ML17221A359
| ML17221A359 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/25/1987 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17221A358 | List: |
| References | |
| NUDOCS 8708280182 | |
| Download: ML17221A359 (17) | |
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7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-335 NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE AND PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATNN AND OPPORTUNITY FOR HEARING The U.S. Nuclear Regulatory Commission (the Commission) is considering issuance of an amendment to Facility Operating License No. DRP-67, issued to Florida Power and Light Company (the licensee), for operation of the St. Lucie Plant, Unit No 1, located in St. Lucie County, Florida.
The amendment would authorize the licensee to increase the spent fuel pool storage capacity from 728 to 1706 fuel assemblies.
The proposed expansion is to be achieved by reracking the spent fuel pool into two discrete regions.
New, high-density storage racks will be used.
The existing storage racks will be removed, cleaned of loose contamination, packaged and shipped off-site.
Region I of the spent fuel pool includes 4 modules having a total of 342-storage cells.
The cell pitch is 10.12 inches.
All cells can be utilized for storage and each cell can accept new fuel assemblies with enrichments up to 4.5 weight percent U-235 or spent fuel assemblies that have not achieved adequate bur nup for Region 2.
Region 2 includes 13 modules having a total of 1364 storage cells.
The cell pitch is 8.86 inches.
All cells can be utilized for storage and each cell can accept spent fuel assemblies with various initial enrichments which have accumulated minimum burnups.
Each cell in each region can accommodate a single Combustion Engineering or Advanced Nuclear Fuel Corporation (formerly Exxon)
PWR fuel assembly or equivalent, from either St. Lucie Unit 1 or Unit 2.
8708280182 870825 PDR ADOCK 0500033' PDR
The new racks are not doubled-tiered and all racks will sft on the spent fuel pool floor.
The amendment application does not involve rod consolidation.
The k ff of the pool will be maintained at less than or equal to 0.95.
Keutron eff absorbers in the form of Boraflex will also be used for criticality control.
The rack vendor has licensed at least 10 other racks of the same design.
The construction process and analytical techniques remain substantially the same as the previous 10 racks.
- Thus, no new or improved technology is utilized in the construction or analysis of the proposed racks.
This amendment was requested in the licensee's application dated June 12, 1987.
The Coomission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c).
A proposed amendment to an operating license for a facility involves no significant hazards consid-eration if operation of the facility in accordance with the proposed amendment would not:
(1) involve a significant increase in the probability or consequen-ces of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
The licensee addressed the above three standards in the amendment appli-
- cation, as restated below.
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated.
In the course of the analysis, FPL has considered the following potential accident scenar ios:
l.
A spent fuel assembly drop in the spent fuel pool.
2.
Loss of spent fuel pool cooling system flow.
3.
A seismic event.
4.
A spent fuel cask drop.
5.
A construction accident.
The probability of any of the fi+st four accidents is not affected by the racks themselves; thus the modification cannot increase the probability of these accidents.
As for the construction accident, FPL does not intend to carry any rack directly over the stored spent fuel assemblies.
All work in the spent fuel pool area will be controlled and performed in strict accordance with specific written procedures.
The crane which will be used to bring the racks into the Fuel Handling Building has been evaluated and meets the requirements of Section 5.1. 1 of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants.,"
In addition, the temporary construction crane which will be used to move racks within the spent fuel pool area will meet the design, inspection, testing, and operation requirements of Section 5.1.1 of NUREG-0612.
This program provides for the safe handling of heavy loads in the vicinity of the spent fuel pool.
Accordingly, the proposed modification does not involve a significant increase in the probability of an accident previously evaluated.
FPL evaluated the consequences of a spent fuel assembly drop in the spent fuel pool (scenario
- 1) and found that the criticality acceptance criterion, k f less than or equal to 0.95, is not violated.
In addition FPL found thIVthe radiological consequences of a fuel assembly drop are not changed from the previous analysis.
The NRC also conducted an evalua-tion of the potential consequences of a fuel handling accident.
Both FPL:
and NRC analyses found that the calculated doses are less than 10 CFR Part 100 guidelines.
The results of an analysis show that a dropped spent fuel assembly on the racks will not distort the racks such that they would not perform their safety function.
Thus, the consequences of this type accident are not changed from the previously evaluated spent fuel assembly drops which have been found acceptable by the NRC.
