ML20206F949
| ML20206F949 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 05/05/1999 |
| From: | Marian Zobler NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| CON-#299-20343 LA, NUDOCS 9905070018 | |
| Download: ML20206F949 (42) | |
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}-zo3h 00CKETED Mh3[1599 l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
'99 MY -6 P3 :07 BEFORE THE ATOMIC SAFETY AND LICENSING B ARD '
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) 1 In the Matter of ) Docket No. 50400-LA l
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- CAROLINA POWER & LIGHT CO. )
)
(Shearon Harris Nuclear Power Plant) )
NRC STAFF'S RESPONSE TO ORANGE COUNTY'S l SUPPLEMENTAL PETITION TO INTERVENE ,
l INTRODUCTION Pursuant to a " Memorandum and Order (Initial Prehearing Order)," issued on l l
February 24,1999, by the Atomic Safety and Licensing Board (Board) designated in the i
above-captioned proceeding and 10 C.F.R. I 2.714(c), the staff of the Nuclear Regulatory j Commission (Staff) hereby responds to " Orange County's Supplemental Petition to Intervene" (Supplement). As discussed below, none of Orange County's (Petitioner) proposed contentions are admissible, therefore, the Petitioner's February 12,1999 Request j for Hearing and Petition to Intervene should be denied.
BACKGROUND ;
l On December 23,1998, Carolina Power & Light Company (Licensee) filed an application for a license amendment pursuant to 10 C.F.R. I 50.90 for the Shearon Harris Nuclear Power Plant (Harris). Letter to United States Nuclear Regulatory Commission from l
James Scarola, Vice President, Harris Nuclear Plant, Cr.rolina Power & Light Co., i December 23,1998 (Application). On January 13,1999, the NRC published Carolina l l
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Power & Light; Notice of Consideration ofissuance ofAmendment to Facility Operating License, Proposed No Sigmficant Hazards Consideration Determination and Opportunity for a Hearing (Notice). 64 Fed. Reg. 2237 (1999). The Notice provided that the proposed amendment would support a modification to Harris to increase the spent fuel storage capacity by adding rack modules to two spent fuel pools (SFP) (pools "C" and "D*). Id. at 2238. The proposed amendment would also support the placement of those pools into service. Id.
On February 12,1999, the Petitioner filed a request for a hearing. " Orange County's 1
Request for Hearing and Petition to Intervene." On March 1,1999 and March 4,1999, pursuant to the Board's Initial Prehearing Order, the Licensee and the Staff filed, respectively, answers to the Petitioner's Petition in which the issues of standing and aspects were addressed. " Applicant's Answer to BCOC's Request for Hearing and Petitioner to 1
I Intervene," March 1,1999, "NRC Staff's Answer to Orange County's Request for Hearing and Petition to Intervene," March 4,1999.
On April 5,1999, pursuant to the Board's Initial Prehearing Order, the Petitioner submitted its Supplement which contained eight contentions. As discussed below, none of 1 the Petitioner's contentions satisfy the Commission's requirements for the admission of contentions. The Petitioner's request for a hearing and petition to intervene should, therefore, be denied. . l l
DISCUSSION
- 1. Leral Standards for the Admission of Contentigns To gain admission as a party, a petitioner for intervention, in addition to establishing standing and raising an aspect within the scope of the proceeding, must submit at least one
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valid contention that meets the requirements of 10 C.F.R. l 2.714.10 C.F.R. l 2.714(b)(1);
Duke Energy Corporation, (Oconee Nuclear Station, Units 1,2, and 3), CLI-99-11, slip op.
at 5 (April 15,1999); Yankee Atomic Electric Company (Yankee Nuclear Power Station),
CLI-96-7,43 NRC 235,248 (1996). For a contention to be admitted, it must meet the standard set fonh in 10 C.F.R. I 2.714(b)(2), which provides that each contention must consist of "a specific state nent of the issue oflaw or fact to be raised or controverted" and must be accompanied by:
(i) A brief explanation of the bases of the contention;
. (ii) A concise statement of the alleged facts or expert opinion which supports the contention . . . together with references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expen opinion;
' (iii) Sufficient information . . . to show that a genuine dispute exists with the
. applicant on a material issue oflaw or fact.
i 10 C.F.R. I 2.714(b)(2). The failure of a contention to comply with any one of these requirements is grounds for dismissing the contention.10 C.F.R. I 2.714(d)(2)(i); Arizona Public Service Company (Palo Verde Nuclear Generating Station, Units 1, 2, and 3),
CLI-91-12,34 NRC 149,155-56 (1991).
In order for a dispute to involve a material issue of law or fact, its resolution must make a difference in the outcome of the proceeding. Oconec, CLI-99-11, slip op. at 6, citing Final Rule, Rules ofPracticefor Domestic Licensing Proceedings -- Procedural Changes in the Hearing Process, 54 Fed. Reg. 33,168, 33,172 (1989). See also 10 C.F.R.
- i 2.714(d)(2)(ii)(a contention must also be dismissed where the " contention,ifproven, would
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be of no consequence . . . because it would not entitle [the) petitioner to relief."). Moreover, l
I contentions that are not supported by some alleged fact or facts should not be admitted nor j should the full adjudicatory hearing process be triggered by contentions that lack a factual and legal foundation. Oconce, CLI-99-11 slip op. at 6,7 citing 54 Fed. Reg. at 33,170.
- 2. Proposed Contentions -
As demonstrated below, none of the contentions proffered by the Petitioners meet the standards discussed above.
GROUPI: TECHNICAL CONTENTIONS:
Contention 1: Inadequate Emergency Core Cooling and Residual Heat Removal In order to cool spent fuel storage pools C and D, CP&L proposes to rely on the Unit 1 Component Cooling Water ("CCW") system, coupled with administrative measures to ensure that the heat load from the pools does not overtax the CCW system. CP&L's reliance on the Unit 1 CCW system and I administrative measures for cooling spent fuel storage pools C and D will unduly compromise the effectiveness of the residual heat removal ("RHR") )
system and the Emergency Core Cooling Systems ("ECCS") for the Shearon Harris plant, such that the plant will not comply with Criteria 34 and 35 of Appendix A to 10 C.F.R. Part 50.
Supplement at 4. In support of this contention, the Petitioner asserts that the Licensee's I proposed design modification, to tie-in the Unit I component cooling water (CCW) system to the heat exchangers of the SFP C and D Fuel Pool Cooling and Cleanup System (FPCCS),
raises issues regarding the Licensee's ability to comply with General Design Criteria (GDC) 34 and 35, residual heat removal and emergency core cooling capability. Id. at 5.
Specifically, the Petitioner asserts that during the recirculation phase of a loss-of-coolant accident (LOCA), the Licensee may be obliged to require its operators to divert some of the CCW system flow from the RHR heat exchangers in order to cool SFP C and D. Id. The e
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. !' Petitioner raises six concerns it contends support its contention and need to be resolved l
before it can be determined that the changes being proposed will comply with GDC 34 and j 1
- 35. Id. at 6- 7.
The Petitioner first claims that based on design information in the Final Safety ,
1 Analysis Report (FSAR) for Harris, the CCW system, currently, without the additional heat j l
load from SFP C and D, would be unable to accommodate a design-basis LOCA. Id. i According to the Petitioner, the CCW system has two heat exchangers each with a design i transfer rate of 50 million Btu / hour.8 Id. The Petitioner contends, however, that during the recirculation phase of a design-basis LOCA the estimated beat load on the CCW system is 160 million BTU / hour. Id. .
j The Petitioner's first concern, however, is outside the scope of this proceeding. The issue the Petitioner raises involves the ability of the CCW to currently accommodate a LOCA. Denial of the Application would have no affect on the Petitioner's concem. Thus, it fails to demonstrate that a genuine dispute exists with the Licensee on a material issue of 8
The Petitioner references Table 9.2.2-1, Amendment No. 27 of the Harris FSAR.
Id. at n. I1. The current Harris FSAR, however, has been updated to Amendment No. 48.
