ML17222A439
| ML17222A439 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/30/1988 |
| From: | Tourigny E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17222A438 | List: |
| References | |
| 88-560-01-LA, 88-560-1-LA, OLA, NUDOCS 8809020070 | |
| Download: ML17222A439 (21) | |
Text
ATTACHMENT A UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of Florida Power and Light Company Company Docket No. 50-335-OLA ASLBP No. 88-560-01-LA St. Lucie Plant, Unit No.
1 Affidavit of Edmond G. Tourigny In Support of Motion for Summary Disposition Mr.
2.
Tourigny being duly sworn according to Iaw states as follows:
Q name is Edmond G. Tourigny.
I am employed in the Office of Nuclear Reactor Regulation (NRR) of the U.S. Nuclear Regulatory Colission.
I am presently the Project Manager for Florida Power and Light Company's St.
Lucie Plant, Unit Nos.
1 and 2.
I have been the Project Manager for the St. Lucie Plant since May 1986.
A summary of my qualifications and experience is attached hereto as Exhibit A, which is incorporated herein by reference.
The responsibilities of my present position include reviewing changes to the Technical Specifications (TS) proposed by the licensee, preparing notices of Significant Hazards Consideration, preparing safety evaluations, environmental assessments (when required),
and environmental impact statements (when required) on the proposed change(s),
and if the proposed change is acceptable, preparing the license amendment and supporting documentation fcr issuance by the Project Director.
Assistance in preparing the safety evaluation and environmental assessment or environ-mertal impact statement (when required) is obtained, as appropriate, from technical specialists in NRC's Office of Nuclear Reactor Regulation.
I managed the safety and environmental review of the St. Lucie spent fuel pool rerack amendment request.
In addition, I wrote parts of the safety evaluation and environmental assessment.
Because of the complexity of review in this matter, a team of technical specialists was assembled to evaluate the licensee's submittals.
This team was augmented by contractual support from Brookhaven National Laboratory.
The team consisted of per sonne1 highly experierced in structures, geosciences physics, radiation, dose assessment, thermal-hydraulics, chemistry and materials.
The team members performed their own analyses of various aspects of the licensee's submittal.
For example, one member of the team performed irdependent calculations of radiation dose to members of the public under normal and accident conditions.
One team member performed
'I his own physics analysis.
One team member independently calculated pool heating and cooling under various scenarios.
Thus, in all significant
- matters, the team was able to evaluate on its own the acceptability of the rerack of the spent fuel pool.
I reviewed the work of every team member and agreed with their analyses and conclusions.
The results of the team's efforts were then summarized in the safety evaluation and environmental assessment which supported authorizing the licensee to rerack the spent fuel pool.
3.
The purpose of this affidavit is to respond to the contentions admitted by the Atomic Safety and Licensing Board in its April 20, 1988 Memorandum and Order, and to support Licensee's Motion of August 5, 1988 for Summary Disposition.
The admitted contentions 3, 4, 6, 8, 9, 11 and 15 were renumbered as I through 7, respectively.
As part of the discovery process, the intervenor was afforded the opportunity to respond to the licensee's interrogatories.
I have reviewed the intervenor's response and concluded that no new information has been identified, nor has the intervenor identified any witnesses.
Based upon these considerations, the staff's safety evaluation and environmenta 1 assessment continues to serve as an adequate basis to authorize the spent fuel pool rerack.
It should be noted that the rerack effort has been completed by the licensee.
4.
Admitted Contention I Admitted Contention I reads as follows:
Admitted Contention I That the ca lculation of radiological consequences resulting from a cask drop accident are not conservative, and the radiation releases in such an accident wi 11 not meet the 10 CFR Part 100 criteria.
(Origina lly Amended Petition Contention 3.)
Memorandum and Order, Appendix A. p. I, (April 20, 1988).
The bases for the contention read as follows:
Bases for Contention The study prepared by the Department of Nuclear Energy, Brookhaven National Laboratory, entitled "Severe Accidents in Spent Fuel Pools in Support of Generic Safety",
NUREG/CR-4982, BNL-NUREG-52093, indicates that, ".
. the calculation of radiological consequences resulting from such an accident are, at this point in time, apparently impossible to determine."
"There is substantial uncertainty in the fission prod'uct release estimates.
These uncertainties. are due to both uncertainty in the accident progression (fuel temperature after clad oxidation and fuel relocation occurs) and the uncertainty in fission product decontamsnation."
(S.6)
In light of such uncertainty, no estimate can be determ>ned to be conservative.
3 Amended Peition,
- p. 4.
In admitting this Contention, the Licensing Board stated that Licensee's response
. should show that its analysis of a cask drop acci-dent bounds those uncertainties that are identifsea in the BNL Report and listed as the bases for this contention.
- Thus, by such conservatisms and
- analysis, Licensee must demonstrate compliance with 10 CFR Part 100 (1987).
