ML20135F468
| ML20135F468 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/06/1996 |
| From: | Hodgdon A NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML20135F471 | List: |
| References | |
| CON-#496-18099 OLA, NUDOCS 9612130075 | |
| Download: ML20135F468 (25) | |
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00CKETED l USNRC l
l Qg:e(serh J j 1p9f:39 UNITED STATES OF AMERICA i NUCLEAR REGULATORY COMMISSION l
{fFI 0C PQ R}
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
GENERAL PUBLIC UTILITY NUCLEAR )
CORPORATION ) Docket No. 50-219-OLA
)
(Oyster Creek Nuclear Generating Station) )
NRC STAFF RESPONSE IN SUPPORT OF LICENSEE'S MOTION FOR
SUMMARY
DISPOSITION INTRODUCTION The staff of the Nuclear Regulatory Commission (Staff) hereby responds, pursuant to 10 C.F.R. 6 2.749(a) and the Atomic Safety and Licensing Board's (Licensing Board's)
Memorandum and Order (Ruling on Intervention Petition), dated October 25,1996, in support of GPU Nuclear Corporation's (Licensee's) Motion for Summary Disposition (Motion), filed on November 15, 1996.
BACKGROUND On April 15,1996, the Licensee submitted a license amendment application in which it sought to change Technical Specification 5.3.1,B, providing that loads greater in weight than one fuel assembly shall not be moved over stored irradiated fuel in the' spent fuel storage facility. 4 The proposed change divided section B into two parts, providing as follows:
9612130075 961206 PDR ADOCK 05000219 i
9 PDR
i B.1. I naa greater than the weight of one fuel assembly shall not be moved over stored irradiated fuel in the spent fuel storage facility, except as noted in 5.3.1.B.2.
- 2. 'Ihe shield plug and associated lifting hardware may be moved over irradiated fuel assemblies that are in a dry shielded canister within the transfer cask in the cask drop protection system.'
On May 6,1996, the NRC staff published in the Federal Register a notice of opportunity for a hearing and a proposed finding of no significant hazards consideration. 61 Fed. Reg.
20,842; 20,842-43; 20,848 (1996). l l
On June 6,1996, the Nuclear Information and Resource Service (NIRS), Oyster Creek Nuclear Watch (OCNW) and Citizens Awareness Network (CAN) filed a request for hearing and petition to intervene.' The Licensee and the Staff filed responses opposing the petition on the basis that the petitioners lacked standing.'
On July 18, 1996, pursuant to the Licensing Board's Order of July 3,1996, the Petitioners filed their Supplemental Petition, in which they set forth a single contention, together i i
' Letter from Michael B. Roche, Vice President and Director, OCNGS, to NRC Document Control Desk (Apr. 15,1996) at unnumbered p. 6 (proposed revised OCNGS Technical Specification page 5.3-1).
2 Nuclear Information and Resource Service, Oyster Creek Nuclear Watch, and Citizens Awareness Network Request for a Hearing and Petition to Intervene on General Public Utility Nuclear License Amendment Request for Oyster Creek Nuclear Generating Station (June 6,1996).
8 GPUN's Answer Opposing Request for Hearing and Petition for Intervention of
[ Nuclear Information and Resource Service, Oyster Creek Nuclear Watch, Citizens Awareness Network] (June 21,1996); NRC Staff Response in Opposition to Request for Hearing and Petition to Intervene of [ Nuclear Information and Resource Service / Oyster Creek Nuclear Watch / Citizens Awareness Network] (June 26,1996).
l with a Reply to the Licensee's Answer and the Staff's Response.' Also pursuant to the Licensing Board's Order of July 3,1996, the Licensee filed its Answer to the Supplemental Petition on
- July 29,1996, and the Staff filed its Response on July 31,1996.8 Both the Licensee and the Staff opposed the admission of the contention as failing to meet the requirements set forth in 10 C.F.R. I 2.714 regarding contentions.
After a prehearing conference held on August 7,1996, the Licensing Board issued a Memorandum and Order (Ruling on Petition to Intervene), October 25,1996. On November 7,1996, the Staff issued the amendment, having made a final finding of no significant hazards consideration.' On November 15, 1996, pursuant to a schedule established in the Licensing i Board's Memorandum and Order of October 25, 1996, the Licensee filed its Motion for Summary Disposition.
DISCUSSION As discussed below, the Licensee has shown through its argument, the affidavit of its licensing expert and supporting documents, that the admitted contention is based on several
- Supplemental Petition of Nuclear Information and Resource Service, Oyster Creek Nuclear Watch and Citizens Awareness Network (July 18,1996); Petitioners Reply to NRC .