The consequences of a loss o+ spent fuel pool cooling system flow (scenario
- 2) have been evaluated and it was found that sufficient time is available to provide an alternate means for cooling (i.e., the fire hose stations) in the event of a failure in the cooling system.
Thus, the consequences of this type accident are not significantly increased from previously evaluated loss of cooling system flow accidents.
. The consequences of a seismic event (scenario
- 3) have been evaluated and are acceptable.
The new racks will be designed and fabricated to meet the requirements of applicable portions of the NRC Regulatory Guides and published standards.
The new free-standing racks are designed, as are
4-the existing free-standing
- racks, so that the floor loading from racks completely filled with spent fuel assemblies, partially filled, or empty at the time of the incident, do not exceed the structural capability of the spent fuel pool.
The Fuel Handling Building and spent fuel pool structure have been evaluated for the increased loading from the spent fuel racks in accordance with the criteria previously evaluated by the NRC and found acceptable.
Thus, the consequences of a seismic event are not significantly increased from previously eva'equated events.
The consequences of a spent fuel cask drop {scenario 4) have been evaluated.
The radiological consequences of the cask drop are well within the guide-lines of 10 CFR 100 and the doses are not increased as compared to the doses analyzed for the presently installed racks.
The cask drop analysis is based on administrative and Technical Specification controls which ensure that minimum requirements f'r decay of irradiated fuel assemblies in the entire spent fuel pool are met prior to movement of the cask into the cask area of the spent fuel pool.
Analyses also demonstrate that k
will always be less than the NRC acceptance criterion.
In addition[.,l 15ffage from a cask drop will not exceed the makeup capabilities of the spent fuel pool.
Thus, the consequences of a cask drop accident will not increase from previously evaluated accident i~analyses].
The consequences of a construction accident (scenario
- 5) are enveloped by the spent fuel cask drop analysis previously performed by FPL.
In addition, all movements of heavy loads handled during the rerack operation will comply with the NRC guidelines presented in NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."
The consequences of a construction accident are not increased from previously evaluated accident, [analysesj.
Therefore, it is concluded that the proposed amendment to replace the spent fuel racks in the spent fuel pool will not involve a significant in-crease in the probability or consequences of an accident previously evalu-ated.
(2)
Create the possibility of a new or different kind of accident from any accident previously evaluated.
FPL has evaluated the proposed modification fn accordance with the guidance of the NRC position paper entitled, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," appropriate NRC Regula-tory Guides, appropriate NRC Standard Review Plans, and appropriate industry codes and standards.
In addition, FPL has reviewed several revious NRC Safety Evaluation Reports for rerack applications similar to its" proposal.
As a result of this evaluation and these reviews, FPL finds that the proposed modification does not, in any way, create the possibility of a new or different kind of accident from any accident previously evalua-ted for the St. Lucie spent fuel storage facility.
(3)
Involve a significant reduction in a margin of safetyf.3 The NRC staff Safety Evaluation review process has established that the issue of margin of safety, when applied to a reracking modification, should address the following areas:
1.
Nuclear criticality considerations 2.
Thermal-hydraulic considerations 3.
Mechanical, material and structural considerations.
The established acceptance criterion for criticality is that the neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions.
This margin of safety has been adhered to in the criticality analysis methods for the new rack design.
The methods used in the criticality analysis conform with the applicable portions of the appropr iate NRC guidance and industry codes, standards, and specifications.
In meeting the acceptance criteria for criticality in the spent fuel pool, such that k
is always less than 0.95, including uncertainties at a
95%/95% probaSRity confidence level, the proposed amendment to rerack the spent fuel pool does not involve a significant reduction in a margin o< safety for nuclear criticality.
Conservative methods are used to calculate the maximum fuel temperature and the increase in temperature of the water in the spent fuel pool.
The thermal-hydraulic evaluation uses the methods used for evaluations of the present spent fuel racks in demonstrating the temperature margins of safety are maintained.
The proposed modification will increase the heat load in the spent fuel pool.
The evaluation shows that the existing spent fuel cooling system will maintain the bulk pool water temperature at or below 150.8'F.
Thus a margin of safety exists such that the maximum allow-able temperature of 217'F is not exceeded for the calculated increase in pool heat load.
The evaluation also shows that maximum local water temper-atures along the hottest fuel assembly are wel~ below the nucleate boiling condition values.