I Yiemorandum and Order, p.
13, (April 20, 1988).
The intervenor contends that a cask drop accident wi 11 cause radiological consequences gr eater than 10 CFR Part 100 criteria.
The Brookhaven National Laboratory (BNL) report on Severe Accidents is used as the basis for the contention.
Thc BNL report assumed a number of events'ould occur in order to postulate significant radiological releases.
Such events include failure of the pool, :hereby causing rapid loss of alI contained cooling water, eventual fuel heatup, and a subsequent zirconium c1adding fire.
I do not believe that a cask drop accident wi 11 cause a radiological release greater than 10 CFR 100 guidelines, provided that the licensee adheres to Technical Specification 3.9.13 which reads "The maximum load which may be handled by the spent fuel cask crane shall not exceed 25 tons" and Technical Specification 3.9. 14 which reads "The irradiated fuel assemblies in the fuel storage pool sha 11 have decayed for at least 1180 hours0.0137 days <br />0.328 hours <br />0.00195 weeks <br />4.4899e-4 months <br />, unless more than one-third I of the] core is placed into the pool, in which case the irradiated fuel assemblies shall have decayed for 1490 hours0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br />."
The basis for this maximum load technical specification
- reads, in part, "structural damage caused by dropping a
load in excess (sic) of a
loaded single element cask could cause leakage from the spent fuel pool in excess of the maximum makeup capability."
The 25 ton limit was placed in the technical specifications when the plant was licensed in l976 as a result of staff analysis of the spent fuel pool structure.
Thus this. limit assures that the initia 1 initiating event described in the BHL report (fai lure of the pool thereby causing rapid loss of all contained water) will not occur.
The basis for the spent fuel decay time technical specification
- reads, in part, "the minimum requirements for decay of the irradiated fuel asser:blies in the entire spent fuel storage pool prior to movement of the spent fuel cask into the fuel cask compartment ensure that sufficient time has elapsed to a liow radioactive decay of the fission products."
The decay time was reeval-uated as part of the spent fuel pool rerack request and found still to be acceptable.
Given a
1490 hour0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> decay time and a~~~ming the cask would damage all the spent fuel in the pool (an impossible event but an acceptable assumption The spent fuel cask crane cannot carry the cask over most of the spent fuel pool because the door of the fuel handling building limits cas'k entry into only a portion of the building; the cross sectional area of the cask would also be smaller than the cross sectional area of the spent fuel pool.
from a design basis perspective),
the resulting dose to an individual at the exclusion area boundary will be 21 rem to the thyroid and less than O.l rem to the whole body.
These doses are well below the 10 CFR 100 guideline doses of 300 rem to the thyroid and 25 rem to the whole body.
I have reviewed the affidavits provided on behalf of the licensee in response to this contention and I agree with the conclusions that a cask drop accident would cause radiological releases within the dose guidelines specified in 10 CFR Part 100.
In summary, a cask drop accident will not cause radiation doses to exceed 10 CFR 100 guidelines provided that the technical specifications on allowable cask loads and minimum spent fue 1 decay tsme are adhered to.
5.
Admitted Contention 2
Admitted Contention 2 reads as follows:
Admitted Contention 2
That the consequences of a cask drop accident or an accident similar in nature and effect are greatly increased due to the presence of a large crane to be built inside the spent fuel pool building in order to faci litate the reracking.
(Origina lly Amended Petition Contention 4).
The licensing Board issued an Order on July 27, 1988, dismissing Admitted Contention 2" with prejudice as. moot."
6.
Admitted Contention 3
Admitted Contention 3 reads as follows:
Admitted Contention 3
The licensee and staff have not adequately considered or analyzed materials deterioration or failure in materials integrity resulting from the increased generation of heat and radioactivity as a result of increased capacity in the spent fuel pool during the storage period authorized by the license amendment.
(Originally Amended Petition Contention 6).
Memorandum and Order, Appendix A, p.
1, (April 20, 1988).
The bases for the contention state:
Bases for Contention The spent fuel pool facility at the St. Lucie plant, Unit Ho. 1, was originally designed to store a lesser amount of fuel for a short period of time.
Some of the problems that have not been analyzed properly:
a)
Deterioration of fuel cladding as a result of increased exposure and decay heat and radiation levels during extended periods of pool storage.
b)
Loss of materials integrity of storage rack and pool liner as a result of exposure to higher levels of radiation over longer periods.
c)
Deterioration of concrete pool structure as as result of exposure to increased heat over extended periods of time.
Amended Petition, pp. 5-6.
Of this contention, the Licensing Board said:
Petitioner argues that the pool was designed to store lesser quantities of spent fuel for a shorter period of time and that licensee has failed to adequately analyze problems that may result from exposure to the increased amount of decay heat and radiation emitted by the larger number of spent fuel assemblies stored.