According to this current version, Table 9.2.2-1 indicates that the design heat transfer for the Component Cooling Heat Exchanger is 50.5 million Btu /hr. See Harris FSAR at 9.2.2-10, Table 9.2.2-1, copy attached here to as Attachment 1. In addition, the Petitioner also
- references Table 9.2.1-3, Amendment No.15 to the FSAR. Supplement at 7, and n.12. As discussed below, the Petitioner states that based on this table the maximum heat load to be extracted from the CCW system by the service water system (SWS) during a LOCA is 160 million Btu / hour. Id. Table 9.2.1-3, however, has also been updated through Amendment No. 48. A copy of Table 9.2.1-3 is attached hereto as Attachment 2. The current table establishes that the maximum heat load to be extracted from the CCW system is 272.6 million Btu /hr.
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i law or fa::t. See 10 C.F.R. l .2714(b)(2)(iii). Moreover, the Petitioner's concern is based on
- a misunderstanding of the Harris FSAR. !
The Petitioners assen that the heat load on the CCW system exceeds the system's l
ability to transfer heat. Supplement at 7. Table 9.2.2-1, however, provides the heat load on the two heat exchangers of the CCW system under the conditions listed in the table, e.g.,
with the service water supply temperature of 95'F,is 50.5 million Btu /hr. Moreover, Table 9.2.1-3 indicates that during the recirculation phase of a LOCA, the CCW system will eject a heat load of 160 million Btu /hr (or as indicated in the most recent version of the FSAR, I
272.6 million Btu /hr) to the Service Water System (SWS).2 Thus,even underthe Petitioner's assumptions, the FSAR tables referenced by the Petitioner indicate that a heat load of 100 million Bru/hr (50 million Btu /hr for each heat exchangers) will be going into the CCW system, during a LOCA, but that the CCW system is capable of ejecting 160 million Btu /hr into the SWS. These tables, therefore, indicate that the CCW system is, in fact, capable of accommodating the heat load from a LOCA. Thus, the Petitioners' first assertion regarding the present capability of the Harris CCW system is without basis and does not suppon the admission of Contention 1. See 10 C.F.R. I 2.714(b)(2)(ii), (iii).
The Petitioner next raises a concem that the Application, Enclosure 9 did not address the time dependence of the CCW system heat load during a LOCA. See Supplement at 7.
According to the Petitioner, the Licensee's analysis must demonstrate that the CCW system has sufficient margin to accommodate both the RHR system and fuel pool heat loads over 2
In fact, the Petitioner recognizes that table 9.2.1-3, provides the estimated maximum her, load to be extracted from the CCW system by the SWS during a LOCA.
See Supplement at 7 n.12.
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7 time, during the LOCA event and subsequently. Id. As discussed below, this concern lacks specificity and basis.
l According to Enclosure 9, the Licensee performed an analysis of the capacity of the 1
CCW system to accommodate the additional heat load of 1.0 MBtu/hr that would be added to the system by the addition of SFP C and D, plus a 6% modeling uncertainty and degraded l l
IST pump performance. Enclosure 9 at 2-3. According to the Licensee, the analysis l indicated that the CCW system did have such capacity, but that the minimum flow to the I
RHR heat exchanger would change from current requirements. Id. In making this determination, the Licensee assumed a heat rejection rate of 111.1 million Btu /hr, a rate consisteatwithexistingpost LOCAcontainmentpressure/temperaturecalculations,suchthat no changa in containment heat removal was prescribed. Id. at 3. Based on this rate, the Licensee then calculated the necessary minimum CCW system flow rate, which was less than specified in the initial safety evaluation report for Hanis. Id. at 3-4. Thus, the Licensee's
, analysis did consider the heat load on the CCW system after a LOCA using existing calculations. The Petitioner's assertion fails to explain why the Licensee's analysis is incorrect. Thus, this second issues fails to provide an adequate basis for Contention 1. See 10 C.F.R. I 2.714(b)(2)(ii),(iii).
The Petitioner also raises the concern that the Licensee failed to address the sensitivity of the CCW and RHR systems performance to factors that may degrade ,
performance frorn nominallevels. Id. According to the Petitioner, relevant factors would include heat exchanger fouling and plugging. Id at 8. This third concern also lacks specificity and basis. As discussed above, the Licensee's analysis focused on the change in l
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the minimum flow to the RHR system from the CCW system, using current assumptions regarding the heat load in the event of a LOCA. This analysis was based on the additional heat load from the SFP C and D, and a 6% modeling uncertainty. Enclosure 9 at 3. The Petitioner fails to explain why the potential for heat exchanger plugging or fouling indicates that the Licensee's analysis is deficient.' This concern, therefore, fails to provide an adequate basis for Contention 1. See 10 C.F.R. I 2.714(b)(ii).
The Petitioner also claims that Enclosure 9 is deficient because the Licensee failed to address the potential for the failure of administrative measures. Supplement at 8. The Petitioner fears that the heat load could be exceeded as a result of human errors. Id. As an initial matter, the heat load limits on SFP C and D will be placed in the Licensee's technical specifications, which are incorporated into the Harris operating license. Furthermore, the Petitioner's concern lacks specificity and basis. The Petitioner, for example, fails to explain what the possible errors would be and what the consequences of such errors would be. Thus this basis does not support the admission of Contention 1. See 10 C.F.R. I 2.714(b)(2)(ii).
The Petitioner also argues that the potential for operator en or during a LOCA would be increased by the Licensee's proposed modification because operators would be required to divent some of the CCW system flow from the RHR heat exchangers in order to meet the cooling needs of SFP C and D while also serving other safety functions. Supplement at 8.
8 The Petitioner notes, as support for its concern, that the Licensee recognized that changes to the CCW system could involve changes in design assumptions including fouling l factors and tube plugging limits and cites to a viewgraph provided during a meeting. Id. A !
review of the viewgraph cited, however, indicates that the Licensee was only considering the possibility of some changes in design assumptions and does not establish that the Licencee's analysis, as presented in Enclosure 9, is deficient.
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This concem, as with the one above, is similarly without basis or specificity. Moreover, as t
indicated in Enclosure 9, current operating procedures direct operators to restore sufficient 4
CCW flow from one CCW train to maintain the temperature at the spent fuel pool during accident conditions. Enclosure 9 at 2. Denying the Licensee's proposed modification would not alter these procedures. Thus, in addition to lacking basis and specificity, the Petitioner's concern fails to demonstrate that a genuine dispute exists with the Licensee on a material issue of law or fact. 10 C.F.R. I 2.714(b)(2)(iii). Thus, this concem does not provide an adequate basis for Contention 1. .
Lastly, the Petitioner claims that the Licensee has not addressed the ability of the onsite and offsite electrical power systems to support residual heat removal, emergency core cooling and other safety functions while also meeting the burden of cooling the spent fuel l
pools. Supplement at 9. The Petitioner's last concern lacks both specificity and basis. The Petitioner claims that the Licensee must address this issue, but fails to explain why. The affidavit of Dr. Gordon Thompson, attached as Exhibit 2 to the Suppleme 4.t, fails to provide the requisite basis and specificity. Other than mentioning that emergency diesel generators would be needed to cool SFPs C and D in addition to pools A and B in the event of a loss of offsite power, Dr. Thompson does not provide any explanation of why the current emergency diesel generators could not provide the necessary power. Thus, the Petitioner' final concern fails to support the admission of Contention 1. See 10 C.F.R. I 2.714(b)(2)(ii).
