Storage Pool Dose Rate Estimates as a Result of a Crit #cality Incident, dated November 20, 1975.
12-1
l1.
U.S. Nuclear Regulatory Commission, " Technology, Safety, and Costs of Decommissioning a Reference Nuclear Fuel Reprocessing Plant," USNRC Draft Report, NUREG-0278.
l 12.
Seismic-Structural Analyses on Reprocessing Building, West Valley, New York Reprocessing Facility:
NRC Staff Comment and Consequence Ana,ysis on the Earthquake Risk.
a.
b.
Letter from Drs. N. M. Newmark and W. J. Hall (senior structural consultants) to Dr. A. T. Clark (NRC) dated January 4,1978 providing j
review and comment on the laboratory analyses listed below.
Los Alamos Scientific Laboratory (LASL) report, " Summary of the c.
Seismic Analysis of the Nuclear Fuel Services Reprocessing Plant at West Valley, New York," (LA-7185-MS) dated March 1978.
d.
LASL report, " Seismic Investigation of the Nuclear Fuel Services, Inc., Reprocessing Plant at West Valley, New York," (LA-7087-MS) dated March 1978.
Lawrence Livermore Laboratory (LLL) report, " Seismic Analysis of e.
the Nuclear Fuel Service Reprocessing Plant at West Valley, N.Y.,"
(UCRL-52266) dated May 24, 1977.
13.
L. J. Holloway, Oakridge National Laboratory (ORNL), and R. W. Andre, Los Alamos Scientific Laboratory, LASL, " Potential Radiological Impact of Tornados on the Safety of Nuclear Fuel Services West Valley Fuel Reprocessing Plant," USNRC Report NUREG/CR-1530.
14.
W. S. Gregory and G. A. Bennett, " Ventilation Systems Analysis During Tornado Condition," LA-5894-PR, Los Alamos Scientific Laboratory, Los Alamos, N.M., March 1975.
15.
W. A. Coffman, " Assessment of the Western New York State Nuclear Services Center Head End Ventilation Building Structural Responses to Tornado Loading," March 1980.
16.
NRC Staff Comment and Consequence Analysis on the Effect of Tornado on Dormant Reprocessing Plant at West Valley, New York.
17.
Letter from R. E. Blanco, ORNL, to Robert M. Bernero, NRC,
Subject:
Effects of Natural Phenomena Events on NFS Waste Tanks 8D-2 and 80-4, February 7, 1977.
18.
Seismic-Structural Analyses on Neutralized High-Level Liquid Waste Storage Tanks, West Valley, New York:
a.
NRC letter forwarding reports to co-licensees, L. C. Rouse, NRC to R. W. Deuster, NFS and J. Larocca, NYSERDA, dated June 20, 1979.
b.
Letter from Dr. W. J. Hall to Dr. A. T. Clark dated March 2, 1979 providing review and comment on LLL report listed below.
12-2
NRC Staff Comment and Censequence Analysis on the Earthquake Risk c.
from the Neutralized Liquid Waste Tanks.
d.
LLL report, " Seismic Analysis of High Level Neutralized Liquid Waste Tanks at the Western New York State Nuclear Service Center, West Valley, New York," (UCRL-52485) dated May 1978.
23.
Nuclear Fuel Services, Inc., " Quarterly Report for period April 1,1976 through June 30, 1979."
6 6
a 12-}
APPENDIX A NRC STAFF COMMENT AND CONSEQUENCE ANALYSIS OF THE EFFECT OF EARTHQUAKE ON SPENT FUEL STORAGE POOL AT WEST VALLEY DOCKET NO. 50-201 In our Interim Safety Evaluation I issued in August 1977, the s'taff discussed fuel receipt and storage at the Western Nuclear Service Center (WNYSNSC) at West Valley, New York.
We indicated that structural analysis of the effect of earthquake on the fuel storage pool was being undertaken by the Lawrence Livermore Laboratory (LLL).
Although our evaluation indicated there would be no danger to the health and safety of the public, the further analysis of earthquake was intended to better determine the risk to employees.
The report prepared by LLL, Structural Analysis of the Fuel Receiving Station Pool at the Nuclear Fuel Services Reorocessing Plant, West Valley, New York, UCRL-52575, presents the results of their seismic analysis.
Livermore con-cludes that structural distress could occur at accelerations above sixteen percent of gravity in the pool region in the upper east corner of the north wall of the storage basin.
We have asked Dr. William J. Hall of Nathan M.
Newmark Consulting Engineering Services to review the results of the analysis.
Dr. Hall's letter, dated March 7, 1979, discussing his review, is attached.
He concludes that a strong earthquake could cause some cracking in the general area cited by Livermore.
His view is that "it is likely that some degree of cracking should take place with an earthquake characterized by about 0.2 g peak horizontal ground acceleration."
As stated previously, our interest in this instance is the health and safety to plant employees.
In our previous evaluations we concluded that there was no risk to the public from operation of the Fuel Receipt and Storage Area, even in the unlikely event of complete loss of water.
The further analysis of seismic loads show that some water seepage is credible following a severe earthquake.
However, because of the silty till soil used as backfill material during construction, very little flow would be expected.
Simple water additions, as are already made to replace evaporation losses, would maintain present water levels.
We have attached the earthquake probability figure previously shown in our interim safety evaluation.
This figure indicates a probability in the range of 2 x 10 4 per year (once in 5000 years) for the occurrence of an earthquake which could cause the possibility of the cracking described.
The basic integrity of the basin would be maintained.
Effects of Earthquake on Crane 8ecause of its very low probability of occurrence, we have not analyzed the effect of seismic loads on the gantry crane used to lift shipping casks during fuel transfer operations.
Because of the limited activity at the site the crane is normally resting in a position away from the fuel unloading basin.
Occasionally, the crane may be used to lift a shipping cask over the unloading basin.
The crane cannot travel over the fuel storage basin.
If it were A-1
p"ossible for the crane to fall during an earthquake, and it was positioned over the cask unloading basin, it does appear to be possible to strike the wall separating both basins.
Again, water leakage from the pool would be slow.
Because of the long cooling time for the fuel, the release of gaseous radioactivity would essentially be limited to krypton-85 contained in the gap between the cladding and the fuel pellets.
This would have no discernable ofisite impact.
The danger of the falling crane would be by far the worst hazard to nearby workers.
It is not known whether or not a criticality could occur as a result of this accident.
The situetion described is not readily amenable to analysis.
We refer again to our previous analysis which estimated maximum doses on the order of one millirem per hour at the water surface if a criticality did occur.
Cask Drop In their 1973 Safety Analysis Report for the proposed modifications of the reprocessing plant, Nuclear Fuel Services considered the dropping of a cask in the unloading basin.
Without analysis it was concluoed that a dropped cask would puncture the unloading basin.
As pointed out previously, the basin is surrounded by glacial soil which would prevent a massive leak.
At our request, LLL analyzed the effect of a cask drop and confirmed the NFS conclusion that the basin floor cosld be punctured from a full drop.
Estimates of the likelihood of such an occurrence indicate a low probability.
We have considered, in addition, the very infrequent use of the crane.1 Restriction on Acolication of Evaluation We note that in our previous safety evaluation of August 1977 and in our above discussion we have taken into consideration the dormant activity at the site.
We caution that other conclusions might be reached if these conditions were changed.
Our analyses cover only the present reduced activity and should not be extrapolated to other modes of operation.
Conclusion Based on the above discussion, we continue to conclude that there is no undue risk to the health and safety of the public or employees from the reduced spent fuel storage activities at the West Valley site.