Petitioner specifies three problems:
(I) deterioration of fuel cladding; (2) loss of integrity of materials making up the storage rack and the pool liner; and (3) deterioration of the concrete of which the pool is constructed.
Amended Petition, 5-6.
At oral argument, Petitioner asserted that the norma 1 temperature of the pool would be increased, subjecting the pool materials, particularly the concrete, to greater stress.
Petitioner asserted that the calculation of these factors were "clearly inadequate."
Yemorandum and Order, page 17 (April 20, 1988).
The Licensing Board limited the scope nf the contention to the length of time authorized by the licensing amendment at issue.
The intervenor believes that the long-term performance of the materials associated with the spent fuel pool has not been demonstrated.
In the basis statement, the fuel cladding, storage
~acks, pool liner, and pool structure concrete were specifically noted.
The fuel cladding in the St. Lucie case is zirca loy and the r ack and liner material is stainless steel.
I believe that the successful long-term performance of spent fuel pool materials has been demonstrated as reported in a number of Commission documents, partic-ularly in the Final Generic Environmental Impact Statement on Handling and Storage of Spent Fuel Light Water Power Reactor
- Fuel, NUREG-0575, August 1979*
- This was a major source document used in the preparation of the staff's Environmental Assessment in support of the Amendment Request.
and the Waste Confidence Decision and Requirements for License Actions Regarding the Disposition of Spent Fuel Upon Expiration of Reactor Operating Licenses, Federal Re ister Notice, Volume 43, No. 207, Friday, August 31, 1984, beginning at.~48.
n a dstaon,the staff specifically focused its attention on material compatability and chemical stability of the stainless steel racks and Boraflex in its Safety Evaluation in support of the amendment.
The following statement was contained in HUREG-0575, Appendix H, Section 3.0 entitled Cladding Stability During Storage of Spent Fuel:
"Fuel handling experience in the U.S., going back to 1959, has not revealed any instance where zircaloy clad uranium oxide fuel has undergone observable corrosion or other chemical degradation during pool storage."
The following statements were contained in the staff's safety evaluation in Section
- 3. entitled "Material Compatabi lity and Chemical Stability."
"The austenitic stainless steel used in the spent fuel storage racks is not suscep-table to stress corrosion cracking and thus, corrosion in the spent fuel storage pool environment should be of little significance during the life of the plant.
The spent fuel pool water is processed by filtration and deminerali-zation to maintain water purity and clarity.
Dissimilar metal contact corrosion (galvanic attack between the stainless steel rack assemblies and zircaloy in the fuel assemblies) should not be significant because the materials are pro-tected by highly passivating oxide films and are, therefore, at similar galvanic potentials".
Similar statements could be made for the pool liner.
In regard to Boraflex, "gualification tests have shown that Boraflex does not possess leachable halogens that could be released into the spent fuel pool water in the presence of radiation.
Similar conclusions have been made regarding the leaching of boron from the Boraflex."
The staff has reviewed the proposed surveillance program for monitoring the Boraflex in the St. Lucie Plant spent fuel storage pool and concludes that the program can revea 1 deterioration that may lead to loss of neutron absorbing capability during the life of the spent fuel racks.
In the unlikely event of Boraflex deterioration, the monitoring program will detect such deterioration and the licensee will have sufficient time to take corrective action.
The reinforced concrete which supports the liner, pool water,
- racks, and spent fuel assemblies is massive in the X-Y plane and in the vertical direction below the pool.
There will be no deterioration of the concrete as a result of the reracking.
The Commission's Waste Confidence Decision goes one step further (as related in this matter which addresses the operating license life of the facility) in considering the long term effects of spent fuel storage.
The fourth Commission Finding states "The Commission finds reasonable assurance that, if necessary, spent fuel oenerated in an reactor can be stored safel and without segno scant env)ronmenta impacts or at east ears e ond the expiration of that reactor's operatin license at that reactor s spent ue stora e basin, or at e)t er ons)te or o
sste
)ndepen en spen ue storage
)nsta at)ons.
(underlining added for emphasis).
In support of this finding an ana lysis was presented on the long-term integrity of spent fuel under water pool storage conditions in addition to an analysis of structure and component safety for extended faci lity operations for storage of spent fuel in water pools.
Based upon the undisputed facts set forth above, the materials associated with the St. Lucie pool and its structure have been adequately evaluated and will perform their designed function for a few decades.
I have reviewed the affidavits provided on behalf of the licensee in response to the contention.
The analyses present a more specific case for the St. Lucie pool and structure materials,
- whereas, I took a more generic approach.
I agree with the licensee that (I) the reinforced concrete spent fuel pool structure will withstand the radiation and heat levels expected as a result of the
- rerack, (2) the structure will withstand the expected thermal
- loads, (3) there are no materials degradation concerns for the stainless steels in the pool liner and storage
- racks, (4) any radiation heat effects on the fuel cladding and fuel assembly materials attributable to increasing the capacity of the pool are negligible when compared to prior reactor exposure, and (5) the suitability of using Boraflex as a neutron absorber for the storage of spent fuel in the pool has been demonstrated.
