ML18085A172
| ML18085A172 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/27/1980 |
| From: | John Lamb, Milhollin G, Shon F Atomic Safety and Licensing Board Panel |
| To: | |
| References | |
| ISSUANCES-OLA, LBP-80-27, NUDOCS 8010310447 | |
| Download: ML18085A172 (48) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Gary L. Milhollin, Chairman Dr. James C. Lamb, III, Member Frederick J. Shon, Member LBP-80-27 In the Matter of
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY, et al.
(Salem Nuclear Generating Station, Unit 1)
Docket No. 50-272 OLA (Spent Fuel Pool)
October 27, 1980 INITIAL DECISION Appearances Mark J. Wetterhahn, Esq., of Conner & Moore, Washington, D.C., and Richard Fryling, Jr., Esq.,
of Public Service Electric and Gas Company, Newark, New Jersey, for the Public Service Electric and Gas Company, et al., Licensees.
R. William Potter, Esq., Keith A. Onsdorff, Esq.,
Menasha J. Tausner, Esq. and Sandra T. Ayres, Esq.,
Assistant Deputy Public Advocates, State of New Jersey, for Mr. and Mrs. Alfred C. Coleman, Jr.
Carl Valore, Jr., Esq., of Valore, McAllister, Aron and Westmoreland, Northfield, New Jersey, for the Township of Lower Alloways Creek.
Richard M. Hluchan, Esq., and Rebecca Fields, Esq.,
Deputy Attorney General, Department of Law and Public Safety, for the State of New Jersey.
June D. MacArtor, Esq., Deputy Attorney General for the State of Delaware.
Barry H. Smith, Esq., Janice E. Moore, Esq. and William D.
Paton, Esq., Office of the Executive Legal Director, U.S. Nuclear Regulatory Commission, Washington, D.C.,
for the NRC Staff.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Gary L. Milhollin, Chairman Dr. James C. Lamb, III, Member Frederick J. Shon, Member In the Matter of
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PUBLIC SERVICE ELECTRIC AND
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GAS COMPANY, et al.
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(Salem Nuclear Generating
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Station, Unit 1)
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Docket No. 50-272 OLA (Spent Fuel Pool)
INITIAL DECISION Summary This is a decision on an application by Public Service Electric and Gas Company (Licensee) to increase the storage capacity of the spent fuel pool at the site of its Salem Nuclear Generating Station, Unit No. 1.
The Licensee wishes to install new storage racks which would permit the storage of additional spent fuel assemblies in the existing pool area.
In the following decision this Board grants the permission sought in the application.
We find that the additional storage can be accomplished without endangering the health or safety of the public, and find no merit in contentions that the new racks will deteriorate or that the Licensee has not considered sufficiently the possible alternatives to the proposed en-largement.
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I.
INTRODUCTION
- 1.
On November 18, 1977 the Licensee applied to the Nuclear Regulatory Commission (NRC) for an amendment to the operating license for the Salem Nuclear Generating Station, Unit 1.
The application was re-filed in revised form on February 14, 1978.
The amendment would allow the storage capacity of the spent fuel pool to be increased from 264 to 1170 spent fuel assemblies.
- 2.
On February 8, 1978 the NRC in response to the application published a "Notice of Proposed Issuance of Amend-ment to Facility Operating License" in the Federal Register (43 Fed. Reg. 5443).
In response to this notice the States of Delaware and New Jersey filed petitions for leave to participate as interested states under 10 CFR §2*. 715(c).
The Township of Lower Alloways Creek, the Sun People, and Mr. and Mrs. Alfred C. Coleman, Jr. filed timely petitions to intervene as parties.
This Atomic Safety and Licensing Board was es-tablished on March 16, 1978 to rule on petitions for leave to intervene, and on April 24, 1978 this Board was designated to conduct hearings.
Mr. Glenn 0. Bright, who was originally designated as a member of this Board, was replaced by Mr.
Lester Kornblith on March 8, 1979 and Mr. Kornblith was re-placed on June 27, 1979 by Mr. Frederick J. Shon.
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- 3.
We held a Special Prehearing Conference in Salem, New Jersey on May 18, 1978 to consider petitions to inter-vene.
The States of Delaware and New Jersey were granted leave to intervene as interested states, and the Colemans and the Township of Lower Alloways Creek were granted leave to intervene as parties.
The petition of the Sun People was denied.
- 4.
The Colemans originally filed 20 contentions, which they later reduced to 13.
After rev~ew we found only four (Contentions 2, 6, 9 and 13) to be admissible.
On February 27, 1979 the Licensee filed a motion for summary disposition of these remaining four.
We granted the motion as to Contentions 9 andl3; thus, only Colemans' Contentions 2 and 6 remained for litigation.
- 5.
Lower Alloways Creek Township originally filed 11 contentions, of which we found only two (Contentions 1 and 3) to be admissible after our first review.
Subsequently, we dismissed Contention 3 in response to the Licensee's motion for summary disposition of February 27, 1979.
Only Contention 1 remained for litigation.
- 6.
We held a second prehearing conference in Salem, New Jersey on March 15 and 16, 1979, at which numerous state-ments were received from members of the general public pursuant to 10 CFR §2.715(a).
The Staff of the Nuclear Regulatory Commission-responded directly to a number of the questions which were raised by these statements.
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- 7.
We held evidentiary hearings in Salem, New Jersey on May 2, 3, and 4, 1979 on the Colemans' Contentions 2 and 6 and on July 10, 1979 on the Township's Contention 1.
Evidentiary hearings were also held.on *July 11, 1979 on two of three questions which the Board itself posed to the parties on April 18, 1979.
These questions sought to determine what the effect would be on Salem's spent fuel pool if an accident similar to that at Three Mile Island 2 were to occur at Salem.
Finally, we held evidentiary hearings on April 28, 29, and 30, 1980 on still another question we posed.
This last question sought to determine, in the event of a gross loss of water from the spent fuel pool, what the difference in consequences would be between those occasioned by the spent fuel pool with expanded storage and the present pool.
- 8.
We have considered the entire record of this pro-ceeding and all of the proposed findings of fact and conclusions of law submitted by the parties.
Any proposed finding of fact or conclusion of law which is not incorporated in this initial decision is rejected as unsupported in law or in fact, or as unnecessary to this decision.
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II.
FINDINGS OF FACT A.
Colemans' Contentions 2 and 6 Colemans' Contentions 2 and 6 state:
- 2.
The Licensee has given inadequate con-sideration to the occurrence of accidental criticality due to the increased density _;
or compaction of the spent fuel assemblies.
Additional consideration of criticality is required due to the following:
A.
deterioration of the neutron absorption material provided by the Baral plates located between the spent fuel bundles.
B.
deterioration of the rack structure leading to failure of the rack and con-sequent dislodging of the spent fuel bundles.
- 6.
The Licensee has given inadequate considera-tion to qualification and testing of Baral material in the environment of protracted association with spent nuclear fuel, in order to validate its continued properties for reactivity control and integrity.
We consolidated the two Contentions for consideration (Board Order dated May 24, 1978) and treat them together here.
- 9.
Evidence on these Contentions was presented by the Licensee and the NRC Staff.
Mr. Edwin A. Liden, Mr. Robert P.
Douglas, Mr. Warren S. Nechodom, and Mr. Thomas G. Eckhart appeared as witnesses on behalf of the Licensee.