The Petitioner also asserts, presumably as an additional basis for Contention 1, that the Licensee considered, but rejected, measures that it could have implemented to ensure compliance with GDC 34 and 35. Supplement at 9-10. The proper focus of this proceeding,
.- l however, is those measures that the Licensee did, in fact, propose in its Application. See I
10 C.F.R. i 2.714(b)(2)(iii); See also The Curators ofthe University ofMissouri, CLI-95-8, i
41 NRC 386,395-96 (1995). Thus, the Petitioner's final basis fails to support the l
admission of Contention 1. I As discussed above, none of the bases the Petitioner offers in support of Contention 1 1
meet the requirements of 10 C.F.R. I 2.714(b)(2). Thus, Contention 1 should be dismissed. I Contention 2: Inadequate Criticality Prevention
. I Storage of pressurized water reactor ("PWR") spent fuel in pools C and D at the Harris plant, in the manner proposed in CP&L's license amendment 4 application, would violate Criterion 62 of the General Design Criteria -
("GDC") set forth in Part 50, Appendix A. GDC 62 requires that:" Criticality )
in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations."
In violation of GDC 62, CP&L proposes to prevent criticality of PWR fuel in pools C and D by employing administrative measures which limit the combination of burnup and enrichment for PWR fuel assemblies that are placed in those pools. This proposed reliance on administrative measures rather than physical systems or processes is inconsistent with GDC 62.
Supplement at 10. In support of Contention 2, the Petitioner notes that in SFP A and B criticality is preversted by maintaining a certain distance between the fuel assemblies and surrounding each fuel assembly with a neutron-absorbing material. Id. The Petitioner asserts, however, that since the proposed fuel storage racks for SFP C and D would allow for closer placement of the spent fuel assemblies than in pools A and B, the Licensee has proposed to rely on administrative measures to protect against criticality. Id. at 11. This, the Petitioner asserts is in violation of GDC 62 which requires that criticality be prevented by physical systems or processes. Id.
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The Petitioner's concem is unsupported and, thus, lacking in specificity and basis.
As indicated in the Application, the Licensee has evaluated the K, for a certain type of fuel l l assembly containing a certain maximum initial enrichment and concluded that the racks can safely accommodate, without borated water, fuel of various initial enrichments and discharge fuel burnup, provided this combination falls within the acceptable domain indicated in Figure 4.2.1 of Enclosure 7.d Application, Enclosure 7, Rev. 3 at 4 4 4. The Licensee states that the bumup criteria will be implemented by appropriate administrative procedures to ensure verified bumup prior to fuel transfer into SFP C or D. Id. at 4-4.
Contrary to the Petitioner's assertion, GDC 62 does not preclude the Licensee from limiting the bumup of the spent fuel to be placed in the spent fuel pool. GDC 62 provides that criticality shall be prevented by physical systems or processes. The burnup of fuel, as well as its enrichment value, is, itself, a physical process which affects criticality. Based on the bumup of the fuel, a Licensee may use, inter alia, fuel storage rack designs that will prevent criticality of the fuel pool by the use of geometrically safe configurations. Thus, the Licensee's proposal to take credit for fuel burnup limits as a means to maintain suberiticality of the spent fuel pool is consistent with GDC 62.
Further, although not a regulation, Staff guidance indicates that one of the methods of compliance with GDC 62 is limiting the burnup value of the fuel to be placed in the spent d
A spent fuel pool will become critical ifit is capable of supporting a neutron chain reaction. See Consumers Power Co (Big Rock Point Nuclear Plant), ALAB-725,17 NRC 562, 564 at n. 2 (1983). This condition is expressed in term of the " effective neutron multiplication factor"(K,,). Id. When the K,,is equal to 1.0 the pool is critical. Id.
4 fuel pool.s As indicated in Draft Regulatory Guide 1.13 (RG 1.13), spent fuel storage rack design may be based on the most reactive fuel assembly to be stored based on a minimum confirmed burnup. RG 1.13 at 1.13-12, copy attached hereto as Attachment 3. Where credit in the fuel rack design is taken for bumup, a minimum fuel assembly burnup must be established. Id. at 1.13-13. RG 1.13 further provides that the nuclear criticality safety analysis should demonstrate that each spent fuel storage facility system is at least 5 %
subcritical (K,,r not to exceed 0.95). Id. at 1.13-9. The analysis should identify the spent fuel assembly characteristic upon which suberiticality depends. Id. Nothing in the Petitioner's Supplement indicates that either the Licensee's proposal or the Staff's guidance is not in conformance with GDC 62 or that the subcriticality of the SFP will not be l maintained. This basis, therefore, fails to support the admission of Contention 2.
The Petitioner further asserts, as a basis for Contention 2, even if the Licensee could rely on RG 1.13, the Licensee's proposed use of administrative measures is not in conformance with the RG 1.13. Id. at 12-13. According to the Petitioner, although some i
parts of RG 1.13 indicate that a licensee can take credit for burnup, other parts of the Guide l
" clearly proscribed such activity." Id. at 12. The Petitioner then references a section in '
l RG 1.13 that provides that a nuclear criticality analysis should demonstrate that criticality i could not occur without at least two unlikely, independent, and concurring faihues or 8
The Commission has recognized that general design criteria arejust that, general criteria. A Licensee is free to propose a variety of methods for conforming with them, but t Staff guidance has been issued to provide a Licensee with at least one acceptable way to j meet them. PetitionforEmergency and Remedial Action, CLI-78-6,7 NRC 406-07 (1978). l Although not controlling, the Staff's guidance is entitled to substantial weight. See Big l Rock, ALAB-725,17 NRC at 568.
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operating limit violations. Id. at 13 citing RG 1.13 at 1.13-9. The Petitioner contends that because ' misplacement of a fuel assembly could cause criticality, the Licensee's j administrative controls do not satisfy RG 1.13. See id.
The Petitioner, however, fails to explain how the Licensee's proposed use of administrative controls is in contradiction with this section of RG 1.13. As discussed above, '
RG 1.13 indicates that the nuclear criticality analysis should be performed, assuming such an event occurs, despite the Licensee's administrative controls.' RG 1.13 does not state that a licensee cannot take credit for burnup. Furthermore, the Petitioner fails to provide any support for its assumption that misplacement of a fuel assembly will,in fact, cause criticality -
in the spent fuel pools. The Petitioner's concem, therefore, fails to provide an adequate basis to support Contention 2. As discussed above, neither of the bases the Petitioner offers provide adequate support to admit Contention 2. Contention 2, therefore, must be dismissed.
10 C.F.R. I 2.714(b)(2).
Contention 3: Inadequate Quality Assurance CP&L's proposal to provide cooling of pools C and D by relying upon the use of previously completed portions of the Unit 2 Fuel Pool Cooling and Cleanup Systems ("FPCCS") and the Unit 2 Component Cooling Water System (CCWS) does not satisfy the quality assurance criteria of 10 C.F.R.
Appendix B. In particular, CP&L fails to satisfy Criterion XIII, regarding Handling, Storage and Shipping, because it has not demonstrated that the piping and equipment have been stored and preserved since the time of completion in a manner that prevents damage or deterioration. CP&L also fails to satisfy Criterion XVI, regarding Corrective Action, because it has
- On April 29,1999, the Staff issued a letter requesting additional information regarding the Licensee's analysis with respect to the possible scenario where a fuel assembly outside the acceptable range for burnup is placed into spent fuel pool C or D. The Staff will evaluate the Licensee's response to determine if the requisite demonstration has been made.
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l' failed to show that it has instituted appropriate measures to correct any ,
l damage or deterioration that has occurred since the piping and equipment were installed. In addition, CP&L fails to satisfy Criterion XVII, regarding Quality Assurance Records, because it has not kept records needed to show that all applicable quality assurance requirements are satisfied. Moreover, the
" Alternative Plan" submitted by CP&L with its license amendment application falls to satisfy the requirements of 10 C.F.R. I 50.55a for an exception to these requirements. Finally, CP&L's proposal to postpone inspection of embedded portions of previously completed the piping until after issuance of the license amendment must be rejected as insufficient to demonstrate compliance with 10 C.F.R. I 50.55a. The inspection proposed by CP&L is necessary to its showing under i 50.55a that the alternative proposed by the Licensee would provide "an acceptable level of quality and safety." This showing must be made hsfgIn the license amendment is issued, not afterwards. .