ISafety Analysis Report, Nuclear Fuel Services, Inc., Reprocessing Plant, West Valley, New York, 1973, p. X-3-3.
A-2
l
.010
~
i
. Figure 1 Estimate of Probability of Earthquake at West Valley, N.Y.
\\
.001 c
w
~
3 Yu o
C.
.g
=
35
.0001
.00001
'e
.05g.10g.1Sg.20g.25g.30g Peak Horizontal Acceleration, fraction of gravity A-3
l l
[
APPENDIX B NRC STAFF COMMENT AND CONSEQUENCE ANALYSIS OF THE l
EARTHQUAKE RISK FROM THE DORMANT NUCLEAR FUEL SERVICES i
REPROCESSING PLANT AT WEST VALLEY, N.Y.
l DOCKET NO. 50-201 i
The Nuclear Regulatory Commission staff has been investigating the risk of earthquake damage to the structure previously used for reprocessing spent nuclear fuel at West Valley, N.Y.
Three separate structural analyses were undertaken.
On October 12, 1976, Nuclear Fuel Services, Inc. submitted a report prepared for them by the Chemical Plants Division of Dravo Corporation.
The report,
" Seismic Competence of the Existing Process Building at the West Valley Repro-cessing Plant" (Report No. 0476.015) assessed the probable competence of the existing reprocessing plant under natural phenomena loading.
Independent analyses for this assessment were performed by the Los Alamos Scientific Laboratory (LASL) and the Lawrence Livermore Laboratory (LLL) at the request of the NRC staff.
Each laboratory has prepared separate reports on their analyses.
In addition, LASL has prepared a summary report on the three independent analyses.
We also have asked Drs. Nathan M. Newmark and William J. Hall of the Nathan M.
Newmark Consulting Engineering Services to review and comment on these analyses.
Drs. Newmark and Hall have extensive experience in seismic analytical methods and in the actual effects of earthquake on structures.
Their conclusions and recommendations are presented in their letter report to the NRC staff dated January 4, 1978.
Based on staf f review of information submitted by Nuclear Fuel Services, Inc.
and its own, independent sources of information, the staff concluded that a site seismic design acceleration of 0.2 g was appropriate for new additions and modifications.
No such criteria have been developed for the existing structures.
The structure was built to Uniform Building Code Zone III specifications during the period 1964-1965.
Zone III indicates an active earthquake zone.
In our investigations, we have assessed the capability of the existing structures against the design value of 0.2 g.
As discussed in our Interim Safety Evalu-ation,1 the staff has also developed information on seismic recurrence intervals for the site.
These estimates are equivalent to predicting the probability of the occurrence of a single earthquake as a function of its magnitude.
This relationship is shown in Figure 1.
The curve of Figure 1 represents a fit to Central Stable Region earthquake histories from modified Mercalli intensities IV, VII, and VIII.
Seismic analysis of the structure is complicated by the nonsymmetric arrangement of its component cells.
The cells are large rooms formed by reinforced concrete INuclear Regulatory Commission Staff, Interim Safe y Evaluation 1, August 1977, Docket No. 50-201, Nuclear Fuel Services, Inc. and New York State Energy Research and Development Authority, Western New York Nuclear Service Center, West Valley, N.Y., Appendix I.
B-1
.~
Pigure 1
.010 l
l
. ' Figure 1 Estimate of Probability of
. Earthquake at West Valley, N.Y.
l
.001 a
l uc o>
\\
g c.
~
c 5
O 4001
'.1.6x10' [M l
l l
.00001
't
.059 10g.15g.20g.25g.30g Peak Horizontal Acceleration, fraction of gravity B-2
w' alls, floors, and ceilings.
They contain the process equipment and, during processing, are a principJe containment barrier for both radiation and radioac-tivity.
The cells are generally perpendicular to each other.
One cell, the General Purpose Cell, is completely imbedded in the surrounding clay soil.
The series of four extraction cells are tall and somewhat narrow.
Ground support for the entire structure comes from both the soil and over five-hundred steel piles.
The analyses considered the structure as an interacting system between the ground support and each cell.
There were differences in the methods by each analyst for modeling this system.
In addition tc the complicated configuration of the structure, each analyst-was faced with the difficulty of defining the condition of failure.
Various modes of failure or other phenomena were analyzed.
For example, the possi-bility of the entire structure overturning or sliding was studied.
It is important to note that failure has been defined in a structural analysis sense, i.e., by comparison of maximum stresses with limits in building codes.
Such " failure" might not even be observable in the actual structure.
As Drs. Newmark and Hall state, "At the excitation level noted we do not see any likelihood of gross collapse of the Process Building as a whole.
We would expect localized cracking at or near penetrations, at reentrant corners, and near discontinuities generally."
Throughout this evaluation, the attempt has been made to strip the analysis of much of the conservatism inherent in the method.
This is not done because the staff does not want to be conservative in its approach to protection of the public; but rather, the removal of conservatism is necessary in order to realistically determine the risk to the public.
The staff continues to require for any new construction the conservatism inherent in its guides and regulations and analyses.
The staff concludes from these analyses that the basic cell structure will be able to withstand the largest credible earthquake.
- However, external walls and supporting systems could fail and cracks appear in walls.
In its further analysis of the consequences the staff assumes a considerable loss of confinement through destruction of the ventilation systems and ejection of the manipulators through thes~e wall penetrations.
A summary of the results of the analyses is shown in Table 1 taken from the LASL summary report (LA-7185-MS).
Although the results of the Dravo study are presented in Table 1 along with the results of the LASL and LLL studies, they are not directly comparable.
The Dravo report indicates that by an earlier method of analysis the structure could successfully withstand ground accelerations of 0.12 g.
The Dravo results are presented for referral as a matter of convenience.
Since the Dravo report was received after NFS elected to withdraw from reprocessing, the staff was unable to pursue questions which could have been helpful in understanding the details of the analysis.
Therefore, the staff did not rely upon the Dravo report to realistically assess the seismic capability of the structure.
Quantity of Radioactivity at Risk The quantity of radioactivity at risk is also not precisely known.
Because of the radiation levels in the cells, remote measuring methods must be employed B-3
Table a Summary of results Earthquake Levels at Impending Structural Distress and Locations A
B Substructure of Chemical plants Lawrence Livermore Los Alamos Scientific Structural member Division of Dravo Laboratory Location Acceleration Location Acceleration Location Acceleration Location I
(g's)
(g's)
(g's)
Shear Wall (Reinforced Concrete) 0.04*
CPC**
0.09 GPC
>0.20 N. R.
0.15 GPC oo 0.17 L1WC L
Shear Wall (Concrete Block)
N. R.
0.03 TOP N.C.
0.07 AOC Pile distress (combined loading) 0.08 N.R.
0.11 0.14 N.R.
Pile distress (tension) 0.05 EPC 0.17 0.16 EPC MCR section (flexture or shear)
N.C.
>0.20 0.14 MCR Intersection of GPC & PMC N.R.
>0.20
>0.20 Joint pounding N.C.
N.C.
>0.20 EPC, CPC N.R. - Not Reported EPC - Extraction & Product Purification Cells N.C. - Not Considered CPC - Chemical Process Cell GPC - General Purpose Cell PMC - Process Mechanical Cell MCR - Mechanical Crane Room ADC - Acid and Of f-Gas Recovery Cell LWC - Liquid Waste Cell TOP - Concrete Block Structures on top of CPC
- Using different assumptions, a 0.12g competence level was once reported.
^^ Based on CPD's Interpretation of steel splice integrity in the CPC south wall.
- Perimeter of building, the level at which 10% of piles would be in tension.
[
to estimate radioactive contamination.