Dr. John Weeks, Mr. Gary Zech and Mr. Edward Lanz appeared as witnesses on behalf of the NRC Staff.
The Colemans presented no testimony, although they and other parties conducted cross-examination.
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- 10.
As stated above, the Licensee proposes in this application to increase the storage capability of the spent fuel pool so as to allow the storage of 1170 spent fuel assemblies instead of 264.
- This would be accomplished by replacing the present spent fuel storage racks with new racks which enab:J_e the assemblies to be stored cl9ser t9getJ:ier.
By storing the assemblies in this denser array, additional assemblies can be accommodated in the existing pool area.
The new racks would decrease the spacing between assemblies from 21 to 10.51inches center-to-center.
The racks consist of an assemblage of hollow, open-ended double-walled stainless steel cells.
Each cell is a square 8.97 inches on a side and 14 feet long.
Each cell will receive.one spent fuel assembly in its cavity.
In order to reduce the number of neutrons travelling from one spent fuel assembly to another, and prevent a self-sustaining chain-reaction within the pool, plates made of Boral (boron carbide and aluminum) are to be fitted and welded into the gap between the double stainless steel walls forming each of the four sides of the cells.
Boral absorbs neutrons.
The result of this construction is that each storage location would consist of a hollow square cell 14 feet long having a sheet of neutron-absorbing Boral enclosed within each of its four stain-less steel sides (Exhibit 6-B, the Staff's Safety Evaluation, at 1-1, 2-17).
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Criticality Calculations
- 11.
Criticality is a measure of the capability of the neutron field within the pool to sustain a chain reaction.
This is expressed 1 by indicating the effective multiplication factor for neutrons (keff)' which is the ratio of the number of neutrons produced from fissions in each generation to the number of neutrons produced in the preceding generation.
To achieve criticality the keff must equal 1.0.
The acceptance criterion established by NRC for spent fuel pools is a cal-culated keff of 0.95 or less (id. at 2-2).
- 12.
The criticality calculations for this application were performed by Exxon Nuclear Company, which is responsible for supplying the new racks to the Licensee.
The calculations indicated that, when loaded with not more than 44.7 grams of U-235 per axial centimeter, the proposed installation will produce a.keff less than 0. 95 (id. at 2-1 through 2-3).
A technical specification on fuel loading will insure that this acceptance criterion will be met (id. at 2-3).
- 13.
This calculation of the keff was made without con-sidering the effect of the boric acid which will be contained in the water which surrounds the racks in the pool.
(Tr. 596).
In response to questions, the Licensee's witnesses stated that the boric acid concentration in the pool water is about 2000 ppm boron (Tr. 444-48, 736) and that~ concentration of this amount
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would be adequate to prevent criticality even without the Bora! plates and under the most disadvantageous fuel loading condition (Tr. 576-77).
Also, the Licensee testified that even without boron in the water, and with an entire Bora! plate missing, in a 5 x 5 array, keff still would be less than 0~9 (Tr. 576).
The NRC Staff testified that the overall results of the criticality calculations for the proposed racks compared favorably with those for other similar spent fuel pools (Exhibit 6-B at 2-2).
Deterioration of the Rack Structure
- 14.
The Colemans' Contention 2 asserts that the above con-sideration -Of criticality is inadequate in light of the possibility that the Bora! and the rack structure could deteriorate in the pool environment.
In response to this I
Contention, the Licensee's witnesses testified that the Type 304 stainless steel specified for the rack structure and the cell walls has been used widely in the nuclear industry, that it is the same material approved for use in fabricating the present Salem racks, and that it was chosen for its compatibility with water containing boric acid at 2000 ppm boron (Exhibit 2 at 2).
The witnesses stated that they are unaware of any deterioration of this type of stainless steel in environments similar to the Salem spent fuel pool (Exhibit 2 at 2, Tr. 455-456).
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- 15.
Dr. John R. Weeks., who testified in behalf of the NRC Staff, stated that no significant deterioration of the racks would occur, and that the stainless steel is protected from corrosion by a tenacious passivating film (Affidavit of John R. Weeks, following Tr. 652 at 2, March 29, 1979).
He also stated that corrosion rates of stainless steel in a spent fuel pool environment are too low to measure (ibid.).
Al-though stress corrosion cracking near welds is possible because of sensitizing by heat (id. at 3), Dr. Weeks concluded that such a phenomenon would be rare and localized and unlikely to affect rack integrity in the fuel pool (id.).
He reports that welded stainless steel liners have been in service for up to twelve years in borated spent fuel pools without failure through stress corrosion (id.).
The Colemans did not file proposed findings of fact opposed to this testimony.
Deterioration of the Neutron Absorption Material (Boral)
- 16.
In response to the assertion that the Boral itself may deteriorate, witnesses for the Licensee stated that the design of the new racks is such that the Boral sheets would be enclosed completely in the welded stainless steel cell walls so as to separate the Boral from the pool water and provide pro-tection against corrosion (Tr. 443, 574).
These witnesses also described manufacturing process controls and non-destructive
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testing of the finished cells which are designed to insure, at a 95% confidence level, that at least 95% of the cells will be leak-tight (Tr. 458, 492, 495, 616-618).
In response to Board questions, the witnesses interpreted this statement as an estimate that no more than 20-30 cells would be expected to leak out of the total 1170 in the modified Salem pool (Tr. 770).
Also, the quality assurance program would include inspection of the racks upon arrival at Salem to insure absence of damage during shipment (Tr. 494).
- 17.
The witnesses generally agreed that the Boral would corrode if it came into contact with the pool water (affidavit of John R. Weeks, supra, at 4).
To determine the extent of corrosion, the Licensee used data gathered during a one year test program and extrapolated to determine the corrosion rates to be expected over the life of the pool (In Camera Tr. 40).
The Licensee's testimony indicated that the method of extrapola-tion used (semi-logarithmic) is consistent with widely accepted practices in industry for determining long-term effects (Tr. 565-567).
Dr. Weeks agreed, and stated. that semi-logarithmic extrapola-tion may even be too conservative (Tr. 693-694).
- 18.
The corrosion observed by the Licensee during the test program consist~d of pitting, edge attac~ and the formation of small bulges in the Boral sheet (Exhibit 2 at 4, Exhibit 5 at i, In Camera Tr. 22).
None of this corrosion, however, would
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significantly reduce the capability of Boral to absorb neutrons because boron carbide is inert not only to the pool water (Affidavit of John R. Weeks, supra, at 4) but even to acid solutions which are much stronger than that of the pool (Tr. 664).
When corrosion does occur the boron carbide particles, rather than falling away, become imbedded in the corrosion products and remain in place (Affidavit of John R.
Weeks, supra, at 2; Exhibit 3 at 2-39).
Thus, their neutron-absorbing capability is not appreciably reduced by corrosion.
Swelling of Cell Walls
- 19.
The Licensee's testimony described an unfavorable incident which occurred at the Monticello facility.
Leaks near the bottom of some of the fuel storage cells allowed water to enter the cells' walls.
The water corroded the Bora!, p~oducing hydrogen gas.
Pressure exerted by the gas caused the cell walls to swell inward (Tr. 439-440) to such an extent that a fuel assembly, if stored in one of these cells, could not be removed (Exhibit 6-B at 2-13).
To alleviate this condition a small vent hole was drilled at the top of each storage cell to allow the gas to escape and to prevent similar pressurization in the future (Tr. 440).