Staff and GPUN Answer Opposing Request for Hearing and Petition for Intervention of l 4
Nuclear Information and Resource Service, Oyster Creek Nuclear Watch and Citizens !
Awareness Network, ( July 18, 1996).
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8 GPUN's Answer to Supplemental Petition for Nuclear Information Resource '
Service, Oyster Creek Nuclear Watch and Citizen Awareness Network (July 19,1996); NRC 4 Staff Response to Petitioners Supplemental Petition (July 31,1996).
- 12tter from Ann P. Hodgdon, Counsel for NRC Staff, to Administrative Judges ,
forwarding a copy of the amendment issued November 7,1996 by the NRC Staff to General Public Utility Nuclear Corporation for its Oyster Creek Nuclear Generating Station (November 12, 1996).
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mistaken premises. The Licensee has also shown that there is no issue, either legal or factual, {
raised by Intervenors' contention and that the Licensee is, therefore, entitled to summary
- disposition as a matter of law. l 1
i By its Motion, the Licensee seeks summary disposition in its favor of the only contention j admitted in the proceeding, in which Intervenors allege that the proposed Technical Specification change, clarifying that the prohibition against lifting loads greater in weight than a single fuel !
l assembly over stored irradiated fuel in the spent fuci storage facility does not prevent the i Licensee from placing a shield plug weighing 4 tons onto the top of the dry shielded canister in the cask drop protection system in the spent fuel storage facility, cannot be granted because the I defense in depth principle underlying NUREG-0612 prevents any change to Technical Specification 5.3.1.B. As admitted by the Licensing Board, Intervenors' contention alleges that the single fuel assembly weight limitation in Technical Specification 5.3.1.B reflects an agency judgment about measures necessary for compliance with NUREG 0612's underlying defense in depth approach and that, as such, it is a vital control that, as a matter of law, cannot be amended. Memorandum and Order (October 25,1996) at 40-42.
By the affidavit of its licensing expert,' Licensee shows that Technical Specification 5.3.1.B preexisted NUREG 0612 by some three years (Fornicola Aff., 11 15, 16, 19), and that it is an artifact not of NUREG-0612 and the generic letters related to NUREG-0612 but rather of an amendment issued in 1977 permitting the Licensee to install high density racks in its spent fuel storage facility (Fornicola Aff.116-15). Additionally, the Licensee points out that the interim protective measure recommended by NUREG-0612 that licensees implement technical
' Affidavit of John C. Fornicola (Fornicola Affidavit).
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specifications like the one at issue was not adopted in the Generic Letters following NUREG-0612. Motion at 9-10; Fornicola Aff at i 19. The Affidavit of Ronald B. Eaton',
attached hereto as Attachment 1 (Eaton Aff.), and the Affidavit of Harold Walker', attached ,
hereto as Attachment 2 (Walker Affidavit), attest to the correctness of Mr. Fornicola's statements regarding these matters. Walker Aff. at 17; Eaton Aff at 115-10. ,
The Licensee characterizes the issues raised by the contention as: 1) What is the regulatory significance of NUREG-0612? and 2) Can the technical specification at issue here be changed? Motion at 5-6. The Staff agrees with the Licensee's characterization. I
- 1. The Regulatory Significance of NUREG-0612 Section 103 of the Atomic Energy Act of 1954, as amended, authorizes the Nuclear Regulatory Commission to issue licenses to persons applying therefor, to, among other things, possess and use utilization and production facilities for industrial or commercial purposes.
Atomic Energy Act of 1954, as amended (AEA) i 103, 42 U.S.C. I 2133 (1995). Such licenses i are subject to such conditions as the Commission may by rule or regulation establish to i
j effectuate the purposes and provisions of the Act. Id. Section 182 requires all applicants for
! a license to provide the Commission with such information as it may, by rule or regulation, s
deem necessary to enable it to find that the utilization or production of special nuclear material l
j will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. AEA i 182,42 U.S.C. I 2232.
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- Affidavit of Ronald B. Eaton in Support of the NRC Staff's Response in Support of j Licensee's Motion for Summary Disposition. (Eaton Affidavit.)
) ' Affidavit of Harold Walker in Support of the NRC Staff's Response in Support of Licensee's Motion for Summary Disposition (Walker Affidavit). Mr. Walker's Affidavit is attached hereto as Attachment 2.