Based upon the above, I believe that materials used in the pool and structure will fulfilltheir design function for the length of time authorized by the license amendment at issue.
7.
Admitted Contention 4
Admitted Contention 4 reads as follows:
That the high-density design of the fuel storage racks will cause higher heat loads and increases in water temperature which could cause a loss-of-cooling accident and/or challenge the reliability and testability of the systems designed for decay heat and other residual heat removal, which could, in turn, cause a major release of radioactivity into the environment.
(Originally Amended Petition Contention 8).
tiemorandum and Order, Appendix A, p. I (April 20, 1988).
The bases for Admitted Contention 4 are stated as follows:
a)
The NRC has stated in numerous documents that the water in spent fuel pools would normally be kept below 122 degrees F.
The present temperature of the water at the St. Lucie plant, Unit No.
1 is estimated to be 110 degrees F.
After the reracking the temperature of the water would rise to 152 degrees F on a
normal basis, and could reach 182 degrees F with a full core load added.
b)
There is also the possibility that a delay in the make-up emergency water could cause the zirconium cladding on the fuel rods to heat up to such high temperature that any attempt at later cooling by injecting water back into the pool could hasten the heat up, because water reacts chemically with heated zirconium to produce heat and possible explosions.
Thus, the zirconium cladding could catch on fire especially in a high-density design and create an accident not previously evaluated.
Amended Petition, pp. 6-7.
In admitting this Contention, the Licensing Board stated that it expects the Intervenor to present direct technical testimony for the record.
Yemorandum and Order, p.
20 (Apri 1 20, 1988).
In this regard, the Intervenor clarified his contention during the Harch 29, 1988 Prehearing Conference by stating that his basic concern with the pool cooling system "comes down to technical calculations."
Tr. 69.
Intervenor further indicated that he or his experts will provide substantial technical evidence that will show (I) the Licensee's temperature calculations are inadequate and will be exceeded, and (2) the Licensee's calculations do not adhere to the guidelines as set forth in the SRP, Section 9.1.3.
Id.
The intervenor contends that the rerack will cause higher heat loads and increases in water temperature which could cause a loss of cooling accident and/or challenge the reliability and testability of the systems designed for decay heat.
The above sequence of events would lead to a major release.
The basis for the contention consists of higher than normal pool water temperature and the possible delay in makeup water leading to zirconium cladding fire.
The spent fuel pool water temperature changes with time and is dependent upon how much spent fuel is placed in the pool, the decay time of each individual assembly before it is placed in the pool, and whether one or two spent fuel pool pumps are operating with the single heat exchanger.
The maximum pool water temperature guidelines allowed by the Standard Review Plan (SRP) are 140'F under normal refueling conditions (approximately I/3 core offload) and below boiling for an abnormal refueling condition (full core offload).
A single active failure to the system is assumed in the normal refueling condition and a single active failure need not be considered under the abnormal refueling condition.
Calculations were performed for both of these conditions ard the results showed that the acceptance criteria of the SRP would be met.
The maximum water temperature (as a function of time) was determined to be 134'F for the normal refueling condition and 167'F for the abrormal refueling condition.
These temperature calculations assumed only one of two pumps were operating.
I am in agreement with the staff calculations and resultant values.
Thus, there would be no loss of coolant because of boiling under these conditions and the reliability and testability of the systems designed for decay heat removal would not be affected.
The spent fuel pool cooling system at the St. Lucie Plant, Unit No.
1 is not seismic Category I.
As such, the SRP identifies the need to address this contingency should an earthquake occur.
This SRP contingency can be considered similar to the intervenor' concern of delay of makeup emergency water, should it be needed.
Implicit with a delay in makeup water is the assumption that the cooling system has failed.
The same two conditions postulated above were assumed coincident with an earthquake which was assumed to disable the spent fuel pool cooling system.
I and the staff reviewed the licensee's analysis.
The calculated boil-off rates and time to boil were estimated to be 33.9 gpm and 16.79 hours9.143519e-4 days <br />0.0219 hours <br />1.306217e-4 weeks <br />3.00595e-5 months <br /> for the normal refueling condition and 69.5 gpm and 7.47 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br /> for the abnormal refueling condition.
In addition, there is a significant delay between the time the pool starts boiling and the water level recedes appreciably.
The licensee calculated results are acceptable.
The licensee identifies a number of makeup water sources'.
These are the Refueling Water Tank, Primary Water Tank, City Water Tank, and Seismic Category I, Intake Cooling Water System.
The Intake Cooling Water System can provide 150 gpm makeup in a reasonable period of time.
The three pumps associated with the system can be powered by the emergency diesel generators and any one pump can fulfillthe function.