Subsequently, the vented cells at Monticello were used to store spent fuel (Tr. 608-609).
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- 20.
At Salem, a similar condition could arise if water leaked into the walls of the cells.
Dr. Weeks testified that water entering the region in which the Boral is enclosed would corrode the Boral, produce hydrogen, and cause swelling of the stainless steel walls (Affidavit of John R. Weeks, supra, at 4).
The Licensee's testimony indicated that the amount of gas pressure within the cell would depend upon the elevation of the leak, with maximum pressure in instances where the leak occurs at the bottom of the cell (Exhibit 2 at 5).
This pressure. could cause the inner wall of the cell to bulge toward the center of the storage cavity (ibid.) and, where no spent fuel assembly has been stored in the cell, deform the wall beyond the elastic limit of Type 304 stainless steel (Tr. 606).
Under these circumstances, the inner wall would not return to its original shape after release of the internal pressure through venting (Tr. 607) so the cell would not be used for storage (Tr. 605).
However, the stainless steel would not rupture (Tr. 607).
The presence of a fuel assembly in the cell would prevent this deformation by obstructing the inner wall's movement (Tr. 606-607).
The force thus imposed on the fuel assembly, however, would not damage the spent fuel rods or their zircaloy cladding (Tr. 7 41-7 4 7 ).
- 21.
The Licensee testified that if a leak develops in a cell already containing a spent fuel assembly, semi-remote tooling will be employed to drill vent holes, release the hydrogen, and prevent subsequent accumulation of gas, which is
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similar to the practice followed at Monticello (Exhibit 2 at 5, Tr. 440).
Licensee and NRC witnesses agreed that the amount of hydrogen released would be too small to pose any risk of combustion or explosion. (Tr. 595, 596, 611, 612, 691, 692).
This venting procedure would be required only for those cases in which the fuel assembly could not be withdrawn from the cell by using a force within the allowable limits of the fuel-handling crane (Exhibit 2 at 5).
Venting would release the pressure and permit routine removal of the assembly (ibid.)~ Because the assembly would prevent the inner cell walls from deforming past their elastic limit, the walls would return to their original shape.
(Tr. 605-606).
Thereafter, the cell would be available for further use (ibid.).
- 22.
Although one could avoid any possibility of swelling by venting all of the cells before installation, the witnesses for the Licensee disagreed with those for the Staff as to whether such a step would be advisable.
The Licensee's testimony was that it would be preferable to maintain water-tight integrity of the cells, so as to separate the Boral from pool water, as an added line of defense against possibly unknown corrosion effects.
In return,* the Licensee would be willing to accept the potential loss of some storage cells through swelling, and the possible burden of venting still other cells, should leaks
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develop (Tr. 619-631, :. 762-764).
The NRC Staff testified that it would be more advisable to vent all of the cells initially to eliminate the risk of swelling because the disadvantages caused by swelling may exceed those of unknown corrosion effects.
(Tr. 696-697, 708-727).
However, the Staff testified further that venting before installation should not be required to insure safety, because the Licensee's capability for venting after swelling was adequate to deal safely with any condition which swelling might create.
(Tr. 711-714, 730-734).
The Staff concluded that swelling would be a disadvantage only from an operational, and not a safety, point of view (ibid.).
Qualification and Testing of Boral
- 23.
Contention 6 asserts that inadequate consideration has been given to the qualification and testing of Boral in the spent fuel pool environment.
As indicated above, both the Staff and Licensee presented extensive testimony showing how Boral behaves in contact with the water from the spent fuel pool, and describing the testing procedures which were used to arrive at the data which was furnished.
This testimony was not contradicted by any other testimony, nor seriously weakened by cross-examination.
- 24.
In addition, the Licensee has committed itself to execute a long-term surveillance program involving the use of coupons to simulate spent fuel storage cells in the spent fuel pool (Exhibit 2 at 6, Tr. 497-499, 583-588).
The coupons will be examined one year after rack installation and every two years thereafter.
The Staff testified that this surveillance program is adequate to detect any degradation of the storage cells.
(Tr. 683-685, 694-695).
Conclusions on Contentions 2 and 6
- 25.
The Board finds, after evaluating the evidence above, that Contentions 2 and 6 are without factual merit.
The evidence establishes that neither the rack structure nor the Boral will deteriorate, that accidental criticality will not occur, and that adequate consideration has in fact been given to the possibility of such occurrences.
Also, we find that adequate consideration has been given to qualification and testing of the Boral to insure its continued integrity and ability to control r~activity in the Salem spent fuel pool environment.
Testimony presented by Licensee and Staff proved that the Licensee's ability to vent any storage cell with spent fuel in it is adequate to protect the public health and safety even if a leak should develop in such a cell.
The Board finds that the above evidence presented by the Licensee and Staff was con-vincing, and finds that the conclusions reached by their witnesses, and tested through ~xamination by the parties and the Board, are sound.
No direct testimony was introduced by the intervenors to refute any of the evidence, interpretations, or conclusions presented by the Licensee or the Staff.
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- 26.
Accordingly, the Board finds that no basis has been established for the allegation that inadequate consideration has been given to deterioration of the rack structure, or the Boral plates to be used as neutron absorbers.
We find that,*
with respect to the issues raised by Colemans' Contentions 2 and 6, the spent fuel pool can be modified and operated as proposed without endangering_ the heal th and safety of the public.
B.
Lower Alloways Creek Township's Contention No. 1 Lower Alloways Creek Township Contention No. 1 states:
The Licensee has not considered in sufficient detail possible alternatives to the proposed expansion of the spent fuel pool.
Specifically, the Licensee has not established that spent fuel cannot be stored at another reactor site.
- Also, while the GESMO proceedings have been terminated, it is not clear that the spent fuel could not by some arrangement with Allied Cnemical Corp. be stored at the AGNS Plant in Barnwell, South Carolina.
Furthermore, the Licensee has not explored nor exhausted the possibilities for disposing of the spent fuel outside of the U.S.A.
- 27.
The Environmental Impact Appraisal (EIA) prepared by the NRC Staff (Exhibit 6-C, dated January 15, 1979) states tha_t the proposed increase in storage of spent fuel will cause no significant environmental impact.
The impact, according to the Staff's testimony, will not be significantly greater than the impact originally described by the Final Environmental Statement for Salem 1 filed in April of 1973 (Exhibit 6-C at 27).
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If we accept as true the Staff's conclusion that the proposed increase in storage would have no significant impact, it follows that any alternative to the increase would have either a greater impact or one which is also insignificant.
This was pointe~ out in a recent decision by the Atomic Safety and Licensing Appeal Board in Portland General Electric Co. (Trojan Nuclear Plant)
ALAB-531, 9 NRC 263, 266 (March 21, 1979).
Under the holding of that case, there would be no need to consider alternatives if we accept the Staff's conclusion.
- 28.
In this proceeding, however, our decision to consider Contention 1 preceded the Appeal Board's decision in Trojan (Memorandum and Order of April 26, 1978), and so did the Staff's decision to consider alternatives in the EIA (Exhibit 6-C, supra, dated January 15, 1979).
In the EIA the Staff considered the alternatives of:
(1) reprocessing spent fuel, (2) storing spent fuel at an independent spent fuel storage installation (ISFSI), (3) storing spent fuel from Salem Unit 1 in the pool at Salem Unit 2, (4) storing spent fuel from Salem at some other reactor site, and (5) shutting down the Salem Unit 1 facility when the racks in the existing pool have been filled (Exhibit 6-C at 12-19).