Supplement at 13-14. 'In support of Contention 3, the Petitioner makes several assertions attempting to show that the Licensee's proposed alternative plan does not meet the quality assurance requirements of 10 C.F.R. Part 50, Appendix B. Specifically, the Petitioner contends that in order to be allowed to use previously completed piping and equipment, the Licensee must satisfy Criteria XIII, XVI and XVII of 10 C.F.R. Part 50, Appendix B. Id. at
- 16. This, according to the Petitioner, the Licensee has failed to do. The Petitioner also asserts that the Licensee has failed to address the issue of the degradation of piping and other equipment over time. See id. at 17-18.
None of the Petitioner's concem raised in support of Contention 3 indicate that the Licensee's proposed alternative plan to meet the requirements of 10 C.F.R I 50.55a is deficient. Pursuant to 10 C.F.R. I 50.55a(a)(3) alternatives to the requirements may be authorized where a licensee demonstrates that the alternative plan would provide an acceptable level of quality and safety or if compliance with the requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and u _ _
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l safety. The Licensee's proposed alternative plan, therefore, must be judged against this ,
1 standard and not the requirements of 10 C.F.R. Part 50, Appendix B. Thus, to the extent that the Petitioner's concerns relate to compliance with Appendix B, they are outside the scope i of these proceeding and thus, do not demonstrate that a genuine dispute exists with the !
Licensee on a genuine issue oflaw or fact. 10 C.F.R. I 2.714(b)(2)(iii). !
The Petitioner further claims, in support of Contention 3, that the Licensee's proposed inspection of embedded piping will not provide an acceptable level of quality and i safety. Supplement at 18. Specifically, the Licensee's proposed use of remote camera inspection for the inspection of one third of the embedded field welds will not likely yield a clear answer regarding the weld quality. Id. According to the Petitioner, a remote camera inspection can provide only limited information about weld quality and cannot provide 2e level of quality assurance that is available from non-destructive testing (NDE) and that such inspections should be performed before the plan is approved. Id. The Petitioner also claims that the Licensee's use of circumstantial evidence to confirm that the remaining 2/3 of the welds were actually inspected is insufficient to show an acceptable level of quality and safety. Id. at 18-19.
The Petitioner's concerns, however, lack an adequate basis. In the Application, Enclosure 8, the Licensee describes its alternative plan of which remote camera testing of those welds embedded in concrete is a part. See Application, Enclosure 8 at 8-11.
Specifically, the Licensee states that in order to address the issue of NDE reco' rds, accessible field welds have been subjected to reinspection and NDE and found to be acceptable. Id. at
, 10. The remote camera inspection will be performed for those welds that are embedded in
l concrete. Id. Other than asserting that the Licensee's plan to rely circumstantial upon evidence and remote camera inspection is not sufficient, the Petitioner fails to provide any explanation of or basis for its assertions that the Licensee's alternative plan does not meet the criteria of 10 C.F.R. I 50.55a(a)(3). Thus, the Petitioner fails to provide an adequate basis for Contention 3. It should, therefore, be dismissed. See 10 C.F.R. I 2.714(b)(2)(ii), (iii).
GROUP II: ENVIRONMENTAL CONTENTIONS Contention 4: Proposed License Amendment Not Exempt From NEPA
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CP&L errs in claiming that the proposed license amendment is exempt from NEPA under 10 C.F.R. I 51.22.
In support of Contention 4, the Petitioner claims that, assuming the Licensee intends to rely on 10 C.F.R. I 51.22(c)(9) as a categorical exclusion , the Licensee has failed to meet j the requirements of that exclusion.
In conformance with its prior practice, the Staff has determined to prepare an environmental assessment (EA) to assess the environmental impacts of the Licensee's Application. Thus, the Petitioner's Contention 4 is moot and should be dismissed.
Contention 5: Environmental Impact Statement Required The proposed license amendment is not supported by an Environmental Impact Statement ("EIS"),in violation of NEPA and NRC's implementing regulations. An EIS should examine the effects of the proposed license amendment on the probability and consequence of accidents at the Harris plant. As required by NEPA and Commission policy, it should also examine the costs and benefits of the proposed action in comparison to various alternatives, including Severe Accident Design Mitigation Alternatives and the alternative of dry cask storage.
Supplement at 35. Contention 5 is premature. Contentions asserting that an EIS should be prepared before the Staff's environmental review has been completed are premature. See
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(1981)("It is unwise, if not improper, to decide without the record support provided by the staff's environmental review, whether a given action significantly afferts the environment.").
Contentions regarding the adequacy of the Staff's environmental review, including whether an EIS should be prepared if it is determined that the proposed action has no significant impact, should await the issuance of the EA. Contention 5 should, therefore, be dismissed.
Contention 6: Scope of EIS Should Include Brunswick and Robinson Storage The EIS for the proposed license amendment should include within its scope the storage of spent fuel from the Brunswick and Robinson nuclear power plants.
Supplement at 37. The Petitioner bases this contention on the assertion that the purpose of the proposed expansion is not only to store spent fuel generated by Harris, but also to j accommodate fuel from Brunswick and Robinson. .Id. As with Contention 5, this contention is premature and should be dismissed. In addition, it should be dismissed because the proposed action does not involve the authorization to receive spent fuel from Brunswick and Robinson. Since Harris is already authorized to receive and store spent fuel from Brunswick and Robinson, this contention is not relevant to this proceeding. See Virginia Electric Co. (North Anna Power Station, Units 1 and 2), ALAB-790,20 NRC 1450, 1453-54 (1984)(Ruling that an amendment to permit the receipt and storage of spent fuel at the North Anna facility from the Surrey facility has no bearing on a separate amendment approving the expansion of the spent fuel pool at the North Anna facility).
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Contention 7: Even if no EIS required, Environmental Assessment Required Even if the Licensing Board finds that no EIS is required, it must order the preparation of an EA.
Supplement at 38. In support of this contention, the Petitioner states that even if an E5 is
- not required, the NRC must still prepare an EA to evaluated unresolved conflicts concerning alternative uses of available resources, as required by section 102(2)(E) of NEPA. Id.
Specifically, the Petitioner asserts that there are unresolved conflicts regarding the appropriate use of space and resources for the storage of spent fuel generated by the Harris, Brunswick, and Robinson facilities. Id. -
As discussed above, the Staff has decided to prepare an EA. Thus, Contention 7 is moot and should be dismissed. Further, the receipt and storage of spent fuel from Brunswick and Robinson is already authorized and is, therefore, outside the scope of the proposed action. See supra Contention 6.
Contention 8: Discretionary EIS Warranted l Even if the Licensing Board determines that an EIS is not required under NEPA and 10 C.F.R. I 51.20(a), the Board should nevertheless require an EIS as an exercise ofits discretion as permitted by 10 C.F.R. Il 51.20(b)(14) and 51.22(b).
i Supplement at 39. As a basis for this contention, the Petitioner contends that special j 1
circumstances exist warranting a discretionary EIS because the proposed action involves ;
l unresolved conflicts concerning altemative use of available resources within the meaning of )
I section 102(2)(E) of NEPA. Id. These conflicts arise, according to the Petitioner, from the ;
fact that the Licensee intends to store spent fuel generated by three difference reactors,
)
Harris, Robinson, and Brunswick. Id. j em . - + % pee w e e m .. e++.%.~4e - m .. A .- e=. .,-...w=c -e a
f-L .
t l
None of the Petitioner's claims indicate that the an EIS should be prepared as a matter of discretion.' As characterized by the Petitioner, these unresolved conflicts arise from the ;
storage of spent fuel from the Brunswick and Robinson facilities. Supplement at 39. 'Ibe storage of spent fuel from Robinson and Brunswick is already authorized by the Harris license. Thus, the consideration of the environmental impacts of the storage of spent fuel fa:n Robinson and Brunswick is not warranted. See Nonh Anna, ALAB-790,20 NRC at 1453-54.
The Petitioner also states that another reason to prepare an EIS is that the Ucensee's proposal appears to be in conflict with the Commission's Waste Confidence decision. See Supplement at 40-41, citing 10 C.F.R. I 51.23. According to 10 C.F.R. I 51.23, the Commission stated its belief that there is reasonable assurance that at least one mined geologic repository will be available within the first quarter of the twenty-first century.