Thus, I conclude that adequate makeup can be provided to the pool so that the fuel would not be put in jeopardy in the highly unlikely event of failure of the cooling system.
I have reviewed the affidavits provided on behalf of the licensee in response to this contention and I agree with the conclusions that operation of the spent fuel pool cooling system will maintain pool temperatures below the values stated in the SRP and should forced cooling be lost and boiling occurs, the fuel would remain well covered with water, at a safe temperature.
It can be concluded that the pool water temperature can be adequately controlled and cooling o
the spent fuel can be adequately accomplished under various refueling conditons, even assuming an earthquake or other initiating event causing all loss of forced pool cooling.
8.
Admitted Contention 5
Admitted Contention 5 reads as follows:
That the cooling system will be unable to accommodate the increased heat load in the pool resulting from the high density storage system and a full core discharge in the event of a single failure of any of the pumps on the shell side of the cooling system and/or in the case of a. single failure of the electrical
'power supply to the pumps on the pool side of the spent fuel pool cooling system.
This inability will, therefore, create a greater potential for an accidental release of radioactivity into the environment.
(Originally Amended Petition Contention 9).
Memorandum and Order, Appendix A, p.
2 (April 20, 1988).
)lo basis was specified to support Amended Contention 5.
In admitting this Contention, the Licensing Board stated that the "Licensee' evidence on this contention should be directed toward applicabil-ity of and compliance with Criterion 44 of 10 CFR Part 50, Appendix A."
memorandum and Order, p.
22 (April 20, 1988).
At the triarch 29, 1988 Prehearing Conference, Intervenor emphasized his concern over the alleged vulnerability of the electrical power supply; in particular, to the effects which humidity, wear, corrosion, elevated temperatures and exposure to radiation would have on components.
Tr. 80.
Essentially, the Contention alleges that, if a pump or pump power supply fails, then the spent fuel pool cooling system will be unable to accommodate the increased heat load associated with the higher density fuel storage under full discharge conditions.
Memorandum and Order, p.
21 (April 20, 1988).
The contention interrelates with Admitted Contention 4 because heat loading and single failure analysis are addressed in the calculation of pool water temperatures under normal refueling conditions, abnormal refueling conditions, and loss of all spent fuel pool cooling.
The contention goes one step further and raises the concern of accidental release of radioactivity into the environment.
As stated in the previous contention, the SRP does not assume a single active failure in the abnormal refueling condition.
I concluded in the previous contention that the spent fuel pool cooling system will maintain pool temperatures below the values stated in the SRP and should forced cooling be lost and boiling occurs, the fuel would remain well covered with water, at a safe temperature.
The radiological dose case which fully addresses the contention is the abnormal refueling condition coupled with complete loss of the spent fuel pool cooling system.
This is a very unlikely event because full core offloading is typically performed only a few times over the life of the facility and the spent fuel pool cooling system is highly reliable.
Makeup water is supplied to the pool to compensate for boil off.
The makeup water could be supplied by the Seismic Category I Intake Cooling Water system which can be powered by the emergency diesel generators.
The results of this analysis were previously sent from Mr. B.
H. Vogler to the intervenor by letter dated April 15, 1988.
I prepared the enclosure to the letter and I will summarize the results here.
The total radiation dose to the thyroid from iodine exposure was calculated to be approximately
- 0. 1 rem.
The5 total radiation dose to the whole body was calculated to be approximately 10 rem.
The radiation dose values are applicable to both the exclusion area boundary and the low population zone boundary.
These values are well below the 10 CFR 100 guidelines doses of 300 rem to the thyroid and 25 rem to the whole body.
I have reviewed the affidavits provided on behalf of the licensee in response to the contention.
The affidavits also focused on electrical power supply fai lure considerations for various systems.
As specified above, the SRP does not call for assumption of a single active failure for the abnormal refueling condition case.
Thus, the affidavits provided on behalf of the licensee go beyond the SRP guidance in this case.
The affidavits did not address the expected radiological 'doses as a result of pool boiling.
My affidavit does.
I agree that the spent fuel pool water will be maintained within acceptable limits and if complete loss of forced cooling 'were to occur, the fuel would be kept covered and maintained at a safe temperature.
In conclusion, and as concluded in Admitted Contention 4, the pool water temperature can be adequately controlled and cooling of the spent fuel can be adequately accomplished under various refueling conditions, even assuming an earthquake or other initiating event causing loss of all forced pool cooling.
In addition, in the highly unlikely event of loss of forced pool cooling, the radiation doses would be very small.
9.
Admitted Contention 6 Admitted Contention 6 reads as follows:
Admitted Contention 6
The proposed use of high-density racks designed and fabricated by the Joseph Oat Corporation is utilization of an essentially new and unproven technology.
(Originally Amended Petition Contention 11).
tfemorandum and Order, Appendix A, p.
2 (April 20, 1988).
