ML20237E181

Forwards Description of Review Methodology & Summary of Results,Including List of Identified Variations from Guidance Contained in Std Review Plan,Per NUREG-1536

ML20237E181
Person / Time
Site:	Trojan  File:Portland General Electric icon.png
Issue date:	08/25/1998
From:	Quennoz S PORTLAND GENERAL ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-1536 TAC-L22102, VPN-048-98, VPN-48-98, NUDOCS 9808310017
Download: ML20237E181 (51)

Text

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Portland General Electric i Trojan Nuclear Plant l 71760 Columbia Riter Huy )

Rainier OR 97048

, (5031 556-3713 L August 25,1998 I  !

l VPN-048-98 1

l Trojan ISFSI Docket 72-017 '

i NRC Document Control Desk l U. S. Nuclear Regulatory Commission l Washington, DC 20555 l

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Dear Sir:

Comnazison of Trojan ISFSI Design to NUREG-1536. " Standard Review Plan for Drv Cask Storage SystemsStandard Review Plan."(TAC No. L22102) l On March 26,1996 (VPN-012-96), PGE submitted an application for a license for the Trojan Independent Spent Fuel Storage Installation (ISFSI). Subsequently, in January of 1997, the NRC ,

issued NUREG-1536," Standard Review Plan for Dry Cask Storage Systems." In recent l

' discussions with PGE representatives, the NRC staff requested that PGE conduct a review to determine the degree to which the Trojan ISFSI conforms to the guidance contained in the Standard Review Plan.  :

l PGE has completed the requested review and the results are included as Attachment I to this I

letter. Attachment 1 contains a description of the review methodology and a summary of the

( results, including a list ofidentified variations from the guidance contained in the Standard

! Review Plan. The discussion of each identified variation also includes the basis for the acceptability of the proposed Trojan ISFSI design or analysis. The review also includes a j chapter by chapter summary of the conformance of the Trojan ISFSI to the criteria contained in the Standard Review Plan.

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l-l VPN-048 L ' August 25,1998 Page 2 of 2 If you have any questions regarding this information, please contact Ray Pate of my staff at (503)

L 556-7480.

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Sincerely,

,d'=p h n= %% = =

Stephen M. Quennoz Trojan Site Executive L Attachments .

i' c: C. J. Haughney, NMSS, w/o attachments -

l T. J. Kobtez, NRC, NMSS,(15 copies)

L. H. Thonus, NRC, NRR, w/o attachments R. A. Scarano, NRC Region IV, w/o attachments David Stewart-Smith, OOE l

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Attachment 1 i

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t l Comparison of Trojan ISFSI License Application with l NUREG-1536," Standard Review Plan  !

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for Dry Cask Storage Systems" l

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Comparison of Trojan ISFSI License Application with NUREG-1536," Standard Review Plan for Dry Cask Storage Systems" I. Introduction The purpose of this document is to compare the Trojan ISFSI license application with the guidance provided in NUREG-1536," Standard Review Plan forDry Cask Storage Systems."

This comparison document is divided into the following four sections: Section I, " Introduction";

Section II, " Scope of Review";Section III, "Results of Review"; and Section IV, " Comparison of Trojan ISFSI with SRP Guidance." Section 111 provides a summary of results, highlights ten variations from the Standard Review Plan (SRP) guidance which were identified during the review, and provides a basis for each variation.Section IV provides a chapter-by-chapter comparison.

Potential differences between the Trojan ISFSI application and the Standard Review Plan arise primarily because the SRP for dry cask storage systems was issued in January of 1997, almost a year after PGE filed the application for the Trojan ISFSI. The Standard Review Plan is formatted for a review of a topical safety analysis report such as would be submitted by a cask vendor in support of a certificate of compliance for a generic dry cask storage system. The sections of the Standard Review Plan are keyed to the format provided in Regulatory Guide 3.61,

" Standard Format and Content of Topical Safety Analysis Reports for a Spent Fuel Dry Storage Facility." PGE prepared the Trojan ISFSI SAR using the guidance of Regulatory Guide 3.48,

" Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage . Installation or Monitored Retrievable Storage Installation (Dry Storage)." The plant specific nature of the Trojan submittal also results in some differences in format and content.