Also, the Staff subsequently testified on the feasibility of storing spent fuel outside of the United States (Affidavit of Gary G. Zech, following Tr. 999).
The alternatives of reprocessing spent fuel or shutting down the facility were not an explicit part of Contention 1.
The Staff concluded in the EIA that reprocessing
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was not feasible because of the President's policy against it, and that the cost of shutting.down the reactor would be far greater than any resulting benefit.
These conclusions were fully supported by the Staff's testimony and were not effectively challenged.
Storage at an Independent Spent Fuel Storage Installation (ISFSI)
- 29.
With respect to storage space which may now exist at independent storage installations, such as that at Barnwell, South Carolina, the Staff testified that such space was not available to Salem (Exhibit 6-C at 14-15).
The Licensee's testimony concurred (Exhibit 2 at 10-11).
This testimony was not contradicted.
- 30.
With respect to storage in newly-contracted ISFSI's, the Township presented testimony by Dr. George Luchak, Professor of Civil Engineering at Princeton University (Testimony of Intervenor Township of Lower Alloways Creek in Respect to Contention No. 1, by George Luchak, Ph.D, following Tr. 918).
Dr. Luchak testified that the Licensee had not indicated to what extent it had pursued the alternative of constructing an ISFSI in concert with other electric utilities.
Also, he testified that a reasonable location for an ISFSI would be in a dry, unpopulated area, such as a desert, and that the Staff's EIA was deficient in failing to provide data on the cost and feasibility of such an ISFSI (ibid.).
He stated that the
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proposed increase in storage at Salem would produce a larger inventory of long-lived radioactivity in the pool (Tr. 952) and that in the event of a severe reactor accident with loss of the pool water, this. higher inventory could cause larger release of radioactivity toward the cities near Salem (Tr. 952-953).
- 31.
Mr. Gary Zech, who testified on behalf of the NRC Staff, stated that the accident postulated by Dr. Luchak was not credible (Tr. 1042-1043).
Mr. Zech described the pool's reinforced concrete construction, which is seismic category 1, as providing a very stable environment for spent fuel (Tr. 1047).
He testified that there is no credible method for loss of water from the pool except possibly through slow evaporation (ibid.),
that there are several sources of back-up water available at
~he pool, and that no credible accident could prevent maintaining water in the pool and cooling the spent fuel (Tr. 1047-1048).
Mr. Zech also stated that increasing the storage capacity does not increase the short-lived radioactivity in the pool because the additional fuel is older fuel and the time elapsed since its removal from the reactor is adequate for decay of the fission products which are in gaseous form (Tr. 1050-1051).
- 32.
We investigated, pursuant to a question of our own, the extent to which the consequences of a gross loss of water in the spent fuel pool would be affected by the proposed
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increase in the quantity of older fuel retained in the pool (see Part II(D), infra, "Board Question Conc~rning a Gross Loss of Water Accident").
We concluded, first, that neither we nor the witnesses who appeared were able to postulate a credible mechanism for a gross loss of water.
Second, we concluded that even if a gross loss of water should occur, there would not be a great difference between the consequences occasioned by the proposed storage configuration and those occasioned by the present one.
These conclusions were not affected by the testimony of Dr. Luchak who, we ruled, was not qualified to testify as to the probability of an accident in the Salem spent fuel pool or as to the consequences of such an accident (Tr. 913).
- 33.
Dr. Luchak was unable to state when an ISFSI might be available.
He relied upon published figures for his opinion that construction would require five years (Tr. 980).
He did not know of any pending applications for permits to construct ISFSI's (Tr. 981).
The Staff testified that the availability of ISFSI's is uncertain for many reasons (Tr. 1005-1007),
and that published estimates of the time needed to construct ISFSI's have been based on the assumption that licensing and environmental problems could be resolved expeditiously.
. (Tr. 1005-1006).
The Staff concluded, after surveying the
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existing information available to it, that an ISFSI would not be available before 1985 at the earliest (Exhibit 6-C at 16).
The Staff also testified that the environmental impacts associated with either storing the* additional spent fuel at Salem or shipping it to another location would probably play no part in the decision to use an ISFSI because both are very small (Tr. 1053),
and that storage at an ISFSI probably would not have a smaller impact on the environment than the proposed increase in storage capacity (Exhibit 6-C at 16).
- 34.
The Staff estimated the cost of constructing an ISFSI at between $24 and $54 million, and the cost of the proposed increase in storage capacity at about $3 million (Exhibit 6-C at 13, 15-16).
The Licensee did not carry out an independent cost analysis for constructing an ISFSI, either alone or with other utilities (Tr. 780-781, 798, 1009-1010) but, based on studies by others, concluded that it would be extremely costly (Tr. 833-835).
Dr. Luchak did not challenge the Staff's cost estimate for constructing an ISFSI, but he did testify that the increase in storage capacity at Salem would result in higher costs for safeguards, security, and maintenance at Salem (Testimony of Intervenor Township, supra, at 3).
He declined to estimate those costs in dollars (Tr. 970).
The Licensee's testimony, in response, was to the effect that expenses for safeguards, security, and maintenance would be far higher at
-*22 -
an ISFSI than at Salem because the ISFSI would be an entirely new facility (Tr. 835).
- 35.
After a review of the above testimony, we conclude that construction and use of an ISFSI would be more costly than the proposed expansion at Salem, that it would produce environ-mental impacts as great or greater than the proposed expansion, that it would not reduce appreciably the risk or consequences of a gross loss of water in the spent fuel pool, and that it is unknown whether an ISFSI can or will be constructed in time to be available for storage of spent fuel from Salem Unit 1 when that storage is needed.
We find that in view of these conclusions., the Licensee has considered this alternative in sufficient detail.
Storage of Spent Fuel from Salem 1 in the Pool of Salem 2
- 36.
The spent fuel storage pool at Salem Unit 2 has been fully constructed (Tr. 811).
The testimony of the Licensee and Staff agreed that, with the existing racks and under the best current estimates, the pool of Unit 1 would be full after 1983 and the pool at Unit 2 would be full after the refueling outage in 1984 (Tr. 1026-1027, 1030, 1104-1105).
Without arrangements for additional storage, it will.be necessary for both Units to shut down.
With the proposed higher density racks, the pool of Unit 1 would be full in 1999 and the pool of Unit 2 in 2000 (Tr. 1105).
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- 37.
If, after filling the present racks in the pool at Salem 1, spent fuel from Salem 1 were stored in the existing racks of the pool at Salem 2, the pool at Salem 2 would be filled by 1983 (Tr. 820.).
However, because the Licensee a ls*o plans to install higher density racks at Salem 2 (ibid.) it would be possible to store spent fuel from Salem 1 in the new racks at Salem 2 in such a fashion as to extend the storage capability of the combined facilities to 1991 or 1992, without installing new racks at Salem 1 (Tr. 1135).
The Staff testified that this extension of storage would still not be long enough, however, to assure that the Licensee could obtain storage else-where (Tr. 1137-1138).
In addition, the Staff testified that storing spent fuel from Salem 1 in the pool at Salem 2 would produce higher occupational exposures than the proposed re-racking of Salem 1 because the spent fuel from Unit 1 would have to be loaded and transported to Unit 2, the spent fuel pools being in separate buildings (Exhibit 6-C at 17).
This would require a license amendment (Tr. 1147).