The bases for the contention read:
Bases for Contention As recently as 8 September 1987, the NRC has provided information concerning these racks to all nuclear power reactor facilities warning of a ".
. potentially significant problem pertaining to gaps "The concern is that separation of the neutron absorbing material used in high-density fuel storage racks might compromise safety."
(NRC Information Notice No. 87-43.
SSINS NO.:
6835).
Again on 23 October 1987, the NRC is requiring more information of FPKL in order to assess the integrity of the Boraflex system.
The answer to this latest inquiry has not yet been made available to the public.
FPSL's response to these and other problems relating to the use of Boraflex incorporated in a system designed by the Joseph Oat Corp. represents an essential modification of the current technology to such an extent that it, -in fact, represents utilization of a new technology and fabrication process that is, thus, unproven and untested.
Amended Petition, page 8.
In admitting this contention the Board stated:
While the use of Boraflex may not be considered "new technology,"
the problems identified in the NRC Staff Board Notifications concerning the reports on the guad Cities and Point Beach plants raise specific questions about the use of Boraflex in the Joseph Oat storage racks.
memorandum and Order, pg 24 (April 20, 1988).
The intervenor contends that the use of high-density racks designed and fabricated by the Joseph Oat Corporation is utilization of an essentially new and uproven technology.
The basis of the contention hinges on problems found with Boraflex, a neutron absorption material that is an integral part of the racks.
The use of the high-density racks designed and fabricated by the Joseph Oat Corporation is not a utilization of a
new and uproven technology.
I also acknowledge that Boraflex gap problems were found elsewhere.
The Boraflex is used for criticality control.
Boraflex has been used in racks fabricated by various companies including Joseph Oat Corporation since the early 1980's.
The St. Lucie racks were fabricated using proven technology.
I visited the Joseph Oat Corporation on October 29, 1987 and observed racks during various stages of fabrication.
Stainless steel half channels were mated together to form a channel or box.
A Boraflex strip was placed on each side of a Region I channel and secured by a cover plate.
Region II channels do not have cover plates 'and the Boraflex is held in place by the mating of two channels.
The channels are subsequently welded together; the module is subsequently welded to the base plate and girdle bars are attached.
guality control personnel independ-ently checked all the above described steps.
The Boraflex gap problem identified elsewhere should not occur at St. Lucie.
The gap problem was due to the restraining of the Boraflex during the fabrica-tion of the racks and that upon irradiation, the Boraflex shrunk and gaps formed because of restraint on parts of the Boraflex.
In the St. Lucie case, the use of adhesives in the attachment of slightly oversized Boraflex to the rack cell was not permitted and the manufacturing process avoided techniques that could p'.nch the Boraflex.
Therefore, the Boraflex should be able to slightly shrink unrestrained under irradiation.
- Lastly, I consider this change of technique in placing the Boraflex on the channel and not permitting restraint to be a
refinement of the whole fabrication process.
It, in itself, does not make the entire fabrication process a
new and uproven technology.
The licensee has developed a surveillance program to verify the performance of the Boraflex.
In the event, of any observed deterioration of the specimen that cou ld have an effect on nuclear criticality, an immediate inspection of the poison panels in the rack will be performed.
The NRC will be advised accordingly.
I believe the surveillance program as outlined in the FPSL submittal of October 20,
- 1967, to the staff to be acceptable.
In addition, in the highly unlikely event that the Boraflex wi 1 1 not be able to fulfillits design function at some time in the future, the licensee's implementation of the corrective actions described in the October 20, 1987 submittal will suffice.
I have reviewed the affidavits provided on behalf of the licensee in response to this contention, and I agree that a 11 aspects of rack construction embodies proven design concepts and well-established fabrication techniques and that the racks incorporate proven technology for Boraflex installation and
'ositioning.
Lastly, the surveillance program is adequate to detect greater than expected degradation and sufficient opportunity is available to take any corrective action that may be necessary.
In summary, I do not believe that the use of high-density racks designed and fabricated by Joseph Oat Corporation to be an essentia 1
new and unproven technology.
10.
Admitted Contention 7
Admitted Contention 7 reads as follows:
Admitted Contention 7
That the increase of the spent fuel pool capacity, which includes fuel rods that are more highly enriched, wi 11 cause the requirements of At!SI-N16-1975 not to be met and will increase the probability that a criticality accident will occur in the spent fuel pool and will exceed 10 CFR Part 50, A 62 criterion.
(Originally Amended Petition Contention 15).
tlemorandum and Order, Appendix A, p.
2 (April 20, 1988).
The bases for the contention rea'd as follows:
Bases fot Contention The increase in the number of fuel rods stored and the fact that many of them may have experienced fuel failure or may be more highly enriched and have more reactivity wi 11 increase the chances that the fuel pool will gc critical, and cause a major criticality accident and perhaps, explosion that wi 11 release large amounts of radioactivity into the environment in excess of the 10 CFR 100 criteria.