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6-Pursuant to these provisions of the Act, the Commission established General Design
! Criteria that relate to, among other things, the movement of heavy loads. 10 C.F.R. Part 50, j Appendix A, General Design Criterion 61. NUREGs and Regulatory Guides, by their very
) nature, serve merely as guidana and cannot prescribe requirements. Curators of she Uniwrsity 1
of Missouri, CLI 95 1, 41 NRC 71, 98 (1995). Although conformance with regulatory guides will likely result in compliance with specific regulatory requirements, nonconformance with such guides does not equate to noncompliance with the regulations. Id. Only statutes, regulations, orders and license conditions can impose requirements upon applicants and licensees. Id. A licensee is free either to rely on NUREGs and Regulatory Guides or to take alternative approaches to meet legal requirements as long as those approaches have the approval of the Commission or NRC Staff. Chrators of the Uniwesity of Missouri, CLI-95-8, 41 NRC 386, 397 (1995). See 10 C.F.R. I 50.109(a)(7).
Harold Walker, Senior Reactor Engineer in the Plant Systems Branch of the Office of Nuclear Reactor Regulation and principal author of the Safety Evaluation Report prepared in connection with the issuance of the amendment that is the subject matter of this proceeding, has provided an affidavit in which he addresses the circumstances giving rise to the publication of NUREG-0612. Walker Aff, at 113-6. In his affidavit, Mr. Walker explains the regulatory !
significance of NUREGs with specific attention to the NUREG at issue here, NUREG-0612.
Walker Aff.,113-7. Mr. Walker's Affidavit also explains the relationship between the applicable General Design Criteria (regulations) and the acceptance criteria to assure that the requirements of the General Design Criteria are satisfied. Walker Aff. at 117, 9. As :
Mr. Walker points out, NUREGs such as NUREG412 represent staff guidance and acceptance
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criteria, and do not substitute for regulations. Walker Aff. at 7. Compliance with a given i
! NUREG is not required. Id.
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- 2. Can Technical Specifications like TS 5.3.1.B be changed?
l 'Ihe Licensee's argument on this issue is somewhat oblique in that its point seems to be l
l that the amendment granted by the Staff on November 7,1996, was merely a clarification and i not a substantive change, in other words, that the change did not really permit the Licensee to 1
take any action not authorized previously. Motion at 11. The Staff agrees with the argument
) as far as it goes and notes, in addition, that another licensee has changed a technical
) specification comparable to the one at issue in Oyster Creek under similar circumstances, as i
l described below.
i The Commission's regulations in 10 C.F.R. I 50.90 provide for applications for
- amendments to a license or permit "whenever a holder of a license or construction permit desires l to amend the license or permit." The only regulatory or statutory limitation on granting an 1
i amendment to a license, which includes changing technical specifications, is that the Commission must find that the license as amended provides reasonable assurance of protection of the public health and safety and that it is not inimical to the common defense and security. Section 182, A.E.A, 42 U.S.C. I 2232; 10 C.F.R. Il 50.92(a); 50.57; 50.40(c). Thus, there is no technical specification that by law cannot be changed. The Licensing Board seems to have recognized this when, in a slightly different context, it stated that, "[N]othing we are aware of in connection with section 50.36 precludes a change in the provisions of such a technical
The Staff also agrees that Consumers Power did not need to change its Technical Specifications for its Palisades plant in order to move fuel out of the spent fuel storage pool to Palisades' ISFSI. Eaton Aff. at i 7.
specification [as TS 5.3.lB] if the licensee can make the appropriate showing." LBP-96-23
> at 36. Beyond that, as stated above, the Staff is not aware of any limitation on amending licenses except the statutory limitations stated in section 182 of the Atomic Energy Act of 1954,
- as amended (42 U.S.C. I 2232), discussed above.
Mr. Eaton's affidavit discusses an amendment issued by the Staff on October 5,1995, i
a j to Sacramento Municpal Utility District for its Rancho Seco plant. Eaton Aff. at i 13. That ,
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amendment modified TS D 3.3, providing that no loads exceeding the combined weight of one . l i l
{ fuel assembly, control component and associated handling tool be handled over irradiated fuel l 1
j assemblies stored in the spent fuel pool, to provide for handling the dry shielded canister (DSC) I i
j top shield plug, the MP-187 cask lifting yoke and yoke extension, and gantry crane lower load block over irradiated fuel assemblies in the DSC in the spent fuel pool. Id.
i Thus, not only do the Atomic Energy Act and the Commission's regulations provide for l amendments to be issued on an appropriate finding by the Commission, the Staff has, in fact,
} issued at least one similar amendment prior to the one contested here, notwithstanding NUREG-l
} 0612.
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- The Licensee argues that its Technical Specification 5.3.1.B was never intended to 4 ,
1 prevent movement of a shield plug onto a canister / cask and that the same is true of the Standard l
. 4 Technical Specification, TS 3.9.7, that is comparable to TS 5.3.1.B. Motion at 20-22. The
[ Staff agrees with that argument.
i l In summary, not only may a licensee move a shield plug over spent fuel despite a l 1
Technical Specification like TS 5.3.1.B (prior to the Nov. 7th amendment) (Palisades), it may
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l 1 ; amend that Technical Specification to clarify that it can move a shield plug over spent fuel in the canister / cask (Rancho Seco).