10 C.F.R. I 51.23(a). The Ucensee, however, stated in the Application that DOE spent fuel storage facilities are not available and are not expected to be available for the foreseeable future. Supplement at 40-41 citing Application, Enclosure 1 at 1. Thus, the Petitioner contends, the Licensee's Application is in conflict with 10 C.F.R. I 51.23. Id. This concern also does not constitute a special circumstance warranting the preparation of a discretionary EIS. According to the Application, the Licensee anticipates a need for an expansion in spent 7
The question of whether the Staff should prepare an EIS pursuant to 10 C.F.R.
Il 51.20(b)(14) and 51.22(b) is a matter of discretion. The Board does not have the authority to direct the Staff, as a matter of discretion, to prepare an EIS. See Carolina Power and Light Co. (Shearon Harris Nuclear Power Plants, Units 1,2,3, and 4),CLI-80-12, 11 NRC 514,516-17 (1980)("[T]he Boards do not direct the staff in performance of their administrative functions.").
5 fuel storage capacity by the year 2000. Application, Enclosure 1 at 1. 10 C.F.R. I 51.23 provides that at least one mined geologic repository will be available within the first quarter of the twenty-first century, but not necessarily by the year 2000. See 10 C.F.R. I 51.23.
Thus, there is no conflict between the basis of the Licensee's proposal and the Commission's regulation. In any event,10.C.F.R. I 51.23 only relates to the environmental impacts of =
spent fuel storage beyond the operating term of a reactor.10 C.F.R. I 51.23. The Petitioner, thus, falls to demonstrate that special circumstances exist warranting a discretionary EIS.
CONCLUSION As set forth above, none of the Petitioner's proposed contentions meet the Commission's requirements for an admissible contention. The Petitioner's February 12, 1999 Request for Hearing and Petition to Intervene should be denied.
Respectfully submitted, Marian Zobler Counsel for N Staff Dated at Rockville, Maryland this 5th day of May,1999
p4 Attachment 1
'- SHNPP FSAR TABLE 9.2.2-1
~
I COMPONENT COOLING SYSTEM DESIGN PARAMETERS L GENERAL
' Service water supply temperature. F 95 Component cooling water design temperature. F 105/120 (max)
COMPONENT COOLING SURGE TANK Number 1 Design pressure, psig Internal 100-External ATM.
- Design tengerature. F -
200 Total volume.. gallons .
2000 Normal water volume, gallons 1000 COMPONENT COOLING HEAT EXCHANGER Number 2 Design heat transfer Btu /hr 50.5 x 10' l Shell T.ube Desi n pressure, psig 150 150 Desi n temperature. F 200 200 Desi n flow rate, lb./hr. 4.57 x 10' 6 x 108 0 erating inlet temperature. F 115.8 95 erating outlet temperature. F 104.9 103.4 l F uid circulated Component Service water cooling water COMPONENT COOLING WATER PUMP y
Horigontal/ centrifugal Design pressure, psig 150 Design temperature. F 200 Design flow rate, gallons per minute 8050 l
Minimum developed head at design flow, ft. 211 COMPONENT COOLING DRAIN TANK Number 1 Desi n Temperature. F 150 Desi n Pressure, osig ATM Tota Volume. gallons 300 Material of Construction Carbon Steel NOTE: Pum) impellers have been modified to develop a minimum of 211 ft. at 805) gpm.
9.2.2-10 Amendment No. 48
.- 1 SHNPP FSAR l
.. -TABLE 9.2.1-3
~-
MAXIMUM SERVICE WATER SYSTEM HEAT LOADS FOLLOWING LOCA
{ ..'
Safety In.iection Phase (approximately 30 min. - I hour in duration)
Heat Load to Service Water System (assumes two SWS loops in operation).
Charging Pumps 0.205 x 10' Btu /hr.
Containment Fan Coolers 210.8 x.108 Btu /hr. l Reactor Auxiliary Building HVAC Chiller 19.02 x 10' Btu /hr.
Standby Diesel Generators 27.4 x 108 Btu /hr. l
. Emergency SW Intake Structure Fan Cooler Units'" 0. Btu /hr.
Total 257.83 x 10' Btu /hr. l Recirculation Phase.
Heat Load to Service Water System (assumes two SWS loops in operation)
Component Coolir; Water Heat Exchanger 272.6 x 10' Btu /hr. l Containment Fan Coolers 210.8 x 10' Btu /hr. l Reactor Auxiliary, Building HVAC Chiller 19.02 x 105 Btu /hr.
Standby Diesel Generators 27.4 x 10' Btu /hr. l Emergency SW Intake Structure Fan Cooler'" 0. Btu /hr.
Total 530.22 x 10' Btu /hr. l l '" The ESW requirement to the intake structure air coolers is based on an L evaluation which permits isolation of ESW.
i l
l l
9.2.1-18 Amendment No. 48
. Attachment 3 bicember 1931
[*
a
- OFFICE OF NUCLEAR REGULATORY RLblANL,h Olvisten 1
- DRAFT REGULATORY GUIDE All0 VALUE/ IMPACT STATEMENT Task CE 913-5
(" \*..**
Contact:
C. Schulten (301)443-5910
- .,(4i.('
- l-PROPOSED REVISION 2* 10 REGULATORY GUIDE 1.13 SPENT FUEL STORAGE FACILITY DE51GN BASIS A. INTR 000CT!04 General Design Criterion 61, " Fuel Storage and Han fonctivity Control," of Appendix A. " General Design Criteria for .
r Plants "
to 10 CFR Part 50, "Comestic Licensing of Product 1 atton Facilities "
requires that fuel storage and handling systes e to ensure adequate l safety under normal and postulated accident It also requires that j these systems be designed (1) with a ca i ruit appropriate periodic l Inspection and testing of components i safety, (2) with suitable i shleiding for radiation protection propriate containment, confine- l ment, and filtering systees, (4) th a sidual heat renoval capability having I 4' re); ability and testability r the importance to safety of decay l heat and other residual heat ,and(5)topreventsignificantreduction !
In fuel storage coolant inven er accident conditions. This guide describes a sethe plable to the NRC staff for implementing Criterion 61. )
\
S. 915CV1519!!
W OANS57.2oftheAmericanNuclearSociety .
l ANS 50 eveloped a standard that detalls sintsus design requirements for ;
"The substantial number of changes in this proposed revision has made it lepractical to indicate the changes with lines in the margin. j nu i..r ,ii R o i.we..i.n.=i.ui m i.e,.. i...nv a.
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O spent fuel storage facilities at nuclear power stations. This standard was approv;d by the American National $tandards Coenittee N18 Nuclear Design Criteria. It was subsequently approved and designated AN51 N210-1976/AN$-57.2, '
" Design Objectives for Light Water Reactor $ pent Fuel $torage Facilities at Nuclear Power $tations," by the American National $tandards Institute on April 12, 1976.
Primary facllity design objectives are:
- a. To prevent loss of water from the fuel pool that would uncover fuel,
- b. To protect the spent fuel from mechanical damage, and
- c. To provide the, capability for limiting the potential offsite exposures
-in the event of significant release of radioactivity from the fuel.
If spent fuel storage facilities are not provided with adequate protective features, radioactive materials could be released to the environment as a result of either loss of water from the storage pool or mechanical damage to fuel within i the pool.
l
, 1. LOS$ 0F WATER FROM $TORAGE POOL l
Unless protective esasures are taken to prevent the loss of water from a fuel storage pool, the spent fv61 could overheat and cause damage to fuel cladding integrity, which could result in the release of radioactive saterf all to the environment. Equipment failures in syntess' connected to the pool could also result in the loss of pool water. A permanent coolant makeup systes designed with suitable redundancy or backup would prevent the fuel from being uncovered should pool leaks occur. Further, early detection of pool leakage and fuel desage can be sede using pool water level sonitors and pool radiation monitors that alarm locally and also at a continuously sanned location to ensure timely operation of building filtration systems. Natural events such as earthquakes or high winds can damage the fuel pool either directly or by the generation of l
sissiles. Earthquakes or high winds could also cause structures or cranes to fall into the pool. Designing the facility to withstand these occurrences without significant loss of watertight integrity will alleviate these concerns.