NFS has provided some measurements of radiation levels in each process cell.
From this information, rough estimates can be made of the corresponding contamination.
Some of this contamination adheres strongly to equipment or cell walls or floors.
In our analysis, we estimate the fraction of " loose" contamination of each individual cell.
In addition, the aerodynamic properties for transport, likely pathways for transport, and the probable ventilation and atmospheric conditions supplying the motive force for transport are estimated.
Estimate of Radioactivity in Cells In our August 1977 interim safety evaluation report on the present dormant operations at West Valley, we estimated that the Process Mechanical Cell contains between 1,000 and 10,000 curies of radioactivity. The estimate was deduced from information supplied by NFS on radiation levels in process cells i
.and from tables showing the relationship between cell contamination and radi-ation level in a Battelle Northwest Laboratory report on decommissioning.2 Since the ascumed cell mixture of fission products contained both beta and gamma emitters, a calculated adjustment was made for the NFS plant.
The adjustment consisted of assuming that gamma emitters were the only contri-butors to gross cell radi:1 tion measurements and accounting for the radioactive decay of gamma emitters such as the cerium praseodymium-144 and ruthenium-rhodium-106 pairs.
NFS has also supplied information on airborne concentrations of radioactivity in each of the process cells.
The densest alpha concentrations are about i
2 x 10 11 curies per cubic meter.
Since atmospheric dilution factors, even under the most restrictive conditions of stable inversion, are larger than 103 at site boundary distances, the offsite concentrations would be less than the maximum permissible concentrations for normal operation for any isotope, even plutonium.
The accidental release of unfiltered cell air under such extreme l
conditions would not be significant.
Effect of Wind The more significant contribution, therefore would be the effect of winds on the contaminated dust and debris in the cells following the run of the postu-lated earthquake.
Meteorological data representing hourly averages taken at the site in 1974-1975 indicate an average wind speed of about 11" miles per hour near ground level.
Within these averages there may be winds of greater speed.
The data also indicates that average winds in any direction under any stability conditions and greater than 24 miles per hour could occur less than 0.2 percent of the time.
Although higher wind speeds of shorter duration than one hour could occur and not be indicated in the joint frequency data, such " spikes" usually would not be more than twenty percent greater than the one hour average.
Therefore, we believe that a 15-mile per hour wind with 20 percent gusting (above and below l
l thI average) fairly represents the average conditions somewhat conservatively.
zTechnology, Safety, and Costs of Decommissioning a Reference Nuclear Fuel Reprocess, NUREG-0278 (final draft version).
8-5
~.
We also have considered 30 mile per hour winds with 20 percent gusting, although the occurrence of such winds during any year would be less than 0.2 percent.
An analyis of the effect of these winds is somewhat speculative since the physical condition of the plant and the particle size distribution of the contaminated dust are not known.
Because the plant is not operating there are no hot liquids or liquids under pressure to provide a further dispersing mechanism.
In order to analy_e the consequence of a large earthquake on the separations plant, the staff has used computer analysis assistance developed at the Los Alamos Scientific Laboratory (LASL).
One computer program, TVENT, has the capability to follow the dynamics of air pressure and flows in cells through-out the plant following the earthquake.
A second computer program, SOLA-ICE, has the capability to show detailed air flow patterns within individual cells using, as input data, the output information for the cell from TVENT.
Figure 2 shows a cutaway' view of the three process cells of interest in this analysis and their relation to each other.
These cells are of interest because they contain the substantial quantities of " loose" radioactivity.
During normal ventilation, the flows are as shown by the arrows in the figure.
Figure 3 shows the simplified model of the system used for the TVENT calculations.
In this analysis we made a number of simplifying assumptions, viz.:
(a) concrete block shear walls were assumed to be nonexistent, (b) steel frame roofs will not provide confinement, (c) complete loss of Mechanical Crane Room, (d) complete loss of Head-End Ventilation Syst'em including the stack, (e) complete loss of the Control Room, (f) complete loss of emergency generator and other utility equipment, (g) all manipulator sleeve parts are opened, and (h) all ventilation ducts to cells are opened.
The effects of these assumptions are that operating gallery areas are directly exposed to outside ambient air.
This is a maximum condition for release as we have assumed the complete loss of confinement, except for the basic confinement of the cells.
Although the emphasis in this investigation is on the health and safety of the public, the information developed is also useful for understanding the risks to personnel onsite.
Particulate Characteristics The staff has no specific information on the characteristics of the radioactive particulate in the cells.
Considering this type of activity in the plant the Process Mechanical and General Purpose Cells probably contain a significant quantity of fuel particles.
The number of fuel particles below 40 microns is B-6
Figure 2 Cutaway yieW of Process Cells a
k.]
v
~) %
3
's
/-
's
\\
l s
I
's l
l
's p
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$1lpCHEMICAL f
PROCESS N
s g
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4+.
c v
J {q)
I l
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~
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s N
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e l
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's
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'./
MECHANICAL
's 7
t CELL l
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Q "',
y
's s
,'d
.N GENERAL N
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..g CELL I
I I
l l
e a e.
I B-7
F.igure 3 Simplified.T. VENT Mo' del of fload Erjd Ventilation System e
1o o
87,000 Ft3 (1)
(2)
Cliemical Process Call 3
7
()
0 0780 Ft3 i
4o A
o,,,,,i
(),.
Process.
h 6
O (5) 15,600 Ft3 i
(4)
Process i
Machanical Ct.ll 1 = Nodo Number I.D.
l (1) = Branch Number I.D.
h-flosistanco to Flow (Duct, Dampor, Filtar)
B-8
.. r
- : --- - =. u :
i i
i probably low as determined from shear studies at Oak Ridge National Laboratory.
On the other hand, dusts which find their way into the plant, and to which radioactivity has adhered, will consist of smaller sizes and will be in the cells in significant quantities.
Visual observation of the cells indicate a dirty condition with substantial dusts and dirt.
Consequence of Normal Winds The air flows in the cells induced by 15-mile per hour winds outside the structure is low.
Table 2 shows a comparison of normal ficws between cells with that created by the average wind.
The flows shown in Table 2 would not be able to entrain particulate material and thereby remove it from the cells.
Studies of the movement of settled particulates by air indicate that for all particulate sizes and densities a threshold friction speed must be exceeded before entrainment can occur.
Table 2 Comparison of Cell Air Flow 3
Flow, ft / min Branch Normal At 15 mph Wind 1
8,600 6,000 2
8,600 6,000 3
0
- 500 4
1,700
-5,000 5
1,700
-5,500 6
10,300 0
Note:
A minus sign indicates flow in a direction opposite to normal flow.
The SOLA-ICE ccde calculates air speeds of about 15 feet per second near the floor of the General Purpose Cell during normal ventilation conditions.
At this flow rate (corresponding to about 10,300 cubic feet per minute exit flow from the cell) no particulate material of densities greater than about 2.5 grams per cublic centimeter would be suspended, entrained, and carried from the cell.
This can be deduced from curves such as those shown in Iverson, et al.,3 on page 383.
The calculated threshold friction speed of about 20 centimeters per second corresponds to an air speed near the floor of 15 feet per second.
The typical wind speeds of 15 miles per hour can induce flows of only about two-thirds this speed or about 10 feet per second at tne General Purpose Cell floor.
Such flows are not sufficient to entrain material for transport external to the building.
"J. D. Iversen, J. B. Pollack, R. Greeley, and B. R. White, " Saltation Threshold on Mars:
The Effect of Interparticle Force, Surface Roughness, and Low Atmospheric Density," ICARUS 29, 381-393 (1976).