Further, if the pool at Unit 2 is filled with fuel from both Units before offsite storage is available, it would then become necessary to rerack the Unit 1 pool at a time when it had been filled with spent fuel, which would produce a higher occupational exposure than would reracking it now, with only one annual discharge present, as proposed (Tr. 1144-1145),
Finally, if such a.
belated reracking of Unit 1 became necessary, *it would also be necessary to load and transport the spent fuel from Unit 2 back to Unit 1 in order to use Unit l's increased capacity, and
24 -
this would increase occupational exposures even more (Tr. 1151).
In view of the uncertainty of off-site storage even by 1991 or 1992, and these higher exposures, we conclude that the alterna-tive of storing spent fuel from Salem 1 at Salem 2 is not prefer-able to the proposed increase in storage at Salem 1.
Offsite Storage at Other Reactors
- 38.
Hope Creek Units 1 and 2, which are currently under construction near Salem, are the only other nuclear facilities owned by the Licensee; *these Uni ts probably wi 11 not be com-pleted before the existing pools are full at both Salem Units (Exhibit 6-C at 17-18).
Hope Creek will use boiling water reactors which have fuel assemblies with dimensions different from the assemblies of the pressurized water reactors used at Salem (ibid.).
To use the Hope Creek pools for Salem's fuel, it would be necessary to replace their racks (id.).
To do so would reduce the storage capacity available for the Hope Creek reactors (id.).
The Staff cited a survey conducted by the Energy Research and Development Agency to the effect that up to 46% of the operating nuclear power reactors in the United States will lose the ability tq refuel.during.the period from 1975.to 1984 unless storage capacity is increased in spent fuel pools or off-site storage is found (id.).
The Staff concluded that under these circumstances, the Licensee could not prudently rely upon the Hope Creek units or any other power facility to provide
25 -
additional storage when the Salem pool is filled.
Since no testimony to the contrary was offered, we must agree with the Staff's conclusions.
Storage Outside the United States
- 39.
The Staff testified that the possibility of disposing of spent fuel from Salem Unit No. 1 outside the United States is nonexistent because of the federal government's policy re-garding nonproliferation of nuclear weapons.
A large scale shipment of spent fuel to a foreign country by a utility company would not be permitted if that country engages in reprocessing (Affidavit of Gary G. Zech at 2, following Tr. 999).
This alternative was not discussed further in the proceeding by any of the parties.
The Board concurs that this is not a viable alternative to the proposed modification.
Conclusions on Contention 1
- 40.
After evaluating the evidence before us, we conclude that the proposed increase in spent fuel storage capacity at Salem Unit 1 will not significantly increase the impact on the human environment caused by the Salem 1 reactor.
We also con-clude that storage outside the United States or at an existing independent spent fuel storage installation is not available, that construction of such an installation by the Licensee would not be an alternative preferable-to the proposed increase, and neither would storage of spent fuel from Salem 1 at Salem 2 or
26 -
at another reactor, if another reactor were available.
In short, we find that Contention 1 has no factual merit in light of the evidence received, and that alternatives to the proposed action have been adequately considered.
C.
Board Questions Concerning the Accident at Three Mile Island
- 41. The accident at the Three Mile Island Nuclear Station, Unit 2, occurred-while this proceeding was pending.
- Because of the serious safety questions which this accident raised, we sought to determine what the effects would be on Salem's spent fuel pool if an accident similar to that at Thre8 Mile Island were to occur at Salem.
We asked the parties to respond to the following questions:
(a)
To what extent did the accident at Three Mile Island affect the spent fuel pool at that site?
(b)
If an accident such as the one at Three Mile Island occurred at Salem, to what extent would the accident affect the spent fuel pool?
To what extent would it have mattered how much spent fuel was present at the pool at Salem?
- 42.
Both the Staff and the Licensee introduced evidence on these questions.
The Staff's testimony was given by Mr. Gary G. Zech and Dr. Jack N. Donohew, Jr.
The Staff also offered Exhibit 12, a package of view plans showing radiation
27 -
fields in the auxiliary building at Three Mile Island.
The Licensee's testimony was given by Messrs. Robert P. Douglas, Edward A. Liden, and Robert A. Burricelli.
No other party presented evidence on these questions; however, all parties and the Board examined the witnesses.
- 43.
With respect to question (a), the Staff's witnesses testified that there was no spent fuel in the pool at Three Mile Island when the accident occurred, that even if there had been the accident would not have affected it (NRC Staff Response, In Part, to Board Questions, following Tr. 1133), and that the pool itself re~ained accessible despite levels of radiation which were higher than normal (Tr. 1236).
The Staff also stated that the equipment for cooling the spent fuel pool and purifying its water was accessible at Three Mile Island after
-~he accident (Tr. 1233~34). The Licensee's witnesses were in accord (Tr. 1291-92) although they were unable to state what the level of radiation actually was in the areas where the equipment was located (Tr. 1293).
This uncertainty was cured by additional testimony from the Staff, which showed that the Staff arrived at its conclusions on accessibility by evaluating actual measurements of radiation fields at Three Mile Island shortly after the accident (Tr. 1324-1339; Exhibit 12).
- 44.
With respect to question (b), the Staff testified that at Three Mile Island radioactive water was pumped automatically from the containment building, which houses the reactor, to tanks in the auxiliary building, which houses the spent fuel pool and the equipment needed to operate it.
This water over-flowed from the tanks into the auxiliary building, producing high levels of* radioactivity there (NRC Staff Response, supra.). The Staff's testimony established that such an automatic transfer could not occur at Salem because valves in the lines leading from the containment to the auxiliary building at Salem close automatically on the safeguards signal (id. at 3).
At Three Mile Island these valves were not designed to close (and did not close) upon the safeguards signal.
The Staff further testified that even if these va~ves (which isolate the containment) did not work properly at Salem, and there were an inadvertent transfer of contaminated water to Salem's auxiliary building, the operation of the spent fuel pool and its support systems would not be seriously affected (id.).
At Salem some of the support systems for the spent fuel pool (the cooling and purification systems) are located in the auxiliary building, which at Salem is a building separate from the building housing the spent fuel pool itself. Parts of the auxiliary building would be contaminated by radioactive water if Salem experienced an accident similar to that at Three Mile Island.
- However, at Salem the parts of the auxiliary building which would.be contaminated are shielded from the support systems for the spent fuel pool to such an extent that the support systems
29 -
would remain accessible for purposes of maintenance (id. at 5).
This conclusion, made by the Staff, was based upon a comparison of the design of Salem's auxiliary building with that of Three Mile Island and, more specifically, on a study of the specific location of the spent fuel pool support systems at Salem, the location of areas which could be contaminated at Salem by an accident similar to that at Three Mile Island, and the radiation fields which existed in the contaminated portions of the auxiliary building at Three Mile Island after the accident (Tr. 1179; 1181).
The Staff concluded that at Salem there would be higher dose rates than normal in the vicinity of the purifica-tion system but that the effect would-not be serious (Tr. 1169).
One source of make-up water for Salem's spent fuel pool could be restricted because of the expected contamination of the noldup tank for the chemical volume and control system, but other available sources would still exist (Tr. 1207).
In addi_tion, valves to provide make-up water to the spent fuel pool at Salem are located in the fuel handling building, so that make-up water would be available without ever having persons enter the auxiliary building (Tr. 1240).
- 45.