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- CONCLUSION As discussed above, the Staffs review of the Licensee's Motion for Summary Disposition confirms that Intervenors' contention does not raise any issue of law or fact and that the f Licensee is entitled to summary deposition as a matter of law.
Respectfully submitted, l (W .
hoc W Ann P. Hodgdon
. Counsel for NRC Staff Lated at Rockville, Maryland this 6th day of December,1996.
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ATTACHMENT 1 UNITED STATES OF AMERICA i NUCLEAR REGULATORY COMMISSION j
l REFOkE THE ATOMIC RAFFTY AND 1TFNRTNG ROARD l In the Matter of )
)
4 GENERAL PUBLIC UTILITY NUCLEAR ) Docket No. 50-219-OLA
- CORPORATION )
)
j (Oyster Creek Nuclear Generating )
Station) )
i i AFFIDAVIT OF RONALD B. EATON IN SUPPORT OF THE NRC STAFF'S RESPONSE IN SUPPORT OF LICENSEE'S MOTION FOR
SUMMARY
DISPOSITION I, Ronald B. Eaton, being duly sworn, do hereby state as follows:
- 1. I am employed by the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation. My business address and telephone number are:
Ronald B. Eaton, Senior Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 (301) 415-3041
- 2. I am currently assigned to serve as the Senior Project Manager, Project Directorate I-2, Division of Reactor Projects I/II, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission.
- 3. I am experienced as a licensing examiner, inspector and NRC project manager of operating nuclear power plants. I am familiar with the regulations pertaining to the operation of l
1 i ; nuclear power plants and the staff's review of the amendment at issue. A summary of my l professional qualifications and experience is attached hereto (Exhibit 1) and is inwporated herein i
by reference.
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- 4. I have reviewed General Public Utility Nuclear Corporation's (GPUN) " Licensee's l
4 Motion for Summary Disposition," dated November 15,1996, and the supporting " Affidavit of John C. Fornicola" (Fornicola Aff.).
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- 5. Based on that review and my experience, it is my professional opinion that the l information and the events described by Mr. Fornicola in his affidavit are correct and historically 4
- factual. This opinion is reinforced by my comparison of the Oyster Creek dry storage issues with l those issues at other power plants. For example, Sacramento Municipal Utility District's l
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. (SMUD's) Rancho Seco facility had a Technical Specification similar to the Oyster Creek l l technical specification at issue and SMUD changed it for the same reason that GPUN did, whereas
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c Duke Power elected to adopt the same rationale for its Oconee facility that Consumers Power l
! adopted for its Palisades plant, namely, that the fuel in the transfer cask is in transit, not stored, l i
in retaining a similar technical specification unchanged. The Staff has not taken exception to l J
either approach.
- 6. Notwithstanding TS 5.3.1.B, GPUN moved a shield plug over irradiated spent fuel j i in a cask in returning irradiated fuel that had been shipped from West Valley to the spent fuel
! storage facility at Oyster Creek in 1984-85. Fornicola Aff.121.
- 7. Even though Palisades has such a TS, it did not need to amend it in order to l transport fuel to its ISFSI. Fornicola Aff.122.
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l 8. Documents cited in the Fornicola Affidavit with regard to the genesis and 1 Staff / Licensee interpretation of T.S. 5.3.1.B establish that the TS relates to stored fuel and not !
1 j to fuel being readied for transportation. ;
j 9. The Standard Technical Specification (STS) recommended by NUREG-0612 1 i
i specifically relates to stored fuel in racks and not to fuel being packaged for transport. Fornicola 1 ,
i Aff.1 19. I l
j 10. On the basis of these considerations, it is my professional opinion that Mr. I Fornicola correctly presented the historical perspective and established that Oyster Creek's TS l 5.3.1.B did not derive from a recommendation of NUREG-0612 but rather came about as a result J l
of an amendment request to install high density racks in the spent fuel storage facility at Oyster j Creek three years prior to the publication of NUREG-0612.
- 11. On October 5,1995, the Staff issued an amendment to Sacramento Municipal Utility District for Rancho Seco Technical Specification D 3.3. Prior to amendment, D 3.3 read I l
as follows:
No loads shall be handled over irradiated fuel assemblies stored in the spent fuel pool, except fuel assemblies, control components, and associated handling tools. The load shall not exceed the combined weight of one fuel assembly, control component and associated handling tool.