For example, there is no SRP chapter that covers the NRC review of Trojan ISFSI SAR Chapter 2, " Site Characteristics," since a topical SAR for a cask system does not address a predefined ISFSilocation.

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II. Scope of Review l

The Standard Review Plan contains fourteen major sections corresponding to the expected chapters of a topical SAR. Each major section contains seven subsections as listed below:

I. Review Objective II. Areas of Review Ill. Regulatory Requirements IV. Acceptance Criteria V. Review Procedure VI. Evaluation Findings VII. References This comparison focuses on Section IV," Acceptance Criteria," and Section V," Review Procedure." Review of the Standard Review Plan indicates that Sections I and II contain general information describing the scope of the review for each chapter. These sections do not, however, contain specific criteria that are subject to comparison to the Trojan ISFSI.Section III of the SRP chapters lists the basic regulatory requirements (CFR Sections) that the NRC staff has relied upon in developing the acceptance criteria of Section IV. Since Section IV contains acceptance criteria that the NRC has found to be sufficient to satisfy the regulatory requirements, Section Ill will not be addressed directly in this comparison.Section VI," Evaluation Findings," contains guidance to the NRC reviewers relative to the conclusions that should be reached and the wording to be included in the Safety Evaluation Report. This section does not contain criteria that can be compared to the Trojan ISFSI SAR or license application.Section VII, " References,"

simply lists the references commonly used in the review process and does not contain criteria that can be compared against the Trojan ISFSI application.

l The Standard Review Plan is written under the assumption that most necessary infom1ation is j contained in the body of the SAR. In the case of the Trojan ISFSI, however, the ISFSI SAR has been supplemented by responses to two rounds of requests for additional information (RAls) and several additional submittals that either provided additional detail or described changes in the originally proposed design of the ISFSI. In addition, PGE submitted much of the detailed l information related to fuel loading activities on the 10 CFR 50 docket (Ref: License Change Application 237). The intent of this review is to confimi that information cited in the Standard Review Plan has been provided to the NRC and provide a " road map" to the location of the l l

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l infonnation. This review is not mtended to identify discrepancies related to the fom1at or content l of the Trojan ISFSI SAR itself.

The purpose of this companson is to aid the NRC reviewers by identifying how the Trojan ISFSI complies with the objective criteria given in the Standard Review Plan. The Standard Review Plan contains a mixture of objective and subjective criteria. For example, much of the Standard l

Review Plan discussion directs the reviewers to draw a conclusion as to whether ccrtain features or analyses are " adequately" described or whether analyses are " adequate." These subjective conclusions are not within the scope of this comparison. If the Standard Review Plan requires that the SAR contain a description of a certain design feature or analysis, this comparison will note whether or not the information has been provided to the NRC and reference the documents that provide the information.

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i III. Results of Review PGE's review indicates that Trojan ISFSI is substantially in conformance with the guidance of the NRC's Standard Review Plan for Dry Cask Storage Systems. The review identified a limited l number of variations from the objective guidance and acceptance criteria delineated in the Standard Review Plan. In each case, PGE has concluded that the Trojan ISFSI design complies with the applicable regulatory requirements and provides an acceptable alternative to the

. guidance of the Standard Review Plan Each of the identified variations is described below.

1. Evaluation of Potential Fuel Cladding Degradation due to DCCG Description of Variation SRP Chapter 4.0, " Thermal Evaluation," recommends that fuel cladding damage resulting from creep cavitation (diffusion controlled cavity growth or "DCCG") should be limited to 15 percent of the original cladding cross-sectional area calculated using the methodology of UCID-21181.

PGE has not performed this specific evaluation for the Trojan fuel cladding.