B-9
Although winds as great as thirty miles per hour are expected less than 0.2 percent of the time at the site, we have examined their effect on the plant under the same circumstances as for the normal winds.
Under this con-dition air flows comparable to normal operation would occur.
The direction of air flow between the Process Mechanical Cell and the General Purpose Cell is reversed.
Very little entrainment would be expected under these conditions.
Re-Use of the Building The staff's conclusions with respect to the safety of the separation building are based on an analysis of the present dormant condition of the plant.
The amount of radioactivity at risk is low and undisturbed.
Our analysis is valid only under these circumstances.
As stated in the LASL summary report, possible modification to the existing structure was briefly considered.
The conclusion that modifications would be costly was not based on an actual analysis of methods and techniques which could be used but rather on the analysis of the seismic effects themselves.
Certainly before the structure were to be used for any new purposes, an analysis of that specific use would be necessary.
Conclusions The staff and its consultants have examined the effect of a major earthquake on the reprocessing plant structure at West Valley, N.Y. and any subsequent consequences.
The structure itself would remain standing after the earthquake and any likely winds would be unable to remove any significant radioactivity from the cells of the structure.
Further analysis is required if the plant is to be used for any processing activity.
B-10
APPENDIX C NRC STAFF COMMENT AND CONSEQUENCE ANALYSIS OF THE EFFECT OF TORNADOS ON DORMANT REPROCESSING PLANT AT WEST VALLEY, NEW YORK In August 1977 the NRC staff issued its initial assessment (Reference 1) of the overall safety of the dormant reprocessing plant at West Valley, New York.
We considered the effect of tornados on the reprocessing plant in that assess-ment (see pages 6 and 7 of the Interim Safety Evaluation I of 1977).
Although our conclusion from that assessment was that the risk of release of radio-activity from the process cells is acceptably low, we intended to obtain additional information on these effects with assistance from the Oak Ridge National Laboratory (ORNL).
This investigation has been completed and the results are discussed below.
Likelihood of Tornado Strike As discussed on pages 6 and 7 of the Interim Safety Evaluation 1 Report, our initial estimate of the recurrence interval for a tornado of the design basis size proposed by the NRC staff in 1974 was ten million years.
A more recent estimate (Reference 2) places the recurrence interval for that size tornado in the West Valley area at about one billion years or greater.
The recurrence interval for a tornado in that area having maximum windspeeds of two hundred miles an hour is estimated at two million years.
It is for this reason that the staff considers a tornado strike with wind speeds greater than two hundred miles per hour to be an incredible event.
However, the parametric 'nalysis by a
ORNL, which is described below, considered wind speeds up to and including three hundred miles per hour.
ORNL Analysis As discussed in the ORNL report (Reference 3) the suctions (pressure drops) created by the tornados can create abnormal flow of air within tne process cells of the plant, either in exceeding normal flow rates or in completely revarsing flow directions; i.e., working against the ventilation fans.
Various cases were analyzed, considering the location and site distribution of available radioactive particulates, sequence of strike (i.e., direction and rate of traverse of the tornado), possibility for various particulate escape pathways, and tornado size.
The TVENT computer code (Reference 4), developed at the Los Alamos Scientific Laboratory (LASL), was used to model the internal flow pathways and identify conditions requiring detailed analysis using the SOLA-ICE computer code (reference 5), also developed at LASL.
SOLA-ICE enables a detailed, two-dimensional analysis of flow conditions within the cells.
Although ORNL considers cases in which confining walls external to the process cells were assumed destroyed, a separate structural evaluation of that possi-bility indicates that the walls would remain intact (Reference 5).
- However, no analysis has been performed of doors, louvers or other possible openings external to the cells.
In any event, the results of the ORNL analysis are shown in Table 1.
The only pathway for release was the route through the manipulator sleeve ports in the Process Mechanical Cell.
C-1
Table 1 Estimate of range of curies of radioactivity induced from the Process Mechanical Cell by direct tornado strike (200 mph)
Fission Products Actinides External walls intact 162 - 1800 36 - 1800 Without external walls.
580 - 28000 131 - 6200 The results presented in Table 1 were first provided to the NRC in the fall of 1979.
The actual numerical values in the table were different as they were preliminary estimates.
The NRC staff felt that, although the numerical values might change, the backflow pathway from the Process Mechanical Cell to the corridors needed corrective action.
On December 10, 1979, the NRC staff met with the licensee to discuss these results.
On March 14, 1980, we were informed by the licensee that the manipulator sleeve ports and other similar cell openings had been sealed.
Our consultant, Dr. Wayne A. Coffman, concurred that the seals would remain intact in a tornado to protect against such a reverse flow condition (Reference 6).
The analysis by ORNL shows that the sealing of the backflow pathway through the manipulator sleeve ports would effectively assure the confinement of the radioactivity in the cells (Reference 3, pages 5-9).
Missile Protection In all cases of the ORNL analysis it was assumed that the integrity of the high efficiency particulate filters would be maintained through the course of the tornado strike.
This aspect was separately analyzed in Reference 5.
Although the Head End Ventilation Building was not specifically designed and constructed with the 200-mph tornado criteria, the analysis indicates sub-stantial protection against it.
A singular exception was discovered in the analyses in that a small portion of the north wall had insufficient impact resistance to withstand a direct missile strike and yet assure filter integrity.
In a meeting with the licensee on April 16, 1980, this aspect of the staff's analysis was discussed.
A corrective action was proposed by the licensee on August 18, 1980 and accepted by the staff on October 10, 1980.
This corrective action consists of a 12-inch thick, reinforced concrete shield set forth in front of the north wall.
This shield provides adequate protection against any credible missile (Reference 7).
Conclusion The analysis provided by our consultants of the credible impacts of a large tornado strike at the West Valley site indicated two critical weaknesses in the structure previously used for reprocessing, viz. 1) a backflow pathway C-2
from the Process Mechanical Cell, and 2) the need for additional protection against tornado generated missiles along the north wall of the Head-End Ven-tilation System enclosure.
Both of these deficiencies were brought to the attention of the licensee and appropriate corrective actions were taken.
Therefore, we conclude that the health and safety of the public is adequately protected against any credible tornado which could strike the site.
References 1.
Interim Safety Evaluation I, August 1977, Docket No. 50-201, Nuclear Fuel Services Inc. and New York State Energy Research and Development Authority, Western New York Nuclear Service Center, West Valley, N.Y.
2.
R. F. Abbey, Jr., " Analysis of Tornado and High Wind Speed Probabilities for West Valley, New York," NRC report, December 1981.
3.
L. J. Holloway and R. W. Andrae, " Potential Radiological Impact of Tornadoes on the Safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant."
I.
Tornado Effects o, Head-End Cell Airflow, NUREG/CR-1530, 1981.
4.
K. H. Duerre, R. W. Andrae, and W. S. Gregory, TVENT, A Computer Program for Analysis of Tornado-Induced Transients in Ventilation Systems, LA-7397-M (July 1978).
5.
W. A. Coffman, " Assessment of the Western New York State Nuclear Service Center Head End Ventilation Building's Structural Response to Tornadic Loading," March 1980, pp. 9-10, 15-20.
6.
Letter, Dr. Wayne A. Coffman to Dr. A. Thomas Clark, dated May 26, 1980.
7.
Letter, Dr. Wayne A. Coffman, Engineering Analysis & Data Services, to Dr. A. Thomas Clark, dated September 22, 1980.
C-3
APPENDIX D NRC STAFF COMMENT AND CONSEQUENCE ANALYSIS OF THE EARTHQUAKE RISK FROM THE NEUTRALIZED LIQUID WASTE TANKS AT THE WESTERN NEW YORK NUCLEAR SERVICE CENTER WEST VALLEY, N.Y.