- i.)
_ . . . _ . --_ ~ - . . _ . _ . . _ .
- 2. MECHANICAL DAMAGE TO FUEL The release of, radioactive saterial from fuel may occur as a result of fuel-cladding fattures or mechanical damage caused by the dropping of fuel elementsorobjectsontofuelelementsduringtherefuelingprocessandat other times.
Plant arrangements consider low-probability accidents such as the dropping of heavy loads (e.g., a 100-ton fuel cask) where such loads are positioned or l soved in or over.the spent fuel pool. It is desirable that cranes capable of carrying heavy loads be prevented from moving into the vicinity af the stored fuel.
Missiles generated by high winds also are a potential cause of mechanical damage to fuel. , This concern can be eliminated by designing the fuel storage facility to preclude the possibility of the fuel being struck by missiles generated by high winds.
- 3. LIMITING P0TENTIAL 0FFSITE EXPOSURES Hechanical damage to the fuel might cause significant offsite doses unless dose reduction features are provided. Dose reduction designs such as negative pressure in the fuel handling building during movement of spent fuel would prevent exfiltration and ensure that any activity released to the fuel handling building will be treated by an engineered safety feature (ESF) grade filtration system before release to the environment. Even if seasures not described are used to maintain the desired negative pressure, small leaks from the building may still occur as a result of structural failure or other unforeseen events.
The staff considers Seismic Category I design assumptions acceptable for the spent fuel pool cooling, makeup, and cleanup systems. Tornado protection requirements are acceptable for the water makeup source and its deliver ,ystem, the pool structure, the building housing the pool, and the flitration .tilation systen. Regulatory Guide 1.52, " Design, Testing, and Maintenance Cr- *la for
- 1tration Post Accident Engineered-$afety-Feature Atmosphere Cleanup System 7 and Adsorption Units of Light-Water-cooled Nuclear Power Plants," a legulatory Guide 1.140, " Design Testing, and Maintenance Criteria for Normal % .ilation Exhaust system Air Filtration and Adsorption Units cf Light-Water-cooled Huclear D
1.13-3 9
N**" *6e..m., ,o,. ,
Power Plants," provide guidelines to. limit potential offsite exposures through the filtration-ventilation system of the pool building.
Occupational radiation exposure is kept as low as is reasonably achievable (ALARA) in '1 activities fr.volving personnel, and efforts toward maintaining exposures Aw'.4 $ e censidered in the design, construction, and operational phases. Guidance .n maintaining exposures ALARA is provided in Regulatory Gufde 8.8 "Information Relevant to Ensuring That occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable."
C. REGULATORY POSITION l
1 The requirements in A" ! N210-1976/ANS-57.2, " Design Objectives for Light
~
i Water Reactor Spent Fuel Storage Facilities at Nuc, lear Power Stations,"" are generally acceptable to the NRC staff as a means for complying with the require-ments of General Design Criterion 61, " Fuel Storage and Handling and Radio- ,
activity Control " of Appendix A. " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 as related to light-water reactors (LWRs),
) '
subjecttothefollowingclarificationsandmodifications:
- 1. In lieu of the example inventory in Section 4.2.4.3(1), the example '!
inventory should be that inventory of radioactive materials that are predicted to leak under the postulated maximum damage conditions resulting from the dropping of a single spent fuel assembly onto a fully loaded spent fuel pool storage rack. Other assumptions in the analysis should be consistent with those given in Regulatory Guide 1.25 (Safety Guide 25), " Assumptions Used for '
Evaluating the Potential Radiological Consequences of a Fuel Handling Accident ;
in the Fuel Handling and Storage Facility for Bolling and Pressurized Water l Reactors."
j
- 2. In addition to meeting the requirements of Section 5.1.3, boiling of the pool water may be permitted only when the resulting thermal loads are proper' ; counted for in the design of the pool structure, the storage racks, and other safety-related structures, equipment, and systems.
" Copies sh se obtained from the American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 60525 1.13-4
- 3. In addition to' meeting the requirements of Section 5.1.3, the fuel ;
D- storage pool should be designed (a) to prevent tornado winds and missiles generated by these winds from causing significant loss of watertight ir.tegrity of the fuel storage pool and (b) to prevent missiles generated by tornado winds from striking the fuel. These requirements are discussed in Regulatory Guide 1.117. " Tornado Design Classification." The fuel storage building.
. including walls and roof, should be designed to prevent penetration by tornado-generated missiles or from seismic damage to ensure that nothing bypasses the ESF grade filtration systes in the containment building.
- 4. In addition to meeting the requirements of Section 5.1.5.1, provisions should be made to ensure that nonfuel components in fuel pools are handled below the ainleum water shielding depth. A systes should be provided that, either through the design of the systes or through adstnistrative procedures, would prohibit unknowing retrieval of these components.
- 5. In addition to meeting the requirements of section 5.1.12.10, the maximum potential kinetic energy capable of being developed by any obeet handled 9 above stored spent fuel, if dropped, should not exceed the kinetic energy of one fuel assembly and its associated handling tool when dropped from the height at which it is normally handled above the spent fuel pool storage racks.
- 6. In addition to meeting the requirements of Section 5.2.3.1, an inter-face should be provided between the cask venting systes and the building ventila-tion system to minimize personnel exposure to the " vent-gas" generated from fl111ng a dry loaded cask with water.
- 7. In addition to meeting the requirements of Section 5.3.3, radioac-tivity released during a Condition IV fuel handling accident should be either contained or removed by filtration so that the dose to an individual is less than the guidelines of 10 CFR Part 100. The calculated offsite dose to an individual from such an event should be well within the exposure guidelines of 10 CFR Part 100 using appropriately conservative analytical methods and assumptions. In order to ensure that released activity does not bypass the 9
1.13-5
i A
' i filtration system, the ESF fuel storage building ventilation should provide and f saintain a negative pressure of at least 3.2 mm (0.125 in.) water gauge within l the fuel storage building. f i
In addition to the requirements of Section 6.3.1, overhead handling r l 8.
systems used to handle the spent fuel cask should be designed so that travel =
directly over the spent fuel storage pool or safety-related equipment is not possible. This should be verified by analysis to show that the physical '
structure under all cask handling pathways will be adequately designed so'that l
/
unacceptable damage to the spent fuel storage facility or safety-related !
equipment will not occur in the event of a load drop. i In addition to the references listed in Section 6.4.4, Safety Class 3, l
- 9. I Seismic Category 1, and safety-related structures and equipment should be subjected to quality assurance programs that meet the applicable provisions l l
of Appendix 8. " Quality Assurance Criteria for Nuclear Power Plants and Fuel
) . Reprocessing Plants," to 10 CFR Part 50. Further, these programs should obtain guidance from Regulatory Guide 1.28 " Quality Assurance Program Requirements f
(Design and Construction)," endorsing AN$! N45.2, and from the applicable provi- l sions of the AN$1 N45.2-series standards endorsed by the following regulatory guides: _
1.30 (Safety Guide 30) "Qua11ty Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment" (N45.2.4).
1.38 "QualityAssuranceRequirementsforPackaging, Shipping, Receiving, !
Storage, and Handling of Items for Water-cooled Nuclear Power Plants" (N45.2.2). ;
1.58 " Qualification of Nuclear Power Plant Inspection Examination, and Testing Personne)" (N45.2.5). l 1.64 " quality Assurance Requirements for the Design of Nuclear Power Plants" (N45.2.11).
1.74 " quality Assurance Terms and Definitions" (N45.2.10).
1.88 " Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records" (N45.2.9).
1.94 "QualityAssuranceRequirementsforInstallation, Inspection, and Testing of Structural Concrete and Structural steel During the Construction Phase of Nuclear Power Plants" (N45.2.5).