Thus, there is no significant reduction in the margin of safety for thermal-hydraulic or spent fuel cooling concerns.
The main safety function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a safe configuration through all normal or abnormal
- loadings, such as an earthquake, impact due to a spent fuel cask drop, drop of a spent fuel assembly, or drop oF any other heavy object.
The mechanical, material, and structural design of the new spent fuel racks is in accordance with applicable portions of the "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications,"
dated April 14,
- 1978, as modified January 18, 1979; Standard Review Plan 3.8.4; and other applicable NRC guidance and industry codes..
The rack materials used are compatible with the spent fuel pool and the spent fuel
- assemblies.
The structural considerations of the new racks address margins of safety against tilting and deflection or movement, such that
~
~ the racks are not damaged during impact.
In addition the spent fuel assemblies remain intact and no criticality concerns exists.
Thus, the margins of safety are not significantly reduced by the proposed rerack.
The staff has reviewed the licensee's no significant hazards consideration determination analysis and agrees with their conclusions.
However, the staff believes that the licensee's no significant hazards consideration (NSHC) analy-sis could have been more explicit in z number of areas.
These areas are (1) pool water temperature under normal and abnormal conditions, (2) recent boraflex problems, and (3) construction accidents.
The licensee'tates in the NSHC analysis that the safety evaluation shows that the existing spent fuel cooling system will maintain the bulk pool water temperature at or below 150.8'F, and that a margin of safety exists such that the maximum allowable temperature of 217's not exceeded.
This statement addresses the abnormal maximum heat load (full core unload) case; the staff's Standard Review Plan (SRP) for full core unload calls for the pool water temper-h ature to be kept below boiling.
Thus, the licensee's analysis and results for this case meet the SRP.
The licensee did not address the maximum normal heat load case in the NSHC analysis.
The staff's review of the licensee's associated safety evaluation concludes that the maximum normal heat load case was also evaluated.
The licensee calculated a maximum pool water temperature of 133.3'F.
The SRP states that the pool should be kept at or below 140'F in this case.
Thus, the licensee's analysis and results for this case meet the SRP.
The licensee's NSHC analysis did not address the recent operational prob-lems associated with boraflex, a neutron absorbing material that fs utilized in many racks to maintain the k ff of the pool less than or equal to 0.95.
I
Although some shrinkage of the boraflex is assumed and accounted for, cracking of the boraflex and the forming of significant axial gaps has not been postu-lated to occur.
It is believed that cracking occurred in some applications because of the rack design and fabrication process which did not allow the boraflex to shrink without cracking.
The staff has reviewed the licensee's associated safety evaluation with particular focus on the method that the bora-flex will be installed.
In Region I, full-length strips of boraflex will be placed between the cell walls and a stainless steel coverplate.
In Region 2, full-length boraflex strips will be placed between the adjacent cell walls.
The licensee's specification for the handling and installation of the bora-flex requires that it will not be installed in a stretched condition.
The specification precludes the use of adhesives in the attachment of the boraflex to the rack cell walls.
FPAL will require that the manufacturing process avoid techniques which could pinch the boraflex.
The design of the racks requires that additional lengths of boraflex, i.e., greater than the active length of a fuel assembly, be installed to account for anticipated shrinkage of the bora-flex.
Based upon the above, the staff does not envision that the proposed racks will experience the boraflex cracking problems experienced elsewhere.
Thus, the k ff of the pool will be maintained less than or equal to 0.95.
eff The most limiting construction accident postulated for the spent fuel pool by the licensee is a 25 ton cask drop accident.
This is the most limiting accident postulated at this time and it will remain the most limited as a re-sult of the proposed rerack.
This appears reasonable because no existing rack or proposed rack weighs more than 25 tons. 'n practice, the technical speci-fications prohibit any load in excess of 2 tons to be carried over irradiated
fuel in the storage pool and also prohibit,the cask crane from picking up any load over 25 tons.
Nevertheless, the licensee's updated safety analysis re-port for Unit No.
1 analyzed this postulated accident in Section 9.1.4, entitled "Fuel Handling System."
The licensee evaluated the radiological consequences of the postulated accident.
The licensee determined that the radiological consequences were within 10 CFR Part.100 guidelines.
The licensee reevaluated the postulated 25 ton cask drop accident in the safety evaluation supporting the amendment request.