Amendment No.123 added language as follows:
This Limiting Condition for Operation permits fuel off-load activities as follows:
The dry shielded canister (DSC) top shield plug, the MP-187 cask lifting yoke and yoke extension, and gantry crane lower load block may be handled with the Turbine Building
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4 Gantry Crane over irradiated fuel assemblies that are in a DSC in the spent fuel pool. (A copy of the amendment package is attached as Exhibit 2).
- 12. The information set forth above is true and mrrect to the best of my knowledge and 4
belief.
Ronald B. Eaton
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Sworn and subscribed to before me this day of December,1996.
Notary Public .
l My commission expires-1 a
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EERICIT 1 i
- PROFESSIONAL QUALIFICATIONS
! OF
- RONALD B. EATON I have served as an NRC staff member since January 26,1981. From January 1981 to April
] 1990, I worked as an operator licensing examiner in the Operator Licensing Branch. I was certified to administer reactor operator and senior reactor operator written and operating i
examinations to candidates for licenses at Babcock & Wilcox (B&W), Combustion Engineering
, (CE), High Temperature Gas Reactor (HTGR) and all non-power reactors in the United States. !
i I also cenified NRC staff and contractors as examiners and audited their performance. From l April 1990 to present, I served as a Senior Project Manager for the Division of Reactor Projects j East. I served as NRC Project Manager for Pilgrim Nuclear Power Station until April 1996 and j I currently serve as NRC Project Manager for Oyster Creek Nuclear Generating Station.
t l Before joining the NRC 1 served 20 years in the U.S. Navy in their Nuclear Power Program.
- I served on 4 nuclear submarines and I nuclear aircraft carrier in the engineering departments.
I qualified on all engineering watch stations including Engineering Watch Supervisor, Propulsion ,
Plant Watch Officer, Engineering Officer of the Watch, and Engineering Duty Officer.
Additionally, I served as a Repair Officer and Duty Officer at the Norfolk Naval Shipyard supervising the repair and overhaul of nuclear submarines and surface ships.
I have completed the NRC technology courses for Combustion Engineering (CE) and Babcock
& Wilcoy (B&W) and numerous other related technical and non-technical courses at NRC and in the U.S. Navy. I attended Hobart College (1958-59) and Old Dominion University (1978-79).
I am an IIT staff resource and a principal author of the Incident Investigation Team (IIT) Report l on the Rancho Seco over-cooling event.
December 1996 l i
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EXHIBIT 2
- Octbber 5, 1995 l Mr. Steve J. Redeker, Manager Plant Closure & Decommissioning
- Sacramento Municipal Utility District P.0. Box 15830
- , Sacramento, California 95852-1830
SUBJECT:
ISSUANCE OF AMEN 0 MENT NO.123 TO FACILITY OPERATING LICENSE (POSSESSION ONLY) N0. DPR SACRAMENTO MUNICIPAL UTILITY DISTRICT
- i. (TAC M92618) i
Dear Mr. Redeker:
l The Commission has issued the enclosed Amendment No.123 to Facility Operating
- License (Possession Only) No. DPR-54 for the Rancho Seco Nuclear Generating j Station. This amendment is in response to your application (PA-191) dated i June 20, 1995, and as supplemented on August 14, 1995. The amendment would 4
permit SMUD to change the Fuel Storage Building crane load handling limits in j order to permit off-load of spent fuel assemblies from the spent fuel pool i (SFP) to the Rancho Seco Independent Spent Fuel Storage Installation.
t j A copy of our Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Reaister Notice.
Sincerely, j ORIGINAL SIGNED BY
! Richard F. Dudley, Senior Project Manager
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i Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation
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Docket No. 50-312
Enclosures:
- 1. Amendment No.123 to License No. DPR-54
- 2. Safety Evaluation cc w/ enclosures:
See next page DISTRIBUTION:
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,y S UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. seseH001 October 5, 1995 k ..... /
Mr. Steve J. Redeker, Manager Plant Closure & Decommissioning Sacramento Municipal Utility District P.O. Box 15830 Sacramento, California 95852-1830
SUBJECT:
ISSUANCE OF AMENDMENT NO.123 TO FACILITY OPERATING LICENSE (POSSESSION ONLY) NO. DPR SACRAMENTO MUNICIPAL UTILITY DISTRICT (TAC M92618)
Dear Mr. Redeker:
The Commission has issued the enclosed Amendment No.123 to Facility Operating License (Possession Only) No. DPR-54 for the Rancho Seco Nuclear Generating i Station. This amendment is in response to your application (PA-191) dated l June 20, 1995, and as supplemented on August 14, 1995. The amendment would permit SMUD to change the Fuel Storage Building crane load handling limits in order to permit off-load of spent fuel assemblies from the spent fuel pool (SFP) to the Rancho Seco Independent Spent Fuel Storage Installation.
A copy of our Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Reaister Notice.