DOCKET NO. 50-201 As stated in our Interim Safety Evaluation I of August 1977,1 the Nuclear Regulatory Commission (NRC) staff has been investigating the potential risk resulting from the effects of a severe earthquake on the neutralized liquid waste tanks and vaults at the Nuclear Fuel Services, Inc. (NFS) reprocessing plant in West Valley, New York.
The staff has contracted with the Lawrence Livermore Laboratory (LLL) to conduct the seismic-structural analyses of the waste tank and vaults.
LLL has reported the results of their analyses in
" Seismic Analysis of High Level Neutralized Liquid Waste Tanks at the Western New York State Nuclear Service Center, West Valley, New York," (UCRL-52458).
We have also asked Dr. William J. Hall of the Nathan M. Newmark Consulting Engineering Services to review and comment on these analyses.
Dr. Hall has extensive experience in seismic and analytical methods and in the actual effects of earthquake on structures.
His comments are presented in this letter report to the NRC staff dated March 2, 1979.
i Based on staff review of information submitted by Nuclear Fuel Services, Inc.
and its own, independent sources of information, the staff concluded that a site seismic design acceleration of 0.2 g was appropriate for new additions l
and modifications.
The tanks and vaults were built to Uniform Building Code Zone III specifications during the period 1964-1965.
Zone III indicates an active earthquake zone.
In our investigations, we have assessed the capability of the tanks and vaults against the design value of 0.2 g.
As discussed in our Interim Safety Evaluation the staff has also developed infor-mation on seismic recurrence intervals for the site.
These estimates are equivalent to predicting the probability of the occurrence of a single earth-quake as a function of its magnitude.
This relationship is shown in Figure 1.
The curve of Figure 1 represents a fit to Central Stable Region earthquake histories from modified Mercalli intensities IV, VII, and VIII.
1
(
Neutralized Waste Storaoe System Descriotion The neutralized nigh-level liquid waste storage system at West Valley consists of two separate 750,000 gallon carbon steel tanks each contained within its own reinforced concrete vault.
One of these tanks contains approximately 585,000 gallons of neutralized wastes, while the other tank serves as a l
spare.
l INuclear Regulatory Commission Staff, Interim Safety Evaluation I, August 1977, Docket No. 50-201, Nuclear Fuel Services, Inc. and New York State Energy Research and Development Authority, Western New York Nuclear Service Center, West Valley, N.Y.
D-1
.010 Figure 1 jEstimate of Probability ofj.
Earthquake at West Valley, N.Y.l j
i i
o
.001
?
~
cg e
~
u
~
u 8
C cc.cO
.0001 1.6x10*
years
.00001 l
I
.05g.109.15g.209 259.30g Peak Horizontal Acceleration, fraction of gravity D-2
1 The tanks and vaults are completely buried within the silty clay soil and have a minimum of eight feet of soil overburden covering the vault roof.
Several surface systems provide services to the underground tanks.
The tank ventila-tion system takes a suction on the tank vapor space, condenses and collects entrained water vapor, and filters airborne particulate matter before discharge to the reprocessing plant ventilation system.
The ability of the surface tank support systems to withstand the effects of earthquake has not been analyzed, and therefore in our analysis these systems are assumed to be damaged.
Seismic-Structural Analysis The results of the LLL analyses indicate that the carbon steel tanks are adequate as designed for ground accelerations up to and including 0.2 g.
The analyses further indicate that the coefficient of friction between the carbon steel tank and the perlite blocks is sufficiently high to preclude sliding of the tanks on the blocks and striking the vault.
The analyses indicate that sig-nificant cracking along the bottom circumference of the vault wall will occur at peak horizontal ground accelerations between 0.13 and 0.16 g.
The amount of cracking can be expected to increase for accelerations between 0.16 and 0.2 g; but no excessive compressive stresses, causing crushing of concrete or gross collapse of the vault, would occur up to 0.2 g.
Some spalling concrete from the inside of the vault roof could fall on the tank during such a severc earthquake.
There is a negligible likelihood that the tank could be damaged by this falling concrete.
As Dr. Hall states, "It is extremely unlikely.
that these [ concrete pieces] would puncture or penetrate the roof of the steel tank in view of the limited distance separating the vault roof and the tank roof, and in view of the tank roof thickness and the strength and ductility properties of the roof material."
Throughout this evaluation, the attempt has been made to strip the analysis of much of the conservatism inherent in the method.
This is not done because the staff does not want to be conservative in its approach to protection of the public; but rather, the removal of conservatism is necessary in order to require for any new construction the conservatism inherent in its guides and regulations and analyses.
The staff concludes from these analyses that the carbon steel high-level liquid waste tanks will remain intact following a severe earthquake.
Some cracking of the vaults can occur, but the vaults will remain standing.
Finally, there is a negligible possibility that damage to the vaults could result in breaching of the tanks.
Quantity of Radioactivity at Risk The radioactive waste in the on-service tank (80-2) exists in two phases - a relatively dense layer of sludge at the tank bottom composed of precipitated solids from the neutralization reaction and a liquid supernatant phase above the sludge.
The sludge is mostly composed of the hydroxides of process chem-icals and fission products which are insoluble in the strong basic pH of the solution.
The supernatant is chiefly sodium nitrate solution with ions of the few fission products (e.g., cesium) which remain soluble in the strongly basic soluticn.
The concentration of cesium-134 and cesium-137 in the supernatant are measured quarterly by NFS and reported to the NRC.
Basea on these measured values and D-3
c'alculations using reprocessed fuel burnup data and fuel discharge and decay data, the fission product and actinide inventory for tank 80-2 have been computed.
The most recent tabulation of this radionuclide invent.ory can be found in Tables 3.7 and 3.9 of the U.S. Department of Energy Commission Report to their Western New York Service Center Study (TID-28905-3).
This inventory, adjusted for decay up through 1979, has been used by the staff as the source term in its evaluation of the consequence of a severe earthquake on the neutralized high-level waste tank storage system.
Potential Releases Followino the Earthouake The equipment servicing the neutralized waste tanks was not designed to urrent e2ismic design criteria, but rather was designed using the Uniform Builcing Code for Zone III.
This 'urface equipment has not been analyzed for its s
capabity to withstand the effects of a severe earthquake.
For the purposes of this analysis, the staff has assumed that, in the event of a severe earthquake, the surface equipment could be damaged such that normal condensation and filtration of the waste tank vapors would no longer occur.
Furthermore, the staff has assumed that the waste tank vapors would then be directly released at ground level rather than condensed, filtered, and then released via the plant stack.
Under this assumption, the radioactivity contained in the tank vapor is that amount of radioactivity from the supernatant that is carried over into the vapor during the process of evaporation.
The decontamination factor (DF) for evaporating semivolatile species from the NFS waste tanks has been measured by comparing the cesium-137 concentration in the supernatant to the cesium-137 concentration in the knockout drum.
For the present conditions in the tank, with waste temperature about 180 F, this DF ranges between 5x105 and 7x10.
5 For the purposes of this analysis, the staff has used a 0F of 5x105 The present rate of evaporation from the waste tank is approximately 65,000 gallons per year.
Several years would be required to evaporate the 585,000 gallons of waste to dryness.
Thus, ample time is available for action
[
to be taken to prevent complete evaporation of the wastes.
A potential consequence ra.Jlting from the effects of a savere earthquake on the high-level waste storage system results from the inhalation of radionuclides that have evolved from the tank vapor space and have been dispersed to the surrounding atmosphere.
Meteorological dispersion factors were developed by the staff for various possible exposure times to individuals at the site boundary.