This was necessary because the proposed amendment allows more spent fuel to be placed in the spent fuel pool; the results are contained in Section 5.3.1.2.
The results indicate that the radiological consequences remain within 10 CFR Part 100 guidelines.
The licensee also states that the proposed spent fuel pool modifications do not increase the radiological conse-C quences of the cask drop accident previously evaluated.
Since the licensee's request to expand the St. Lucie I spent fuel storage pool capacity satisfies the following conditions:
(1) the storage expansion method consists of replacing existing racks with a design that allows closer spacing between stored spent fuel assemblies; (2) the storage expansion method does not involve rod consolidation or double-tiering; (3) the k ff of the pool is maintained less than or equal to 0.95; and (4) no eff new technology ot unproven technology fs utilized in either the construction process or the analytical techniques necessary to justify the expansion, the Commission concludes that the request does not involve a significant hazards consideration in that it:
(1) does not involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) does not create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) does not involve a significant reduction in a margin of safety.
Because the submittal and the above discussion by the licensee appear to demonstrate that the standards specified in 10 CFR 50.92 are met, and because reracking technology has been well-developed and demonstrated, the Comaission proposes to determine that operation of the facility in accordance with the proposed amendment does not involve a significant hazar ds consideration.
The Commission is seeking public comnents on this proposed determina-tion.
Any cements received within 30 days after the date of publication of this notice will be considered in making any final determination.
The Comnission will not normally make a final determination unless it receives a request for a hearing.
Written comnents may be submitted by mail to the Rules and Procedures Branch, Division of Rules and Records, Office of Administration, U.S. Nuclear Regulatory Comnission, Mashington, D.C.
- 20555, and should cite the publica-tion date and page number of this FEDERAL REGISTER notice.
Written comments may also be delivered to Room 4000, Maryland National Bank Building, 7735 Old Georgetown Road, Bethesda, Maryland from 8:15 a.m. to 5:00 p.m.
Copies of written conments received may be examined at the NRC Public Document
- Room, 1717 H Street, N.M., Mashington D.C.
The filing 'of requests for hearing and petitions for leave to intervene is discussed below.
By September 30, 1987, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written petition for leave to intervene.
Request for a hearing and petitions for leave to intervene shall be filed in accordance with the Comnission's "Rules of Practice for Domestic Licensing Proceedings" in 10 CFR Part 2. If a request for a hearing or petition for leave to intervene is filed by the above date, the Comoission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of hearing or an appropriate order.
As required by 10 CFR 52.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding.
The peti-tion should specifically explain the reasons why intervention should be permit-ted with particular reference to the following factors:
(I) the nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest.
The petition should also iden-.
tify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene.
Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing con-ference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which
~ ~ are sought to be litigated in the matter, and the bases for each contention set forth with reasonable specificity.
Contentions shall be limited to matters within the scope n< the amendments under consideration.
A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a
party.
W The Comnission hereby provides notice that this is a proceeding on an application for a license amendment falling within the scope of section 134 of the Nuclear Waste Policy Act of 1982 (NWPA), 42 U.S.C.
510154.
Under section 134 of the NWPA, the Conmission, at the request of any party to the proceeding, is authorized to use hybrid hearing procedures with respect to "any matter which the Commission determines to be in controversy among the parties."
The hybrid procedures in section 134 provide for oral argument on matters in controversy, preceded by discovery under the Commission's rules, and the desig-nation, following argument, of only those factual issues that involve a genuine and substantial dispute, together with any remaining questions of law,'o be resolved in an ad)udicatory hearing.
Actual adiudicatory hearings are to be held on only those issues
~ound to meet the criteria of section 134 and set for hearing after oral argument.
The Cmmission's rules implementing section 134 of the NWPA are found in 10 CFR Part 2, subpart K, "Hybrid Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors" (published at 50 FR 41662, October 15, 1985) 10 CFR 52.1101 et
~se Under those rules, any party to the proceeding may invoke the hybrid hearing procedures by filing with the presiding officer a written request for oral argument under 10 CFR 2.1109.
To be timely, the request must be filed within ten (10) days
of an order granting a request for hearing or petition to intervene.
(As outlined above, the Ceanission's rules in 10 CFR Part 2, subpart 6, and 52.714 in particular, continue to govern the filing of requests for a hearing or petitions to intervene, as well as the admission of contentions).
The presiding officer shall grant a timely request for oral argument.