Sincerely,
}r f l Richard F. Dudley, Senior roject Manager Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-312
Enclosures:
- 1. Amendment No.123 to License No. DPR-54
- 2. Safety Evaluation cc w/ enclosures:
See next page
~.- Mr. Steve J. Redeker Rancho Seco KE: lear Generating i .
Station 1 .
Docket No. 50-312 l cc:
4 Mr. Jan Schort, General Manager Mr. Thomas D. Murphy Sacramento Municinal Utility District Atomic Safety and Licensing Board
. 6201 S. Street Paral
- P. O. Box 15830 U.S. Nuclear Regulatory Commission j Sacramento, California 9h813 Washington, D.C. 20555 Thomas A. Baxter, Esq. Ms. Helen Hubbard
! Shaw, Pittman, Potts & Trowbridge P. O. Box 63
, 2300 N. Street, N.W. Sunol, California 94586 Washington, D.C. 20037 Ms. Dana Appling, General Counsel i,
Mr. Jerry Delezinski Sacramento Municipal Utility l Licensing Supervisor District i
Sacramento Municipal Utility District 6201 S. Street i
Rancho Seco Nuclear Station P. O. Box 15830 14440 Twin Cities Road Sacramento, California 95813
, Herald, California 95638-9799 James P. McGranery, Jr., Esq.
Mr. Robert B. Borsum, Licensing Dow, Lohnes & Albertson Representative Attornsys At Law
] B & W Nuclear Technologies Suite 500
! Nuclear Power Division 1255 Twenty-Third Street, N.W.
- 1700 Rockville Pike - Suite 525 Washington, D.C. 20037-1194 Rockville, Maryland 20852 Mr. Steve Shu
! Regional Administrator,' Region IV Radiologic Health Branch U.S. Nuclear Regulatory Commission State Department of Health Services j 611 Ryan Plaza Drive, Suite 400 P. O. Box 942732
- Arlington, Texas 76011-8064 Sacramento, California 94234 i Sacramento County Mr. James R. Shetler Board of Supervisors Deputy Assistant General 700 H. Street, Suite 2450 Manager, Operations Sacramento, California 95814 Rancho Seco Nuclear Generating Station I Mr. Leo Fassler 14440 Twin Cities Road Assistant General Manager and Herald, California 95638-9799 Chief Operations Officer Sacramento Municipal Utility District 6201 S. Street P. O. 15830 ,
Sacramento, California 95813 I Mr. Charles Bechhoefer, Chairman 1 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission l
Washington, D.C. 20555
\ Mr. Richard F. Cole Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555
ase uq g k UNITED STATES j
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30006-0001
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SACRAMENTO MUNICIPAL UTILITY DISTRICT i
I DOCKET NO. 50-312 RANCHO SECO NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE (POSSESSION ONLY)
Amendment No. 123 License No. DPR-54 l
- 1. The U.S. Nuclear Regulatory Comission (the Comission or the NRC) has
- found that
A. The application for amendment filed by the Sacramento Municipal Utility District (the licensee) dated June 20, 1995, and supplemented on August 14, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will be maintained in conformity with the application, i
the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by
- this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be cenducted in compliance with the Comission's regulations set forth in 10 CFR
- Chapter I; D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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. 2. Accordingly, the license is amended by changes to the Permanently Defueled Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License (Possession Only) No. DPR-54 is hereby amended to read as follows:
(2) Permanentiv Defueled Technical Specifications
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The Permanently Defueled Technical Specifications contained in Appendix A, as revised through Amendment No.123, are hereby incorporated in the license. Sacramento Municipal Utility District
- shall maintain the facility in accordance with the Permanently Defueled Technical Specifications.
1 l 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
.L M d Seymour H. Weiss, Director Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: october 5, 1995
ATTACHMENT TO LICENSE AMENDMENT NO. 123 FACILITY OPERATING LICENSE (POSSESSION ONLY) NO. DPR-54 DOCKET NO. 50-312 i
- Replace the following pages of the Appendix A Permanently Defueled Technical
- Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
i Remove Insert D3/4-4 D3/4-4 j BD3/4-6 803/4-6 l
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.. l RANCHO SECO UNIT 1
.. PERMANENTLY DEFUELED l ,.