The airborne concentrations of the released radionuclides at the site boundary were computed and compared with the maximum permissible concentra-tions (MPCs) for release of airborne radionuclides to unrestricted areas during normal operation.
Tne MPCs are tabulated in 10 CFR 20, Appendix "B",
Table II.
For shorter postulated exposure times (e.g., one hour), the meteoro-I logical dispersion factors (X/Q) represent the poorest or least amount of dispersion that can be expected during a given day.
The X/Q values for longer exposure times (e.g., several weeks) represent a weighted average of all meteorological dispersion conditions that can ce expected within the particular time period.
l 0-4 l
The results of the comparison between calculated airb rne radioactivity concen-trations and the MPCs indicate that an individual at the closest site boundary exposed to the worst meteorological dispersion that could be expected in any given hour would inhale airborne concentrations of radioactivity that are less than 25% of 10 CFR 20 limits.
If an individual were to be exposed to the airborne radioactivity at the site boundary for six weeks, experiencing the effects of average meteorological dispersion that can be expected during that time, the average airborne concentrations would be less than 0.5% of 10 CFR 20 limits.
Frequently, when evaluating the consequences of extreme accidents such as a severe earthquake, comparisons are made with the much higher accident limitations contained in 10 CFR 100.
In this case, the predicted radioactivity concentrations are less than that which ir permitted for coatinuous, unrestricted release of radioactive effluents under normal operating conditions.
Following the postulated earthquake, action would be taken by operators to restore any damaged tank ventilation systems.
The time required for operators to restore ventilation, condensation, and filtration following an earthquake is unknown.
If damage to aboveground equipment is extensive, several weeks may be required to complete repairs.
Temporary systems could be installed in an emergency if necessary.
A temporary tie-in to the main reprocessing plant ventilation system could also be accomplished in order to diminish the amount of activity released.
It is expected that personnel performing this work would require respiratory protection.
Radiolytic Decomoosition of the Waste The radiation field in the high-level liquid waste causes radiol 3 tic decomposition of water into hydrogen and oxygen gas.
This same radiation field is sufficient to heat the liquid waste and cause evaporation of the water.
The gamma flux due to cesium-137 is the most important sour:e of the radiation field in the supernatant.
Based upon the amount of gamma flux pre-sent, the staff has estimated that the amount of hy irogen gas produced by radiolytic dacomposition may be as high as 5.5 standard cubic feet per hour.
At the same time, the production rate of steam due to evaporation of water caused by the same radiation field is approximately 3300 standard cubic feet per hour, which is six hundred times the production rate of hydrogen.
During normal operation, the hydrogen gas produced in the waste will col Rct in the tank vapor space and be removed by the ventilation system.
The mea-sured hydrogen concentration in the effluent vapor stream has been less than one percent.
After discharge to the atmosphere, the hydrogen will rapidly diffuse and will also recombine with oxygen in the atmosphere.
?ollowing a severe earthquake, the staff has assumed, in the absence of formal structural analysis, that the tank ventilation system is damaged and inoper-ative.
In order to conservatively determine potential radiological consequences, the staff has further assumed that severed piping from the tank permits the ground level release of vapors from the tank.
Hydrogen gas would be released along with other components of the vapor.
Regardless of the condition of the tank support systems following a severe earthquake, the production rate of steam is so much larger than the production rate of hydrogen that conditions within the tank vapor space would be further driven from flammibility limits.
0-5
O Conclusion The staff and its consultants have examined the effects of a major earthquale on the neutralized high-level liquid waste storage tanks at the reprocessing plant ir. West Valley, New York.
The carbon steel tanks as designed would withstand the effects of the earthquake, and the surrounding concrete vaults would remain standing.
In the absence of a formal seismic-structural analysis of the aboveground structures and equipment that provide support services to the tanks, the staff has assumed that damage to this aboveground equipment would permit the release of radioactivity from the tank vapor space.
The resulting airborne concentrations of radionuclides at the site boundary would be less than permitted for unrestricted release during normal operations.
I
'l 0-6
(
APPENDIX E NRC STAFF COMMENT AND CONSEQUENCE ANALYSIS OF THE EARTHQUAKE RISK FROM THE ACID LIQUID WASTE TANKS AT THE WESTERN NEW YORK NUCLEAR SERVICE CENTER WEST VALLEY, NEW YORK 00CKET NO. 50-201 As stated in our Interim Safety Evaluation I issued in August 1977 (Ref. 1),
the Nuclear Regulatory Commission (NRC) staff has been investigating the potential risk resulting from the effects of a severe earthquake on the acid liquid waste tanks and vault at the Nuclear Fuel Services, Inc. (NFS) repro-cessing plant in West Valley, New York.
The staff has contracted with the Lawrence Livermore Laboratory (LLL) to conduct the seismic-structural analyses of the waste tanks and vault.
LLL has reported the results of their analyses in " Seismic Analysis of the Acid Liquid Waste Tanks at the Western New York Nuclear Service Center, West Valley, New York."
(UCRL-52600).
We have also asked Dr. William J. Hal.1 of the Nathan M. Newmark Consulting Engineering Services to review and comment on these analyses.
Dr. Hall has extensive experience in seismic analytical methods and in the actual effects of earthquake on structures.
His comments are presented in his letter report to the NRC sta dated May 25,1979 (Ref. 2).
Based on staff review of information submitted by Nuclear Fuel Services, Inc.
and its own inoependent source of information, the staff concluded that a site seismic design acceleration of 0.2 g was appropriate for new additions and modifications.
The tanks and vault were built to Uniform Building Code Zone III specifications during the period 1964-1965.
Zone III indicates an activity earthquake zone.
In our investigations, we have assessed the capability of the tanks and vault against the desigr. value of 0.2 g.
As discussed in our Interim Safety Evaluation, the staff has also developed information on seismic recurrence intervals for the site.
These estimates are equivalent to predicting the probability of the occurrence of a single earth-quake as a function of its magr.itude.
This relationship is shown in Figure 1.
The curve of Figure 1 represents a fit to Central Stable Regier. earthquake histories from modified Mercalli intensities IV, VII, and VIII.
Acid Waste Storace System Descriotion The acid high-level liquid waste storage system at West Valley consists of two separate 15,000 gallon stainless steel tanks, both contained within a single reinforced concrete vault.
One of these tanks contains approximately 12,000 gallons of acid wastes, while the other tank serves as a spare.
The tanks and vaults re completely buried within the silty clay soil and have a minimum of six feet of soil overburden covering the vault roof.
Several surface system, provide services to the underground tanks.
The tank ventil-ation system takes a suction on the tank vapor space, condenses and collects entrained water vapor, scrubs entrained acidic vapors, and filters airborne particulate matter before discharge to the reprocessing plant ventilation system.
The ability of the surface tank support systems to withstand the effects of earthquake has not been analyzed; and in our analysis, these systems are assumed to be damaged.
E-1
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Figure 1 JEstimate of Probability ofj Earthquake at West Valley, N.Y.
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Seismic-Struc,tural Analysis The results of the LLL analyses indicate that the stainless steel tanks and vault are adequate as designed for ground accelerations up to and including 0.2 g.
The analyses of the tanks further indicate that the only component of the tanks that could exceed yield strength is the base plate of the tank support legs.
Although some yielding cf the base plate could occur, the supporting function of the legs would not be impaired.
The vault is expected to experi~
seme cracking near the roof to wall junction at peak horizontal ground accelerations between 0.18 and 0.2 g.
This cracking is limited so that l
the vault would remain stancing after the earthquake.
Throughout this evaluation, the attempt has been made to strip the analysis of much of the conservatism inherent in the method.