The presiding officer may grant an untimely request for oml argument only upon a showing of good cause by the requesting party for the failure to file on time and after providing the other parties an opportunity to respond to the untimely request.
If the presiding officer grants a request for oral argument, any hearing held on the application shall be conducted in accordance with the hybrid hearing procedures.
In essence, those procedures limit the time available for discovery and require that an oral argument be held to determine whether any contentions must be resolved in an adjudicatory hearing.
If no party to the proceeding requests oral argument, or if all untimely requests for oral "argument are denied, then the usual procedures in 10 CFR Part 2, subpart 6 apply.
Subject to the above requirements and any limitations in the order granting leave to intervene, those permitted to intervene become parties to the proceed=
ing and have the opportunity to participate fully in the conduct of any hearing which is held, including the opportunity to present evidence and cross-examine witnesses at such hearing.
If a hearing is requested, the Ccemission will make" a final determination on the issue of no significant hazards consideration.
The final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it effective, notwithstanding the request for a hearing.
Any hearing held would take place after issuance of the amendment.
If the final determination is that the amendment involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.
Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period.
However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards con-sideration.
The final determination will consider all public and State comments received.
Should the Coomission take this action, it will publish a notice of issuance and provide for opportunity for a hearing after issuance.
The Comnis-sion expects that the need to take this action will occur very infrequently.
A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Coamission, U.S. Nuclear Regulatory Comoission, Mashington, D.C.
20555, Attention: Docketing and Service Branch, or may be delivered to the Commission's Public Document
- Room, 1717 H Street, N.W.,
Mashington, D.C., by the above date.
Mhere petitions are filed during the last ten (10) days of the notice period, it is requested that the petitioner promptly so inform the Commission by a toll-free telephone call to Mestern Union at (800) 325-6000 (in Missouri (800) 342-6700).
The Mestern Union operator should be given Datagram Identification Number 3737 and the following message addressed to Herbert Berkow:
petitioner's name and telephone number; date petition was mailed; plant name; and publication date and page number of this FEDERAL REGISTER notice.
A copy of the petition should also be sent to the Office of the General Counsel-Bethesda, U.S. Nuclear Regulatory Conmission, Washington, D.C.
- 20555, and to Harold F. Reis, Esq.,
Newman
& Holtzinger, 1615 L Street, N.W.,
Washington, D.C.
20036, attorney for the licensee.
Nontimely filing of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing will not be entertained absent a determination by the Comoission, the presiding officer or the Atomic Safety and Licensing Board designated to rule on the petition and/or request, that the petitioner has made a substantial showing of good cause for the gran-ting of a late petition and/or request.
That determination will be based upon a balancing of the factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the application for amendment dated June 12, 1987, which is available for public inspection at the Commission's Public Document
- Room, 1717 H Street, N.W., Washington, D.C., and at the Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 33450.
Dated at Bethesda, Maryland, this 25th day of August, 1987.
FOR THE NUCLEAR REGULATORY COMMISSION Herb r N. Berkow, Director Pro ect Directorate II-2
,Division of Reactor Projects-I/II
August 19, 1987 DI IBUTION
~Doc et:file w/o-encl-.
D. t1iller w/encl.
E. Tourigny w/encl.
DOCKET NO(S).
50 335 and 50-389 Nr. C. 0. Moody Group Vice President Nuclear Energy Florida Power and Light Company Post Office Box 14000 Juno Beach, Florida 33408
SUBJECT:
ST.
LUCIE UNITS 1
AND 2 V
The following documents concerning our review of the subject facility are transmitted for your information.
Notice of Receipt of Application, dated Draft/Final Environmental Statement, dated Notice of Availability of Draft/Final Environmental Statement, dated Safety Evaluation Report, or Supplement No.
dated Environmental Assessment and Finding of No Significant Impact, dated Notice of Consideration of Issuance of Facility Operating License or Amendment to Facility Operating License, dated Q Bi-Meekly Notice; Applications and Amendments to Operating Licenses Involving No N ill'l*d C
id i,d d~l (l]
Exemption, dated Construction Permit No.
CPPR-
, Amendment No.
da'ted Facility Operating License No.,
Amendment No.
dated Order Extending Construction Completion Date, dated Monthly Operating Report for transmitted by letter dated Annual/Semi-Annual Report-transmitted by letter dated
Enclosures:
As stated Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation cc:
See next page
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