TECHNICAL SPECIFICATIONS a
D3/4.3 FUFL STOR AGE BUTI DING LOAD HANDI ING LIMITS
[ T IMITING CONDITIONS FOR OPERATION 1
D3.3 No loads shall be handled over irradiated fuel assemblies stored in the spent fuel pool, except fuel assemblies, control components, and associated handling tools. The load shall
, not exceed the combined weight of one fuel assembly, control component and associated handling tool.
l l' This Limiting Condition for Operation permits fuel off-load activities as follows: The dry shielded canister (DSC) top shield plug, the MP-187 cask lifting yoke and yoke extension,
- and gantry crane lower load block may be handled with the Turbine Building Gantry Crane over irradiated fuel assemblies that are in a DSC in the spent fuel pool.
l j APPLICABILITY Whenever irradiated fuel assemblies are stored in the spent fuel pool l
ACTION:
) With the above requirements not met, place the fuel storage building fuel handling bridge,
- overhead crane, and gantry crane in a safe condition. ll l
S'URVEILLANCE REOUIREMENTS l
i D4.3.1 Perform a dead weight load test at the rated load on the fuel storage building fuel handling bridge and overhead crane within 7 days prior to initial use during fuel handling ,
j operations.
i j D4.3.2 Perform a FUNCTIONAL TEST of the fuel storage building fuel handling bridge
. interlocks within 7 days prior to the commencement of fuel handling operations and at least i once per 7 days thereafter during fuel handling operations.
j D4.3.3 Perform acceptance and verification testing of the MP-187 cask lifling yoke and yoke extension, and the DSC top shield plug support cables in accordance with ANSI N14.6-1986.
Proposed Amendment No.49',191' Amendment No. 123 D3/4-4
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RANCHO SECO UNIT 1 PERMANENTLY DEFUELED
! TECHNICAL SPECIFICATIONS D3/4 3 FUFL STORAGE BUII DING LOAD HANDLING LIMITS BASES l
Load handling in the Fuel Storage Building and the associated Permanently Defueled Technical Specification (PDTS) requirements are described and analyzed in Defueled Safety l Analysis Report (DSAR) Section 9.6.2.4, " Safety Provisions." The load handling limitations of PDTS D3/4.3 (1) meet the guidelines established in NUREG-0612 for the
- control of loads at nuclear power plants and (2) provide reasonable assurance that loads in excess of the combined weight of a fuel assembly, control component, and associated
. handling tool (i.e., the design basis load) will not, be handled over irradiated fuel.
Therefore, the potential for accidental heavy load drops over spent fuel is minimized.
Several years of operating experience has shown that these requirements are sufficient to '
protect plant personnel safety and public health and safety. .
l Restricting the movement ofloads over spent fuel assemblies to those loads that are within the design basis load provides reasonable assurance that in the event a load is dropped over spent fuel (1) the activity released will be limited to that contained in a single fuel assembly, and (2) the distortion of fuel in the spent fuel pool storage racks will not result in a critical
- array. These assumptions are consistent with the analysis for the Fuel Handling Accident addressed in DSAR Chapter 14," Accident Analysis."
i PDTS D3/4.3 facilitates the off-load of spent fuel assemblies from the spent fuel pool to the Rancho Seco Independent Spent Fuel Storage Installation (ISFSI). This specification allows the movement of the dry shielded canister (DSC) top shield plug (which weighs less than 7,000 pounds) with the Turbine Building Gantry Crane over spent fuel assemblies in a DSC in the spent fuel pool, using the four DSC top shield plug support cables, the MP-187 cask lifting yoke and yoke extension, and the gantry crane lower load block.
The load handling limit is based on the MP-187 cask lifting yoke and yoke extension, and
- the four cables used to support the DSC top shield plug being designed and tested in accordance with ANSI N14.6-1986. ANSI N14.6-1986 establishes the design, fabrication, testing, maintenance, and quklity assurance program requirements for (1) special lifting devices that lift containers holding radioactive material weighing more than 10,000 pounds and (2) container attachment members that affect the function and safety of the lift. Also, since NUREG 0612 considers the lower load block a heavy load, this gantry crane component, which was designed and tested to ANSI B30.2, is included in PDTS D3/4.3.
No credible accident scenario is associated with handling the DSC top shield plug or the MP 187 cask lifting yoke and yoke extension with the gantry crane, which includes the lower load block, over spent fuel assemblies that are in a cask /DSC in the spent fuel pool.
Also, the gantry crane is designed such that it can only handle loads over the SFP cask pit
, area of the spent fuel pool and can not move a load over the SFP fuel storage racks.
i BD3/g Amendment n . 123
i p %'% UNITED STATES l j
- .. g NUCLEAR REGULATORY CDMMISSION )
a f WASHINGTON, D.C. WOBMeM l j
j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
} SUPPORTING ANENDMENT N0.123T0 FACILITY OPERATING LICENSE i
j (POSSESSION ONLY) NO. DPR-54
! SACRAMENTO MUNICIPAL UTILITY DISTRICT
) RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 i
1.0 INTRODUCTION
By letter dated June 20, 1995, and supplemented on August 14, 1995, the !'