The Licensee presented the testimony of Messrs. Liden, Douglas, and Burricelli on these same questions (Licensee's Response to Licensing Board's Question 1 and Part 1 of Question 3 Relating to Impact of a Three Mile Island Type Incident on the Salem Unit 1 Spent Fuel Pool, following Tr. 1264).
These
'i 30 -
witnesses agreed with the Staff that both the spent fuel pool and its support systems would be accessible for maintenance if Salem experienced an accident in which contaminated water were transferred to the auxiliary building (id. at 3).
In addition, these witnesses testified that the ventilation system in Salem's auxiliary building is designed to. prevent the movement of airborne radioactivity from one area which might be contaminated to another (id.).
This system, which is typical of most nuclear plants, brings air in through clean areas such as corridors and exhausts it through areas which might be contaminated (Tr. 1280).
Thus, according to the Licensee, gaseous radioactivity in Salem's auxiliary building is not expected to contaminate areas containing the support equipment for the spent fuel pool (Licensee's Response, supra, at 3).
- This ventilation equipment, which is operated remotely from the control room, is located in a space between the contain-ment building and the fuel handling building (Tr. 1287).
The equipment would be accessible for maintenance even in the event of an accident of the type at Three Mile Island, especially if temporary shielding were used to reduce radiation levels (Licensee's Response, supra~ at 4).
- 46.
In sum, the testimony by both the Licensee and Staff showed that the type of accident which occurred at Three Mile Island would not seriously affect the spent fuel pool at Salem or any of its supporting equipment.
The Board finds this testimony
' to be convincing.
No other party offered testimony on this point, nor did any other party propose a finding of fact or conclusion of law on this point.
Also, there was no sug-gestion that any effect on the spent fuel pool which might occur from such an accident would depend on whether the pool contained the additional spent fuel assemblies sought to be authorized by this application.
We find that if an accident of the type which occurred at Three Mile Island were to occur at Salem, there would be little, if any, effect on the spent fuel pool as now authorized, and little, if any, effect on the pool with the expanded storage capacity requested by the Licensee.
We consider our questions concerning Three Mile Island to have been adequately answered.
D.
Board Question Concerning a Gross Loss of Water Accident
- 47.
In our Memorandum and Order of Febnuary 22, 1980 (LBP-80-10, 11 NRC 337, 346) we directed the parties to answer the following question:
In the event of a gross loss of water from the storage pool, what would be the difference in consequences between those occasioned by the pool with the expanded storage and those occasioned by the present pool?
- 48.
As we explained in that Memorandum and Order, our review of Commission policy on dealing with large accidents had led us to conclude that that policy, as it then stood, would require us to have on record further evidence of the consequences of such an accident in order that we might decide, as a threshold matter, whether the change induced in those con-sequences by the fuel pool expansion would require further evaluation in the form of an environmental impact -statement (11 NRC at 346).
To receive this evidence, we held evidentiary hearings in April, 1980.
- 49.
Testimony was proposed by the Licensee, the Staff, and Lower Alloways Creek Township.
We rejected the Licensee's testimony as not responsive to our question (Tr. 1376).
We rejected the portion of the Township's proposed testimony prepared by Dr. Fankhauser as not sufficiently connected with the difference between the present and proposed storage con-figuration (ibid.).
For reasons stated on the record, we struck substantial portions of the Township's proposed testimony prepared by Dr. Webb (Tr. 1377-81 and 1679-82).
We received the Staff's proposed testimony into evidence (Tr. 1387).
- 50.
Dr. Richard C. Webb sponsored the Township's testimony concerning the consequences of a gross loss of water from the spent fuel pool (Testimony of Richard C. Webb, Ph.D., in Respect to Board Question #3, following Tr. 1697).
Dr. Webb stated that if such a loss occurred the radioactivity in the pool could be dispersed over a very large area, such as the eastern seaboard of the United States (id. at p. 14).
We asked Dr. Webb to
~
explain the mechanism by which this dispersion could occur and the factual assumptions, if any, upon which his statement depended.
Dr. Webb responded that he had reached his conclusion
33 -
by calculations which simply assumed that a large amount of radioactivity would escape from the pool (Tr. 1699-1702).
We then asked Dr. Webb whether he could identify any mechanism by which this assumed* release of radioactivity could or would occur.
He responded (Tr. 1706-1709) that this mechanism was described in Part II of his testimony of April 9, 1980 entitled "Consequences of Zirconium Fire:
Fission Product Release (Postscript)" (following Tr. 1697).
When asked to describe or state more specifically what mechanism he was referring to, he was unable to do so; he responded simply that there were "many factorsn (Tr. 1708) and that the description of the mechanism was spread generally through the pages of Part II of his testi-mony (Tr. 1709).
We also asked Dr. Webb to indicate to what extent this mechanism might depend upon or be influenced by the presence of spent fuel four years old or older (spent fuel in this category is the subject of the Licensee's application).
Dr. Webb pointed to a statement in his testimony--made without any supporting analysis or data--that a zirconium fire "could conceivably spread to old spent fuel" [from fresher fuel], to another statement in his testimony that "one must assume that
- [a zirconium] fire wi 11 spread" (Tr. 1710-1711), and to an addendum in which he said he addressed a zirconium fire which had occurred at Bettis Laboratories (Tr. 1711).
The
34 -
first two statements are simply unsupported assertions, and Dr. Webbvs testimony in the cited addendum does not in fact discuss the extent to which any mechanism for release of radiation is influenced by the presence of spent fuel four years old or older.
It seems clear to the Board that Dr. Webb believes a zirconium fire could start in the pool if a gross loss of water should occur and that the fire could spread to fuel four years old or older.
However, there is nothing specific to show whether such a fire could propagate or what specific difference the densification of storage would make.
When Dr. Webb attempted to analyze this question for the existing (open) racks he could only conclude:
"Intensive efforts on my part failed to solve this formidable mathematic problem."
(Part III of Supplement of April 8-9, 1980 to February 27, 1979 testimony, at p.l).
- 51.
With respect to the probability of propagation, Dr. Webb did not offer any firm opinion.
He stated:
"I considered the likelihood (probability) and I concluded that any judgment on likelihood is un-scientific and pure speculation.
So I considered the matter and disposed of it that way."
(Tr. 1731).
He asserted in response to a Board question that he simply disagreed with the use of "probability" in making these cal-culations (Tr. 1732).
- 52.
In the matter of dispersion of fission products once released from the pool, Dr. Webb's conclusions concerning the impacts of fission product releases depend upon a knowledge of meteorology.
However, Dr. Webb is not a meteorologist (Tr.
1687), nor has he published any articles in meteorological journals (Tr. 1688).
He testified that he had studied meteorology (Tr. 1697).
- 53.
When one views Dr.
Webb's testimony as a whole, it*
is impossible to glean from it any clear picture either of a mechanism by which a large amount of radioactivity could escape from the pool, or the assumptions of fact which might be appropriate to such a mechanism.
Neither does the testimony discuss any clear relation between such a mechanism and the presence of spent fuel four years old or older in the pool.
- 54.
In general, we found much of the testimony ill-organized and difficult to follow.
It was unsuitable for assessing the probability that a serious accident could be caused by a substantial loss of water.
It was of even less help in trying to determine whether the total risk presented by the fuel pool would substantially increase because of the proposed expansion.
- 55.
The Staff's testimony was more productive.
It was sponsored by Mr. Walter F. Pasedag, Environmental Evaluation Branch, Division of Operating Reactors, U.S. Nuclear Regulatory Commission.