1.116 " Quality Assurance Requirements for Installation, Inspection. I and Testing of Mechanical Equipment and Systems" (N45.2.8).
4 1.123 " Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants" (N45.2.13).
- 10. The spent fuel pool water temperatures stated in Section 6.6.1(2) exceed the limits recommended by the NRC staff. For the maximum heat load during 9 ' Condition a single active Ifailure,occurrences with normal the pool water temperature cooling should be kept systems in at or below 60'C(140'F). Under abnormal maximum heat load conditions (full core unload) and also for Condition IV occurrences, the pool water temperature should be kept below boiling.
- 11. A nuclear criticality safety analysis should be performed in accord-ance with Appendix A to this guide for each systes that involves the handling, transfer, or storage of spent fuel a.semblies at LWR spent fuel storage facilities.
- 12. The spent fuel storage facility should be equipped with both electrical interlocks and mechanical stops to keep casks free being transported over the spent fuel pool.
- 13. Sections 6.4 and 9 of ANS-57.2 list those codes and standards referenced in ANS-57.2. Although this regulatory guide endorses with clarifications and modifications ANS-57.2, a blanket endorsement of those referenced codes and D
1.13-7
' " " -
- w +- - sm., ,, ,_ _ _
\
I x standards is not intended. (Other regulatory guides may contain some such l
endorsements.),
D. IMPLEMENTATION i
The purpose of this section is to provide information to app 1(cants regard-ing the NRC staff's plans for using this regulatory guide.
This proposed revision has been released to encourage public participation !
~
in its development. Except in those cases in which an applicant proposes an i
acceptable alternative method for complying with specified portions of the Commission's regulations, the method to be described in the active guf'de reflecting public comments will be used in the evaluation of applications for construction permits and operating licenses docketed after the implementation date to be specified in the active guide. Implementation by the staff will in no case be earlier than June 30, 1982. j l
{
l 4
1.13-8 !
i
~
. 4 1
APPENDIX A D .
NUCLEAR CRITICALITY SAFETY t
- 1. SCOPE OF NUCLEAR CRITICALITY $AFETY ANALYS!$
i 1.1 A nuclear criticality safety ~ analysis should be performed for each systen j that involves the handling, transfer, or storage of spent fuel assemblies at j light-water reactor (LWR) spent fuel storage facilities. !
1 1.2 The nuclear criticality safety analysis should demonstrate that each LWR spent fuel storage facility systes is subcritical (k,gg nottoexceed0.95).
1.3 The nuclear criticality safety analysis should include consideration of all cred E h normal and abnormal operating occurrences, including: j Accidental tipping or falling of a spent fuel assembly, j s.
- b. Accidental tipping or falling of a storage rack during transfer. l
- c. Hisplacement of a spent fuel assembly, j
- d. Accumulation of solids containing fissile materials on the pool floor or at. locations in the cooling water systes,
- e. Fuel drop accident.s.
- f. Stuck fuel assembly / crane uplifting forces,
- g. Horizontal motion of fuel before complete removal from rack,
- h. Placing a fuel assembly along the outside of rack, and
- 1. Objectsthatmayfallontothestoredspentfuelassemblies.
1.4 At all locations in the LWR spent fuel storage facility where spent l fuel is handled or stored, the nuclear criticality safety analysis should
-demonstrate that critica.11ty could 31 occur without at least two unlikely, ,
independent, and concurrent failures or operating limit violations, a
1.5 The nuclear criticality safety analysis should explicitly identify spent fuel assembly characteristics upon which subcriticality in the LWR spent fuel storage favility depends.
D 1.13-9
=
1.6 The nuclear criticality safety c.ui/f 2 should explicitly identify design limits upon which subcriticality dt;,et4 %.4t require physical verification at the completion of fabrication or cons nuction.
1.7 The nuclear criticality safety analysis should explicitly identify operating limits upon which suberiticality depends that require implementation in operating procedures.
2.- cat.CULATION HETHODS AND CODES Methods used to calculate subcriticality should be validated in accordance with Regulatory Guide 3.41, " Validation of Calculational Methods for Nuclear Criticality Safety." which endorses ANSI N16.9-1975.
- 3. METHOD TO ESTABt1SH $UBCRITICALITY 3.1 The evaluated multiplication factor of fuel in the spent fuel storage racks, k g, under normal and credible abnormal conditions should be equal to or less than an established maximum allowable multi. . cation factor, k,;
i.e.,
ksIIa The factor, k,, sn.suld be evaluated from the expression:
l k, = k,, + Aksb + Aku
- AI sc l where k,, a the computed effective multiplication factor; k an is calculated by the same methods used for be..chmark experiments for design storage parameters when the racks are loaded with the most reactive fuel to be stored,
) )
1,13-10
Ak sb = the bias in the calculation procedure as obtained from the D comparisons with experiments and including any extrapolation to storage pool conditions, Ak, = the uncertainty in the benchmark experiments, and Ak,,a the combined uncertainties in the parameters listed in para-graph 3.2 below.
3.2 The combined uncertainties, Ak,,, include:
- a. Statistical uncertainty in the calculated result if a Monte Carlo calculation is used,
- b. Uncertainty resulting from comparison with calculational and experimental results,
- c. Uncertainty in the extrapolation from experiment to storage rack condi-9 tions, and
- d. Uncertainties introduced by the considerations enumerated in para-graphs 4.3 and 4.4 below.
3.3 The various uncertainties may be combined statistically if they are independent. Correlated uncertainties should be combined additively.
3.4 All uncertainty values should be at the 95 percent probability level with a 95 percent confidence value.
3.5 For spent fuel storage pool, the value of k, should be no greater than 0.95.
- 4. STORAGE RACK ANALYSIS ASSUMPTIONS 4.1 The spent fuel storage rack module design should be based on one of the
~
(o110 wing assumptions for the fuel:
1.13-11
- s. The most reactive fuel assembly to be stored at the most reactive poin.t'in the assembly life, or j i
- b. The most reactive fuel assembly to be stored based on a minimum 1
confirmed burnup (see Section 6 of this appendix).
Both typts of rack modules may be present in the same storage pool.
4.2 Determination,of the most reactive spent fuel assembly includes considera-tion of the following parameters: .
- a. Maximum fissile fuel loading,
- b. Fuel rod diameter,
- c. Fuel rod cladding material and thickness,
- d. Fuel pellet density,
- e. Fuel rod pitch and total number of fuel rods within assembly,
- f. Absence of fuel rods in certain locations, and
- g. Burnable poison content.
4.3 The fuel assembly arrangement assumed 1.. storage rack design should be the arrangement that results in the highest value of k, considering:
- a. Spacing between assemblies,
- b. Moderation between assemblies, and
- c. Fixed neutron absorbers between assemblies. .
4.4 Determination of the spent fuel assembly arrangement with the highest value l
-of k, shall include consideration of the following: ;
l
- a. Eccentricity of fuel bundle location within the racks and variations in_spacingamongadjacentbundles,
- b. Dimensional tolerances,
- c. Construction materials,
- d. Fuel and moderator density (allowance for void formations and temper-ature of water between and within assemblies),
(
1.13-12 e-,,w .
^-
-- _-~~..--,o---.-_-. .
l l
- a. Presence of the remaining amount of fixed neutron absorbers in fuel assembly, and
- f. Presence of structural material and fixed neutron absorber in cell walls between assemblies.
4.5 Fuel burnup determination should be made for fuel stored in racks where credit is taken for burnup. The following methods are acceptable:
- a. A minimum allowable fuel assembly reactivity should be established, and a reactivity measurement should be performed to ensure that each, assembly meets this criterion; or
- b. A minimum fuel assembly burnup value should be established as deter-mined by initial fuel assembly enrichment or other correlative -
parameters, and a measurement should be performed to ensure that each fuel assembly meets the established criterion; or
- c. A minimum fuel assembly burnup value should be established as deter-mined by initial fuel assembly enrichment or other correlative param-eters, and an analysis of each fuel assembly's exposure history should be performed to determine its burnup. The analyses should be performed under strict administrative control using approved written procedures.