! Sacramento Municipal Utility District (SMUD or the licensee) proposed to amend
{ the Facility Operating License (Possession Only) No. DPR-54, by changes to the j Permanently Defueled Technical Specifications (PDTS) for the Rancho Seco i Nuclear Generating Station (Rancho Seco or the plant). The proposed amendment l I (PA-191) would permit the licensee to change the Fuel Storage Building crane l
! load handling limits in order to permit off-load of spent fuel assemblies from the spent fuel pool (SFP) to the Rancho Seco Independent Spent Fuel Storage !
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- Installation (ISFSI).
2.0 DISCUSSION AND SAFETY EVALUATION 2.1 Discussion I The plant is shut down and the reactor is permanently defueled and has not
- operated since June 7, 1989. All of the spent fuel from the reactor has been i transferred to the spent fuel pool in the fuel Storage Building, a seismic Category I structure.
In order to implement the changes described in Section 1.0 above, the licensee proposes to modify the fuel handling technical s~ pecification as follows:
PDTS D3/4.3 " Fuel Storage Building Load Handling Limits" A section is added that would enable the licensee to lift the dry shielded canister irradiated(DSC)l fue assemblies in a DSC that has been loaded with up totop shield twenty-four spent fuel assemblies in the deep pit area in the northwest section of the SFP. However, the load would not pass over spent fuel in the storage racks. Lifting hardware surveillance requirements are also added. The Bases section of this PDTS would be modified by adding the justification for the changes and references to the appropriate NRC guidance and industry standards.
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! 2.2 Reason for the Chance 1
i The current PDTS for Rancho Seco prohibit the handling of any load over spent i fuel that exceeds the weight of one fuel assembly and its associated lifting i tools. In order to remove spent fuel from the SFP to an onsite ISFSI, the
! fuel must be loaded into a DSC. The closure to the DSC, '.he top shield plug,
! must then be applied; but the plug weighs substantially more than a single i fuel assembly. Therefore, the licensee was required to apply for a technical i specification change in order to remove spent fuel to the ISFSI.
2.3 Safety Evaluation l The staff performed its safety evaluation by first determining the most i damaging potential accident that could occur during fuel handling operations.
Such an accident would be the highest drop of the heaviest load over the greatest number of spent fuel assemblies. Based on the configuration of the j spent fuel pool and fuel storage building, this accident would be the drop of j a 9 ton shield plug and lifting assembly from a height of about 30 feet onto j the top of a D3C loaded with 24 spent fuel assemblies.
t l The staff determined that such a drop is not credible. This determination is i based on our assessment of the very large factors of safety used in the design j of the licensee's crane and crane hardware. The crane upper and lower load j blocks will be disassembled and thoroughly inspected before use. The crane
- wire rope cable will be replaced with new cable. When installed on the crane
! in its 14 part (7 loop) configuration, the cable will have a safety factor of l 10 for a 2oad of 125 tons. The crane main hoist will be proof tested to i 125 percent of its rated 185 ton capacity and the cantilever section will be tested to 125 percent of its 130 ton capacity. The lifting assemblies and i hardware will be tested in accordance with and will meet the requirements of
! ANSI Standard N14.6. Each of the four sling and turnbuckle assemblies has j sufficient capacity to support the total weight of the shield plug.
1 i Additionally, crane operators handling heavy loads in the vicinity of the i spent fuel pool will be qualified in accordance with ANSI Stanrhrd B30.2,
! Chapter 2-3, for cab-operated cranes. Operators will also anpiete an l extensive training program including classroom instructiv, .nd oral or j written examination, and a practical examination requiring the lift and j precise positioning of a heavy load in a laydown area.
1 l Although, the greatest potential shield plug drop height is about 30 feet,
} the licensee informed the staff in an August 23, 1995, teleconference that the planned load path to be used in lowering the shield plug onte-the DSC will
! limit the greatest actual drop height to about three feet. Such an event
! would not result in any damage to the spent fuel. The staff also determined
! that the crane main hook has interlocks and mechanical steps that prevent j loads from DSC loading operations which exceed the weight of a single fuel j assembly from passing over spent fuel in the storage racks.
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! Based on the above considerations the staff has concluded that the proposed changes to PDTS D3/4.3 are acceptable.
3.0 STATE CONSULTATION
l In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of this amendment. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
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The amendment changcr Mquirements with respect to installation or use of j facility components located within the restricted area as defined in 10 CFR l
- Part 20 and changes surveillance requirements. The NRC staff has determined i that the amendment involves no significant increase in the amounts, and no i significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative
- occupational radiation exposure. The Commission has previously issued a l proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (60 FR 45184). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above,
- that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such
, activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common
, defense and security or to the health and safety of the public.
1 Principal Contributor: Morton B. Fairtile
- Date
- october 5, 1995 4
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