This is not done because the staff does not want to be conservative in its approach to protection of the
)
public; but rather, the removal of conservatism is necessary in order to realistically determine the risk to the public.
The staff continues to require for any new construction the conservatism inhere:.t in its guides and regulations i
and analyses.
The staff concludes from these analyses that the stainless steel high-level liquid waste tanks will remain intact following a severe l
Some cracking of the vault can occur, but the vault will remain j
standing.
l Ouantity of Radioactivity at Risk l
The radioactive waste on the on-servce stainless steel tank (80-4) exists in a single liquid, acidic phase.
The fissico products and actinides remain soluble in the strongly acidic solution.
The radionuclide inventory in 8D-4 has been previously reported by NFS and others (Refs. 3 and 4) on the basis of fuel burnup and decay data and Thorex reprocessing data.
A small samole of acidic waste from 8D-4 was analyzed by the Oak Ridge National Laboratory (ORNL) in 1978.
The results of the ORNL analyses were included in the DOE study report on the West Valley site (Ref. 41 This radionuclide inventory, adjusted for decay up through 1979, has been used by the staff as the source term in its evaluation of the consequence of a severe parthquake on the acidic high-level waste tank storage system.
Potential Releases Following the Earthquake The equipment servicing the acid waste tanks was not designed to current seismic design criteria, but rather was designed using the Uniform Building Code for Zone III.
This equipment has not been analyzed for its capacity to withstand the effect of a severe earthquake.
For the purposes of this analy-sis, the staff has assumed that, in the event of a severe earthquake, the equipment could be damaged such that normal condensation and filtration of the l
waste tank vapors would no longer occur.
Furthermore, the staff has assumed
}
that the waste tank vapors would then be directly released at ground level rather than condensed, filtered,.and then released via the plant stack.
Under the assumption, the radicactivity contained in the tank vapor is that amount of radioactivity from the liquid that is carried over into the vapor during the process of evaporation.
The decontamination factor (DF) of 5 x 10s E-3
reviously used by the staff in its analysis of the effects of earthquake on the neutralized high-level waste storage system (Ref. 5) is again used by the staff in this analysis.
The rate of evaporation from 8D-4 has not been measured, but is presently quite low due to the low temperature of the waste (110 F to 120 F).
In the event of a severe earthquake, with the tank cooling systent possibly disabled, the w:ste temperature and evaporation rates could be expected to rise.
In the absence of measured evaporation rate data for tank 80-4, the staff has calculated an evaporation rate based on the conservative assumption that the entire waste heat generation is removed by evaporation, with no waste heat removed by the mechanisms of convection, conduction, or radiation.
This conservatively calculated evaporation rate is 4.9 gallons per hour.
A potential consequence resulting from the effects of a severe earthquake on the high-level waste storage system results from the inhalation af radionuclides that have eva bed from the tank vapor space and have been dispersed to the surrounding atmosphere.
Meteorological dispersion factors were developed by the staff for various possible exposure times to irdividuals at the site boundary.
The airborne concentrations of the released radionuclides at the site boundary were computed along witn the maximum inhalation doses to hypo-thetical individuals positioned at the site boundary.
The doses were computed using the REDIQ computer code (Ref. 6) which uses the Task Group Lung Model for computing inhalation doses.
The results of dose calculations indicate that an individual at the closest site boundary exposed to the worst meteorological dispersion that could be expected in any given hour would receive a lifetime (50 year) dose commitment of approximately 0.7 millirem to the lungs, 3 millirem to the liver, 7.4 mil-lirem to the bone, and 1 millirem to the whole body.
The predominant isotopes contributing to the dose are strontium-90 and plutonium-238.
If an individual were to be exposed to the airborne radioactivity at the site boundary for six weeks, experiencing the effects of average meteorological dispersion that can be expected during the time, that individual would receive a lifetime (50 year) dose commitment of approximately 9 millirem to tre lungs, 40 millirem to the liver, 100 millirem to the bone, and 13 millirem to the whole body.
Again, the principal contributors to the dose are strontium-90 and plutonium-238.
Following the postulated earthquake, action would be taken by operators to restore any damaged tank ventilation systems.
The time required for operators to restore ventilation, condensation, and filtration following an earthquake is unknown.
If damage to aboveground equipment is extensive, several weeks may be required to complete repairs.
Temporary systems could be installed in an emergency if necessary.
A temporary tie-in to the nain reprocessing plant ventilation system could also be accomplished in order to diminish the amount of activity released.
It is expected that personnel performing this work would require respiratory protection.
Radiolytic Decomoosition of the Waste The radiation field in the high-level liquid waste causes radiolytic decomposition of water into hydrogen and oxygen gas.
The same radiation field E-4
is sufficient to heat the liquid waste and cause evaporation of the water.
The decay of cesium-137, stror. tium-90, and yttrium-90 produce the most important sources of the radiation field in the waste.
Based upon the amount of radiation present, the staff has estimated that the amount of hydrogen gas produced by radiolytic decomposition may be as high as 2.2 standard cubic feet per hour.
The production rate of steam due to evaporation of the water is many times larger than the production of hydrogen.
During normal operation, the hydrogen gas produced in the waste will collect in the tank vapor space and be removed by the ventilation sy-The mea-sured hydrogen concentration in the effluent vapor stream has. en less than one percent.
After discharge to the atinosphere, the hydrogen will rapidly diffuse and will also recombine with oxygen in the atmosphere.
Following a severe earthquake, the staff has assumed, in the absence of formal structural analysis, that the tank ventilation system is damaged and inoper-ative.
In order to conservatively determine potential radiological consequences, the staff has further assumed that severed piping from the tank permits the ground level release of vapors from the tank.
Hydrogen gas would be released along with other components of the vapor.
Regardless of the condition of the tank support systems following a severe carthquake, the production rate of steam is so much larger than the production rate of hydrogen that conditions within the tank vapor space would be further driven from flammability limits.
Conclusion The staff and its consultants have examined the effects of a major earthquake on the acid high-level liquid waste storage tanks at the reprocessing plant in West Valley, New York.
The stainless steel tanks as designed woul' withstand the effects of an earthquake, and the surrounding concrete vault would remain standing.
In the absence of a formal seismic-structural analysis of the above-
, ground structures and equipment that provide support services to the tanks, the staff has assumed that damage to this aboveground equipment would permit the release of radioactivity from the tank vapor space.
The resulting air-borne concentrations of radionuclides at the site boundary would produce a dose to individuals at the site boundary that is only a fraction of the natural radiation background (about 100 millirem / year whole body).
The dose would also be many times lower than doses specified under the accident design values of 10 CFR 100.
i E-5
References 1.
Nuclear Regulator Commission Staff, Interim Safety Evaluation I, August 1977, Docket No. 50-201, Nuclear Fuel Services, Inc. and New York State Energy Rsearch and Development Authority, Western New York Nuclear Service Center, West Valley, N.Y.
2.
Letter from Dr. W. J. Hall, Nathan M. Newmark Consulting Engineering Services, to Dr. A. T. Clark, U.S. Nuclear Regulatory Commission dated May 25, 1979.
3.
" Alternative Processes for Managing Existing Commerical High Level Radioactive Wastes," Battelle Pacific Northwest Laboratory (NUREG-0043),
dated April 1976.
4.
" Western New York Nuclear Service Center Study," Companion Report, U.S. Department of Energy (TID-28905-3).
5.
NRC Staff Comment and Consequence Analysis of the Earthquake Risk from the Neutralized Liquid Waste Tank at the Western New York Nuclear Service Center, West Valley, N.Y., enclosure (3) to letter from L. C.