Amended Petition, p.
11.
The Licensing Board amended Intervenor!s contention as originally filed to delete reference to failed fuel, and admitted the contention.
See Hemorandum and Order, p.
28 (April 20, 1988).
In admitting this contention, the Licensing Board stated that "Lcjriticalitycontrol is one of the basic concerns when fuel is being stored, and the methods used to achieve this control are of great importance.
The contention is therefore admitted."
t~iemorandum and Order, p.
28 (April 20, 1988).
The intervenor is concerned that a criticality accident has a higher probability of occurring as a result of increasing the spent fuel capacity and allowing higher enriched fuel in the pool.
Criticality consideration is a prime input in the design of'he spent fuel racks.
Since the spacing between the fuel assemblies in a given rack is now smaller and higher enriched fuel would be permitted, a neutron absorbing medium (Boraflex in this case) for both Regions must be used in order to meet regulatory requirements.
In addition, credit for burnup is also permitted for the Region II racks.
The K
of an individuai fue'I assembty decreases whi'le the fuel assembly is produc'f" energy in the reactor.
The SRP calls for K f to be not greater than 0.95 for normal storage conditions assuming that all thVracks contain fuel assemblies and are fully covered with water.
Bo credit is allowed for the boron in the pool water, but credit may be taken for the neutron absorbing capability of Boraflex in both regions and for fuel burnup in Region II.
Credit may be taken for the boron in the pool water
for an abnormal storage condition since the assumption of two independent and concurrent failures need not be considered.
The K
would decrease for most abnormal storage conditions but couid increase for~Qme.
For example, a loss of the cooling system will result in an increase in pool water temperature but K ff will decrease.
Conversely, if a fresh assembly was placed in a Region II storage location, the K << would increase.
However, the boron in the water is worth approximately 0.2 fh K ff and this would more than offset this abnormal storage condition.
The staff's review of the licensee's physics input parameters
( including using a maximum of 4.5 w/o fuel), methodology (including uncer tainty analysis),
and results, concluded that the K << wi 11 be no greater than 0.95.
The value calculated by the licensee was 0.$ 409 for Region I and 0.9435 for Region II.
The staff concluded that most abnormal storage conditions will not result in an increase in K
but for some events, the K f could increase;
- however, the K rr decrease carfhd by boron in the water mar&ban offsets the reactivity adffftion caused by credible accidents.
I am in agreement with the staff's conclusions and no credible criticality accident is possible.
Thc staff and I believe that subcriticality control is so important that
. technical specifications have been put in place.
Technical Specification 5.6.1 requires that (1) a boron concentration of at least 1720 ppm will be kept in the pool water, (2) non-temporary placement of spent fuel in Region II will only be permitted if the assembly meets the burnup requirements of Figure 5.6-1, and (3) all fuel assemblies must have an initial U-235 enrichment less than or equal to 4.5 weight percent.
I have reviewed the affidavits provided on behalf of the licensee in response to this contention.
I agree with the licensee that the design of the St.
Lucie 1 storage racks provide assurance that a criticality accident cannot occur under any credible postulated condition.
Based upon the above discussion, I conclude that a criticality accident will not occur at the St. Lucie Unit No.
1.
The applicable technical specifications must continually be met.
11.
In summary, I am satisfied that the St. Lucie Unit 1 Spent Fuel Pool racks were adequately
- designed, fabricated, installed, and will operate as a safe spent fuel storage system unti I the expiratiorr of the license in 2016 and that all NRC requirements have been met.
I am satisfied that the contentions have been adequately addressed, and I support the licensee's Notion for Summary Disposition.
The foregoing statement are true and correct to the best of rqy knowledge and belief.
r /gin ~> +
r W /
// qgz~w~zp J Subscribed and sworn to before me this~of
, isfI'on our>gny
/ (g iotary u
s c Ny commission expires:
I 0
a tn D
Exhibit A Edmond G. Tourigny Project Manager, U.S.N.R.C.
Washington, 0.
C.
gualification and Experience Page 1 of 4 Experience U.S. Nuclear Re u lator Commission; Washin ton, D.C., Operatin Reactors Prospect Yiana er; ctober 1
resent Just under seven years operating nuclear power plant experience including:
managing hundreds of safety and environmental reviews associated with license amendments, NRC rule exemptions, and ASHE code reliefs; performing independent technical evaluations in support of license amendments, NRC rule exemptions, and ASHE code reliefs; conducting extensive plant system walkdowns comparing technical specifications, implementing procedures, PXIO's, and FSAR descriptions; evaluating plant operational events/licensee response/licensee resolution; evaluating plant changes and modifications made under 10 CFR 50.59; Extensive knowledge of NRC power plant regulations, regulatory
- guides, and standard review plan; Able to perform evaluations associated with accident ana lysis, fuel reloads, inservice inspection, fire protection, equipment qualification, and missile protection; Received extensive NRC trainsng in Westinghouse, Combustion Engineering, Babcock and Wilcox, and GE NSSS; and Balance of P 1 ant;.