His professional qualifications appear at Tr. 1387.
Dr. Allan S. Benjamin of Sandia Laborato~ies joined Mr. Pasedag in presenting the Staff's case.
Dr. Benjamin's
36 -
qualifications appear at Tr. 1389.
Dr. Benjamin, one of the authors of what became known in the hearing as the "Sandia Report", acted as a consultant to Mr. Pasedag (Tr. 1390).
The report (NUREG/CR-0649) is entitled "Spent Fuel Heatup Following Loss of Water During Storage" (Tr. 1399-1400).
- 56.
Mr. Pasedag testified that, for fresh spent fuel, continued denial of water cooling in the spent fuel pool could lead to oxidation and failure of the clad, to overheating of the U02 fuel, and possibly to the release of fission products into either the present or the proposed pool.
(Direct Testimony of Walter F. Pasedag in Response to Board Question No. 5, following Tr. 1387, at p.4),
Also, in the new, denser storage configuration proposed by the Licensee, there would be less natural convection after a loss of water than there would be with the present, more widely spaced configuration.
- Thus, there would be a higher likelihood that the recently discharged fuel would reach oxidation temperatures (with possible clad melting) in the proposed configuration.
(Further Testimony of Walter F. Pasedag in Response to Board Question No. 5, following Tr. 1387, at p. 2).
The decay time required to assure that the fuel's decay heat generation would not result in oxidation temperature~ (above 900°c) in the higher density storage con-figuration is about one year (id. at p. l; Tr. 1441).
- 57.
We pursued with Dr. Benjamin the notion of a possible fire in the pool after a gross loss of water.
It appeared that
37 -
he was the person upon whom the Staff relied most specifically in matters of heat transfer and oxidation (Pasedag Further Testimony, supra, at p. l; Tr. 1390).
Dr. Benjamin was familiar with the analytic techniques utilized in the heatup analysis*
contained in the Sandia Report:
In general, we found his testimony to be cogent and well founded.
According to Dr. Benjamin, it would not be possible to have "flames" in the spent fuel pool despite the high temperatures which would follow a gross loss of cooling water (Tr. 1393).
The freshly discharged spent fuel would become hot, oxidize, glow, and emit heat by thermal radiation (Tr. 1393, 1394).
The oxidation would not, however, spread from one spent fuel element to another by what is commonly thought of as a "fire"--a de-flagration with rapid convection and spreading of flames; it could propagate only through a process in which heat radiating from the recently discharged spent fuel might raise the temperature of older spent fuel assemblies which had been stored nearby (Tr. 1391, 1392).
Dr. Benjamin did not believe that one could rule out the possibility that this rise in temperature could cause these older assemblies to oxidize (Tr. 1392, 1399).
Mr. Pasedag, however, testified that any oxidation of these older assemblies would be limited, and would not lead to a release of fission products substantially greater
38 -
than those released by the recently discharged fuel (Pasedag, Further Testimony, supra, at p. 2).
Both witnesses stated that the calculations required to form a solid conclusion on the propagation of oxidation were beyond the scope of the Staff's review of the application (Tr. 1391, 1418).
According to Mr. Pasedag, one can nevertheless be confident that even if some oxidation of the older spent fuel assemblies occurred, the oxidation would be limited to those stored nearest the recently discharged assemblies, would probably not be sufficient to melt the clad, and would certainly not be sufficient to melt the fuel (Tr. 1448):
If the clad were indeed to melt on the older assemblies ( 4 years old or older) the radioactive release would be limited to the fission products contained in the gap between the clad and the fuel pellets (Tr. 1449).
There would be essentially none of the more volatile fission products left in the gap of the older fuel--they would have decayed or plated out (ibid.).
Altogether, the possibility of radioactive release from the older spent fuel is limited to a few isotopes and would be small even in the event thermal radiation should cause some of that fuel to oxidize (id.).
- 58.
We find the above testimony by the Staff to be persuasive and not meaningfully contradicted by any other testimony.
On cross-examination Dr. Benjamin stated that it
39 -
would be possible through further analysis to predict more precisely whether oxidation could propagate to the older fuel, and that the calculations for such an analysis could be performed by one person in a few months (Tr. 1483).
We do not believe*, how-ever, that further study is needed to reach our decision.
Mr. Pasedag's testimony convinced us that even if oxidation did propagate to the older fuel the resulting radioactive release would not be significant in comparison to the radioactive release from the recently discharged fuel.
When we consider that Dr. Webb was unable to describe any credible mechanism. for propagation despite a specific invitation to do so, and consider that a gross loss of water is in itself an event of very low probability, we do not believe that further study of propagation is necessary to answer our question.
We are satisfied that in the event of a gross loss of water from the spent.fuel pool, there would not be a great difference between the consequences occasioned by the proposed storage configuration and those occasioned by the present one.
E.
Class 9 Accidents
- 59.
In the course of the hearing we asked the following question of the parties:
The proposed Annex to Appendix D, 10 CFR Part 50, appears to define a Class 9 accident as a sequence of failures which are more severe than those which the safety features of the plant are designed to prevent.
The sequence of failures at Three Mile Island produced a breach of the containment and a
40 -
release of radiation which could not be prevented by the safety features.
Was the
- occurrence at Three Mile Island therefore a Class 9 accident?
Was the risk to health and safety and the environment "remote in probability," or "extremely low" at Three Mile Island, as those terms are used in the Annex?
(Tr. 922-23).
- 60.
When we asked this question, we were of the opinion that if in fact a "Class 9" accident had occurred after only a few hundred reactor-years of operation, this fact would be important in interpreting the meaning and scope of the policy contained in the proposed Annex.
The Annex assumed that Class 9 accidents, although severe, were so "remote in pro-bability" that their environmental risk was "extremely low".
- 61.
All parties replied to this question, the Intervenors generally taking the position that the accident was Class 9, the Licensee taking the position that it was not, and the Staff taking the position that it was, but notifying us of a Staff minority opinion that it was not.
- 62.
In the time since we asked the question and received the replies, the Commission has revised its policy on review of Class 9 accidents.
On June 9, 1980 the Commission published a statement of interim policy entitled "Nuclear Power Plant Accident Considerations Under the National Environmental Policy Act of 1969", 45 Fed. Reg. 40101 (June 13, 1980).
In that
41 -
statement the Commission withdrew the Annex and abolished the system of accident classification.
The effect of this change is to moot our question, since no problem of applying or interpreting the proposed Annex can now arise.
Even severe accidents may now be considered in the Commission's environ-mental review.
- 63.
This new policy applies to cases in which the Commission's Staff has not completed its environmental impact statement.
Thus, it does not apply to this case.
It is worth noting, however, that our inquiry into the consequences of a gross loss of water from the spent fuel pool may have anticipated the change in policy.
Or at least, objections to inquiries such as ours will no longer be made on the ground that certain accidents belong to a forbidden category.
For the accident sequence which we investigated, we are satisfied that the record now contains an analysis which would be adequate even under the new policy.
III.
CONCLUSIONS OF LAW
- 64.
The grant of the license amendment requested in this proceeding is not a major Commission action significantly affecting the quality of the human environment.
Therefore, it does not require the preparation of an environmental impact statement under the National Environmental Policy Act of 1969 (NEPA), 42 u:s.c. §4321, et. seq., or under Part 51 of the
42 -
Commission's regulations.