Rouse, NRC, to D. W. Deuster, NFS, and J. Larocca, NYSERDA dated June 20, 1979.
6.
D. L. Strenge, E. C. Watson, W. E. Kennedy, "REDIQ - A Computer Program for Estimating Health Effects from Inhalation and Ingestion of Radio-nuclides," Battelle Pacific Northwest Laboratories (BNWL-2110),
December 1976.
E-6
o APPENDIX F NRC EVALUATION OF THE SAFETY ASSOCIATED WITH THE DEFECT IN PAN 80-2 AT WEST VALLEY, NEW YORK
Background
The NFS high-level liquid waste storage system consists of:
(1) the waste tank; (2) a steel pan which surrounds the bottom of the tank and projects upward a portion of the height of the waste tank; and (3) a concrete vault.
The underground vault is embedded in a glacially deposited silty till soil of low permeability and high ion exchange capacity.
The facility was constructed with two of these vault pan-tank systems for the primary liquid high-level waste generated by reprocessing operations.
One tank, designated 80-2, con-tains high-level waste (s585,000 gallons) while tank 80-1 is maintained as a spare.
A drawing of a sectional elevation view of one of the waste storage tanks is attached.
As part of a continuing safety evaluation of the storage of the high-level waste in tank 80-2, NRC requested on June 27, 1978 that NFS estimate the time required to transfer waste from 80-2 to the spare tank 80-1.
Included in the various tests, outlined by NFS to us in their October 12, 1978 letter, was the addition of water to the 80-2 pan to verify instrument response.
An alarm in the pan, designed to warn that waste was leaking from the tank to the pan, failed to.*espond as expected when pan water level reached the alarm set point.
Subequent investigation of this problem resulted in NFS receiving indication of rising watee level in the concrete vault, indicating that the 80-2 pan leaked.
An additional problem was discovered with the 80-2 vault liquid level detection system.
The as-built detection probe is twelve (12) inches shorter than designed and thus does not extend to within one inch of the vault bottom as required by construction drawings.
In ligia of these discoveries, the pan under the spare tank 80-1 was similarly tested.
The 80-1 pan did not leak.
The annual space between the tank and the vault in 80-1 has been entered by both NFS personnel and NRC staff.
Nt conditions were found indicating that 80-1 would be unsuitable to receive waste.
Puroose of the Pan A pan is provided beneath each tank to hold waste if a small leak should The pan has a capacity of approximately 30,000 gallons, while the tank occur.
has an operational maximum capacity of 600,000 gallons.
Therefore the pan can hold only a relatively small fraction of the entire waste tank volume and was not designed to serve as a comolete, integral waste barrier in the event of a tank rupture.
Rather, it was designed to act as an interim collection chamber or barrier for small leaks.
Ther e has been no leakage from the NFS waste tanks.
Experience at the Department of Energy (DOE) defense sites, Hanford and Savannah River, has shown that leakage rates from waste tanks have been very small - normally i
fractions of a gallon per minute (gpm) and never more than a few gpm.
Thus, a
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F-1
i 1
sound pan with a small volume and a520 gpm pump can serve as an adequate viaste i
collec'. ion chamber pending waste transfer frcm a leaking tank.
In the unlikei, event of a large tank leak, waste would overflow the pan into the vault.
The vault is provided with a steam eductor to remove waste from the vault and return it to the pan.
Explanation of Pan Defect The cause of the pan defect is unknown at present.
A possible explanation may be the torch cuttir.g of semicircular holes in the skirt around the perlite blocks.
The semicircular holes, fabricated during construction, were designed to allow passage, and therefore, detection of liquid waste from beneath the tank to the annular space, if a tank leak were to occur.
During the cutting operation, it is theorized that base metal in the pan may have been penetrated forming a flow path to the vault.
(See note on sectivn G-G on attached drawing.)
NFS review of documentation concerning postconstruction and preacceptance testing of the waste storage system has revealed no evidence that a standing water operational test was conducted to test pan integrity priot to use.
Thus, the po'ssibility exists that the defect in the 80-2 pan has existed since construction.
Another possible cause of the defect is accelerated local corrosion.
Such a phenomenon could have been enhanced in regions around pan welds and/or by pockets of moisture at the pan to pea gravel interfaces.
Because of the high radiation field in the annular space around the tank, the tank cannot be entered by personnel for direct inspection.
Some inspection is possible by remote methods.
The staff has recently contracted with Rockwell International-Hanford to perform various inspections.
Such inspections cannot be conducted without a thorough safety analysis of each technique.
Upon completion of an appropriate inspection plan, including a safety analysis, inspection procedures will begin.
The condition of the tank, vault, and pan may be better understood after the inspection.
These investigations to be performed by Rockwell, under contract to NRC, are expected to take place over the next two years.
Repair of Defect Attempts to repair the defect in the pan under tank 80-2 may be neither beneficial nor possible and certainly should not be undertaken without con-siderable analysis.
A repair undertaken in haste and as a result of inadequate planning could produce further problems in the 80-2 waste storage system.
Tank Condition In an interim safety evaluation report issued in August 1977, the staff concluded that tank 80-2 will continue to safely contain the high-level waste:
F-2
1.
Tank 80-2 has not leaked since being placed in service in 1966.
2.
Corrosion coupon inspection from tank 8D-2 indicates a general corrosion rate much less than design.
3.
Tank temperatures are less than design.
4.
Tank 8D-2 was heat-treated for stress relief during construction to avoid stress corrosion cracking.
5.
All carbon steel tanks at DOE defer's sites, which have been stress relieved and placed in similar chemical service, have not leaked.
Additionally, recent results of seismic analyses indicate that even the largest credible earthquake would not cause tank failure.
Transfer of Waste from 8D-2 to 80-1 The use of spare tanks is considered in Technical Specification 5.4, Spare Waste Storage Capacity, of NFS License No. CFS-1.
This specification requires replacement of a spare tank at the earliest practical date if a spare tank is activated (waste is added) or found unsuitable.
Although not explicitly stated in the specification, it is the staff view that waste should be trans-ferred only if the present tank leaks or is found through proper engineering inspections to be deteriorated and in imminent danger of leaking.
The equip-ment and procedures for waste transfer have been prepared by NFS and are ready for use, although no transfar of this nature has been made at the site.
Conclusions As presently stored, the principle confinement barriers for the waste area are:
- 1) the carbon steel tank, 2) the concrete vault with its surrounding water injection system, and 3) the glacial deposits (silty till) in which the tank vault system is imbedded.
None of these barriers has been compromised.
The threat to the public, or to operating personnel, has not changed appreciably as a result of the defect discovered in the pan.
Therefore, the staff position is that the high-level waste in tant,80-2 should not be transferred at this time to the spare tank 8D-1.
In addition, no attempt should be made to repair the defect in the pan under tank 8D-2 without further detailed analysis of its cause or the possible effects of the repair.
F-3
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s APPENDIX G GLOSSARY AEC Atomic Energy Commission (now NRC)
AOC Acid and off gas recovery cell CPC Chemical process cell DOE Department of Energy DF decontamination factor EPC extraction and product purification cells
. FRS fuel-receiving station GPC general purpose cell LASL Los Alamos Scientific Laboratory LLL Lawrer.ce Livermore Laboratory LWC liquid waste cell MCR mechanical crane room MPC maximum possible concentrations NFS Nuclear Fuel Services, Inc.
NRC.
U.S. Nuclear Regulatory Commission ORNL Oak Ridge National Laboratory PMC process mechanical cell SAI Science Applications, Inc.
TOP concrete block structures on top of CPC UPC uranium product cell O
e l
G-1