These procedures should provide for independent checks of each step of the analysis. by a second qualified person using nuclear criticality safety assessment criteria described in paragraph 1.4 above.
The uncertainties in determining fuel assembly storage acceptance criteria should be considered in establishing storage rack reactivity, and auditable records should be kept of the method used to determine the fuel assembly storage acceptance criterion for as long as the fuel assemblies are stored in the racks.
Consideration should be given to the axial distribution of burnup in the fuel assembly, and a limit should be set on the length of the fuel assembly that is permitted to have a lower average burnup than the fuel assembly average.
1.13-13
t
, 5. USE OF NEUTRON ABSORBER $ IN STORAGE RACK DESIGN 5.1 Fixed' neutron absorbers may be used for criticality control under the following conditions:
- a. The effect of neutron-absorbing materials of construction or added fixed neutron-absorbers may be included in the evaluation if they are designed and fabricated so as to preclude inadvertent removal by mechanical or chemical action. ,
- b. Fixed neutron absorbers should be an integral, nonremovable part of the storage rack.
- c. When a fixed neutron absorber is used as the primary nuclear criticality safety control, there should be provision to:
9 '
(1) Initially confirm absorber presence in the storage rack, and (2) Periodically verify continued presence of absorber.
5.2 The presence of a soluble neutron absorber in the pool water should not normally be used in the evaluation of k,. However, when calculating the effects of Condition IV faults, realistic initial conditions (e.g., the presence of soluble boron) say be assumed for the fuel pool and fuel assemblies.
l
- 6. CREDIT FOR BURNUp IN STORAGE RACK DESIGN t 6.1 Consideration should be given to the fact that the reactivity of any given spent fuel assembly will depend on initial enrichment, assu depletion, amount of burnable poison, plutonium buildup and fission product burnable poison depletion, and the fact that the rates of depletion and plutonium and fission product b'ulldup are not necessarily the same.
D 1 1.13-14
i 6.2 Consideration should be given to the practical implementation of the spent
' fuel screening process. Factors to be considered in choosing the screening method should ' include:
- a. Accuracy of the method used to determine storage rack reactivity;
- b. Reproducibility of the result, i.e., what is the uncertainty in the result? .
- c. Simpilcity of the procedure; 1.e., how much disturbance to other operations is involved?
- d. Accountability, i.e., ease and completeness of recordkeeping; and
- e. Auditability.
f S
a .
1.13-15
1 DRAFT VALUE/ IMPACT STATEMENT
- 1. PROPOSED ACTION 1.1 Description .
l Each nuclear power plan.t has a spent fuel storage factitty. General Design Criterton 61, " Fuel Storage and Handling and Radioactivity Control," of Appendix A.
" General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities," requires that fuel storage and handling systems be designed to ensure adequate safety under normal and postulated accident conditions. The proposed action would prov'ide an acceptable method for implementing this criterion. This action would be an update of Regulatoiy Guide 1.13. "$ pent Fuel Storage Facility Design Basis." l 1.2 Need for Proposed Action Since Regulatory Guide 1.13 was last published in December of 1975, addt-
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tional guidance has been provided in the form of ANSI standards and NUREG reports. The Office of Nuclear Reactor Regulation has requested that this guide be updated.
1.3 Value/!spect of Proposed Action 1.3.1 JLR$
. The applicants' basis for the' design of the spent fuel storage facility will be the same as that used by the staff in its review of a construction permit or operating license application. Therefore, there should be a minimus number of cases where the applicant and the staff radically disagree on the design criteria. ,
1.3.2 Government Acencief, Applicable only if the agency, such as TVA, is an appilcant. .
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1.3.3 Industry '
The value/ impact on the applicant will be the same as for the NRC staff.
1.3.4 Public )
Nomajorimpactonthepubliccanbeforeseen. I 1.4 Decision on Prooosed Action ]
1 The guidance furnished on the design basis for the spent fuel storage facility l should be updated. l l
- 2. TECHNICAL APDR0ACH l TheAmericanNuclearSocietypublishedANS-57.2(ANSIN210),"DesignObjectives I for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations."
Part of the update of Regulatory Guide 1.13 would be an evaluation of this standard and possittle endorsement by the NRC. Also, recommendations made by Task A-36, t which were published in NUREG-0612. " Control of Heavy Loads at Nuclear Power 9 Plants," would be included.
- 3. PROCEOURAL APPR0ACH Since Regulatory Guide 1.13 already deals with the proposed action, logic dictates that this guide be updated.
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- 4. STATUTORY CONSIDERATIONS 4.1 NRC AUTHORITY Authority for this regulatory guide is derived from the safety requirements of the Atomic Energy Act of 1954, as amended, through the Commission's regulations, in particular General Design Criterion 61 of Appendia A to 10 CFR Part 50. j
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4.2 Need for NEPA Assessment )
The proposed action is not a major actibn as defined by paragraph 51.5(a)(10) of 10 CFA Part 51 and does not require an environmental tapact statement.
S. CONCLU$10N Regulatory Guide 1.13 should be updated.
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DOCXETED UNITED STATES OF AMERICA DI' NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3 MY -6 P3 :08 I l
OFF , -
In the Matter of ) Ra , *
) ADJL ,y CAROLINA POWER & LIGHT COMPANY ) Docket No. 50-400-LA
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(Shearon Harris Nuclear Power Plant) )
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CERTIFICATE OF SERVICE i l
I hereby certify that copies of"NRC STAFF'S RESPONSE TO ORANGE COUNTY'S SUPPLEMENTAL PETITION TO INTERVENE" in the above captioned proceeding have been served on the following by electronic mail with conforming copies deposited in the Nuclear l Regulatory Commission's. internal mail system, or as . indicated by an asterisk, by E-mail with l conforming copies deposited in the United States mail, first class, or as indicated by a double ;
asterisk by deposit in the NRC's intemal mail system this 5th day of May,1999.
G. Paul Bollwerk,III, Chairman Frederick J. Shon Administrative Judge Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board Mail Stop: T 79.3 Mail Stop: T-3F-23 U.S. Nucleav.cgulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 (E-mail copy to GPB@NRC. GOV) (E-mail copy to FJS@NRC. GOV)
Dr. Peter S. L.am Office of the Secretary Administrative Judge ATTN: Rulemaking and Adjudications
. Atomic Safety and Licensing Board Staff Mail Stop: T 3F-23 Mail Stop: O 16-C-I U.S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission
. Washington, DC 20555 Washington, DC 20555 (E-mail copy to PSL@NRC. GOV) (E-mail copy to HEARINGDOCKET
@NRC. GOV)
Office of the Commission Appellate Atomic Safety and Licensing Board Panel" Adjudication" Mail Stop: T 3F-23 Mail Stop: O 16-C-1 U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555
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l Adjudicatory File ** James M. Cutchin, V l Atomic Safety and Lic'ensing Board Atomic Safety and Licensing Board MailStop: T3F-23 Mail Stop: T 3F-23 U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission l Washington, DC 20555 ' Washington, DC 20555 (E-mail copy to JMC3@NRC. GOV) l Diane Curran, Esq
- John H. O'Neill, Jr.* I Harmon, Curran, Spielberg William R. Hollaway*
& Eisenberg, L.L.P. Counsel for Licensee 1726 M Street, N.W., Suite 600 Shaw Pittman Potts & Trowbridge Washington, DC 20025 2300 "N" Street, N.W. !
(E-mail copy to Washington, DC 20037-1128 .
_dcurran@harmoncurran.com (E-mail copy to l
, john _o'neill@shawpittman.com, william.hollaway@shawpittman.com Steven Carr*
Legal Department '
Carolina Power & Light Co. ,
411 Fayetteville Street Mall P.O. Box 1551- CPB 13A2 Raleigh, North Carolina 27602 (E-mail copy to steven.carr@cplc.com n
Marian L Zobler Counsel for NRC S 4
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