The basis for this conclusion is our review of the record of this proceeding, particularly the evidence supporting the Staff's Environmental Impact Appraisal (Exhibit 6-C).
In the Appraisal, the Staff describes the environmental impact of the proposed modification, the need for the increase in storage capacity, the environmental impact of postulated accidents, alternative possibilities for spent fuel storage, and the overall balance of costs and benefits.
The evidence adduced fully supports the Staff's conclusion that this action will not significantly affect the quality of the human environment.
None of the testimony or cross-examination by intervenors or interested states showed that the Staff's conclusion was incorrect, or that the evidence supporting that conclusion was inadequate.
- 65.
As we pointed out above in our discussion of Contention 1, there is authority for the proposition that NEPA does not re-quire us to consider the possible alternatives to a proposed action if that action is not one which will significantly affect the quality of the human environment.
Portland General Electric Co. (Trojan Nuclear Plant) ALAB-531, 9 NRC 263, 266 (1979).
We have nevertheless considered the Township's Contention 1 and have found, as stated above in our Findings of Fact, that no alternative has been identified which would produce smaller
43 -
environmental impacts than the action proposed by this application,
- 66.
When the Commission published its "Intent to Prepare Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel," 40 Fed. Reg. 42801 (September 16, 1975), the Commission stated that five factors should be "applied, weighed and balanced" for particular license applications made while the Generic Statement was still under consideration.
These five factors were considered by the Staff in its Environmental Impact Appraisal, but they were not directly made matters in controversy before us, and no evidence other than the Staff's Appraisal was introduced with respect to them.
Ypon a review of the evidence presented, we conclude that the Staff's conclusion regarding these five factors is fully warranted.
Further, we note that the Staff published the Final Generic Environmental Statement on Handling and Storage of Spent Light Water Power Reactor Fuel, NUREG-0575, in August of 1979.
This final statement finds that, as a general matter, enlargements of storage capacity at spent fuel pools are economically and environmentally acceptable (NUREG-0575 at ES-10, ES-11).
- 67.
The Commission's statement of interim policy entitled "Nuclear Power Plant Accident Considerations Under the National Environmental Policy Act of 1969," 45 Fed. Reg. 40101 (June 13, 1980), which withdraws the proposed Annex to Appendix D to
4 44 -
10 CFR Part 50, does not affect this proceeding.
As pointed out above in our discussion of Class 9 accidents, this new policy does not apply to proceedings in which the Staff has completed its environmental review.
The new policy is not a*
basis for 11opening, reopening, or expanding any previous or ongoing proceeding."
43 Fed. Reg. at 40103.
- 68.
There is reasonable assurance that the activities authorized by the requested amendment to the operating license can be conducted without endangering the health and safety of the public.
- 69.
The activities authorized by the requested amendment to the operating license will be conducted in compliance with the Commission's regulations.
- 70.
The issuance of the requested amendment to the operating license will not be inimical to the common defense and security or to the health and safety of the public.
IV.
ORDER
- 71.
Wherefore, it is ORDERED, iri accordance with the Atomic Energy Act, as amended, and the regulations of the Nuclear Regulatory Commission, and based on the findings and conclusions set forth herein, that the Director of Nuclear Reactor Regulation is authorized to make appropriate findings in accordance with the Commission's regulations and to issue the appropriate license amendment authorizing the requested replacement of spent fuel storage racks at Salem Station Unit 1.
- 72.
It is further ORDERED in accordance with 10 CFR
§§ 2.760, 2.762, 2.764, 2.785, and 2.786, that this Initial Decision shall be effective immediately and shall constitute the final action of the Commission forty-five (45) days after the issuance thereof, subject to any review pursuant to the above-cited Rules of Practice.
- 73.
Exceptions to this Initial Decision may be filed within ten (10) days after service of this Initial Decision.
A brief in support of the exceptions shall be filed within thirty (30) days thereafter [forty (40) days in the case of the NRC Staff].
Within thirty (30) days of the filing and service of the brief of the Appellant [forty (40) days in the case Of the Staff] any other party may file a brief in support of, or in opposition to, the exceptions.
Dated at Bethesd~, Maryland THE ATOMIC SAFETY AND LICENSING BOARD
... /
(J Lamb, III;"Member G
ilhollin, Chairman this 27th day of October, 1980.
I EXHIBITS Relating to the Initial Decision in Public Service ElectTic and Gas Company (Salem Nuclear Generating Station, Unit 1), Docket No. 50-2 72-*SP.
Soonsor No.
Licensee II II II II II II NRC Staff II 1-A 1-B 1-C 1-D 1-E 1-F 1-G 1-H 1-I 1-J 1-K 2
3 4
s 6-A 6-B Letter of 11/18/77 from Librizzi to Lear.
Letter of 12/13/77 from Librizzi to Lear.
Letter of 2/14/78 from Librizzi to Lear.
Letter of 5/17/78 from Librizzi to Lear.
Letter of 7 /31/78 from Librizzi to Schwencer.
- Letter of 8/22/78 from Librizzi to Schwencer.
Letter of 10/13/78 from Librizzi to Schwencer.
Letter of 10/31/78 from Librizzi to Schwencer.
Letter of 11/20/78 from Librizzi to Schwencer.
Letter. qf 12/22/78 from Librizzi to Schwencer.
Letter of 1/4/79 from Librizzi to Schwencer.
Affidavit of Edwin A. Liden, 2/21/79.
Exxon Nuclear Company Document XN-NS-TP-009 entitled "Fuel Storage Racks Corrosion Program, Baral-Stainless Steel,"
11/9 /78.
Application for Withholding Information From Public Disclosure (with supporting affidavit), 1/10/79.
Exxon Nuclear Company Document XN-NS-TP-009/NP, entitled "Fuel Storage Racks Corrosion Program, Boral--Stainless Steel (Non-proprietary Version)" March, 1979.
Letter of 1/15/79 from Mr. Schwencer to Mr. Librizzi.
Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to the Modification of the Spent Fuel Storage Pool, Facility Operating License No. DPR-70, Public Service Electric
I.,,}
I.,..,
.d Gas Company, Salem Nuclear ~erating Station Unit Number 1, Docket Number 50-272.
6-C Environmental Impact Appraisal by the Office of Nuclear Reactor Regulation Relating to the Modification of the Spent Fuel Pool, Facility Operating License Number DPR-70, Construction Permit *Number CPPR-53, Public Service Electric and Gas Company, Salem Nuclear Generating Station Unit Number 1, Docket Number 50-272.
7 Corrosion of Materials in Spent Fuel Storage Pools, John R.
Weeks, July 1977 (BNL-NUREG 23021).
8 Corrosion Considerations in the Use of Boral in Spent Fuel Storage Pool Racks, J.R. Weeks, January, 1979 (BNL-NUREG 25582).
New Jersey and Delaware 9
Letter of 12/20/77 from Cunningham to Smith.
10 Letter of 1/19/78 from Crockett to Beckjord.
Licensee 11 Affidavit of Thomas Eckhart dated June 18, 1979.
NRC Staff 12 Package of view plans of the auxiliary building at Salem.
Colemans 13 NRC Inspection Report 50-263/79-02 concerning the Monticello Nuclear Generating Plant, dated April 10, 1979.
Licensee 14 Licensee's Response to Licensing Board Question 5 Regarding a Gross Loss of Water from the Salem Spent Fuel Pool, dated April 10, 1980.