Trained on Sequoyah Simulator (W), Calvert Cliffs Simulator (C.E.),
and Bellefonte Simulator (BXW);
Project Nanager assignments included St. Lucie, Unit Nos.
1 X 2, Forth Calhoun, ANO - 2, and Trojan;
Page 2 of 4 U.S. Nuclear Regulator Commission; Washin ton, D.C.,
adioactsve aste 1ana ement ro ram ana er prs 1979-cto er 1
Two and one half years low level waste management experience including:
development and maintenance of multi-year program plans which contained goals, objectives, planned accomplishments, and resource requirements; coordination of technical work performed inside NRC and outside NRC by other agencies and contractors; budget preparation and execution; Interfaced with outside groups such as state legislators, public advisory groups, etc.;
Evaluated proposed Congressional bills and prepared Congressional testimony; Interfaced with high level waste program managers.
U.S. Nuclear Re ulator Commission.
Washin ton, D.C., Nuclear ue c e and l1atersa s
ro ram Ana st pri 19 prs 1979)
Na-'or author of Congressionally requested study entitled "Regulation of Federal Radioactive Waste Activities", (NUREG-0527).
Was responsible for ca tegorizing a
1 1 federal nuclear waste by type and location.
Has in-depth knowlege of OOE sites which process,
- store, and dispose nuclear waste.
Hajor author of Commission requested study entitled "The License Renewal Study for Parts 30, 40, and 70 Licenses."
Study was the first in-depth analysis of why the NRC issues Parts 30, 40, and 70 licenses for 5 years.
Extensive knowledge of NRC fuel cycle and material regulations.
U.S. Nuclear Re viator Commission, Washin ton, O.C.
Special Assistant; anuary prl Viajor author of first Commission requested NRC Five Year Plan.
Page 3 of 4 The Penns lvania State Universit Universit
- Park, PA ra uate sssstant; anuar ecem er Served as a Graduate Assistant to the Director of the Nuclear Reactor Facility including:
development and utilization of nuclear design computer codes for in-core analysis of commercial LWR's; assisting in the instruction of students in the use of computer codes for core design and fuel management.
Graduate student in Industrial Engineering.
Awarded full tuition and stipend.
Westin house Electric Cor oration, Pittsbur h, PA Harketin n sneer; cto er 19 1 - Decem er 1
Performed technical and economic analysis of nuclear fuel cycl es.
Provided support to nuclear fuel and nuclear plant sales managers.
Nuclear fIaterials and E ui ment Cororation A olio PA iuc ear n sneer; arc cto er Two and one half years of power reactor design experience including:
proposal preparation to market commercial nuclear fuels; reactor core physics analysis; fuel management studies; nuclear computer code benchmarking studies.
The Penns lvania State Universit, Universit
Graduate student in Nuclear Engineering.
Awarded full tuition and stipend.
Page 4 of 4 Formal Education MASTER OF ENGINEERING - INDUSTRIAL The Pennsylvania State University, Industrial Engineering Department, 1976.
Thesis:
The Reliabi lit With and Without Repair and Availabilit of Four E ement stems.
MASTER OF ENGINEERING - NUCLEAR The Pennsylvania State University, Nuclear Engineering Department, 1970.
Thesis:
The Neutronics Calculation of a
Pressurized Water Reactor.
BACHELOR OF SCIENCE - NUCLEAR
,Lowell Techno logy Institute, Nuclear Engineering Department, 1968.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of FLORIDA POWER AND LIGHT COMPANY (St. Lucie Plant, Unit No. I)
CERTIFICATE OF SERVICE Docket No. 50-335-OLA (SFP Expansion)
I hereby certify that copies of "RESPONSE OF NRC STAFF IN SUPPORT OF LICENSEES MOTION FOR
SUMMARY
DISPOSITION" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or as indicated by an asterisk through deposit in the Nuclear Regulatory Commission's internal mail system, this 30th day of August, 1988:
B. Paul Cotter, Jr.,
Chairman*
Administrative Judge Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Richard F. Cole*
Administrative Judge Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Atomic Safety and Licensing Board Panel (I)*
U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Atomic Safety and Licensing Appeal Panel (5)*
U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Adjudicatory File*
Atomic Safety and Licensing Board Panel Docket U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Glenn 0. Bright*
Administrative Judge Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Michael A. Bauser, Esq.
Harold F. Reis,. Esq.
Newman
& Holtzinger, P.C.
1615 L Street, N.W.
Washington, D.C.
20036 Docketing and Service Section*
Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Campbell Rich 4626 S.E. Pilot Avenue
- Stuart, Florida 34997 en amain og er Senior Supervisory Trial Attorney
v s~