Basis for Variation from SRP l

This UCID-21181 report postulates pre-existing cracks or flaws in zircaloy metals grow under the stress and temperatures normally seen during dry fuel storage conditions. Although this failure mechanism has not been specifically observed in Zirconium cladding, empirical data is '

available on tests run on brass, copper, and magnesium'. The UCID-21181 model has never been

. validated against cavitation data and, in fact, voids and cracks are very infrequently seen in irradiated Zircaloys2, l l

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The available technical reports are in good agreement on the combination of temperature and stress relationships that define the onset of DCCG. The postulated clad temperature range of concern is >300' C, and the fuel rod hoop stresses of concern are >120 MPA . Experimental M W Schwartz and M C Witte.1987, Spent Fuel Cladding Integrity During Dry Storage, UCID-21181, Lawrence Livermore National Laboratory, Livermore, CA 2M A McKinnon, R E Enzinger, D L Baldwin, and S G Pitman.1998, Data Needs for Long Term Dry Storage of LWR Fuel, EPRI TR- 108757, Battelle Pacific Northwest Division, Richland, WA 4

data has shown that at 350 C, especially at stress over 100 MPA that grain boundary cavities can be fomied. The DCCG failure mechanism was considered in the development of the conservative long term maximum storage temperature limit of 380 C recommended by PNL t

reports 4835,6189, and 6364. These PNL reports form the basis for the Trojan long term fuel cladding temperature limit.

In 1995-6, Lawrence Livem1 ore National Labs (LLNL), developed an improved model for Zircaloy DCCG. When the improved DCCG model is applied to Zircaloy using updated and I recommended Zircaloy properties, DCCG behavior disappears as a practical concern for Zircaloy clad spent fuel rods. With this model, DCCG temperature limits for Zircaloy increase by over 300 C. I i

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In the Trojan fuel evaluation case, only a limited number of fuel rods approach the region where l the onset of DCCG would be of concem. Trojan specific values, as documented in various Westinghouse Topical Reports, for the maximum fuel rod pressure for the worst case fuel assembly is ~92 MPA. The fuel cladding temperature limits applied to Trojan fuel are all based on the worst case fuel rod, resident in the worst case fuel assembly that has the highest bumup and highest initial fill gas pressure. These limits are used to represent over 208,000 spent fuel rods at Trojan. All other fuel rods have much lower fuel rod stress values than calculated here.

J Nonetheless, the cladding stress value for the maximum fuel rod pressure is well below the stress j recognized as required for the onset of DCCG. This fuel rod pressure is based on a long term  ;

temperature limit of 380 C, whereas actual steady state conditions are calculated to be less than 315 C, if this value is applied to the PNL equations, then the actual stress level would be even j lower, ~33 MPA.

Another approach to the evaluation of the potential for DCCG failures is to apply the time to I

rupture due to creep mechanisms as derived in EPRI TR-103949, Temperature Limit Determination for the Inert Dry Storage of Snent Nuclear Fuel and use Trojan data to determine expected time to failure for all fuel related failure mechanisms. In each case, temperature is held constant until rod failure occurs. This expression does not account for temperature reductions due to radioactive decay /nuclide half-life which would drop temperature levels to where creep is no longer an active mechanism (i.e.,20 years).

tr= 10 nio,wn2mmnnp2o l

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a = initial hoop stress in Mpa tr = time to rupture in hours T = temperature in degrees Kelvin

- In case: 1) 100 Mpa at 380 C

2) 92 Mpa at 315 C (expected maximum values for Trojan fuel) i Each case is evaluated to show expected failure rates of a fuel rod. In case 1) the fuel rod will fail in 669.7 years while in case 2) the fuel rod will not fail (i.e.,1.2X105 years). This expression .

gives approximate values for all creep mechanistic failures and it bounds the DCCG model. As  !

I shown by these evaluations, below 100 Mpa and less than 380 C, the DCCG failure mechanism is not of any consequence.

EPRI Report TR-108757 concludes (based on domestic and international empirical data):

1. There is no evidence that DCCG is active as a primary breach mechanism of Zircaloy clad fuel; and the validity of the DCCG mechanism is more pertinent during the first 20 l years because it appears to be inoperative at lower temperatures. l

2. The most active mechanisms of fuel degradation occur in the first 10 years of storage, ,

i where the temperature is decreasing from its maximum to ~125 C. Beyond 10 years, the temperature is low enough that further fuel degradation is precluded in the remaining 90 years of dry storage, p

3. It is feasible to dry store nuclear spent fuel for 100 years.

Trojans fuel falls well below the documented guidelines for the onset of DCCG and where DCCG is thought to be a failure mechanism. Since the onset of DCCG is never

_ reached, DCCG is not a concem for fuel to be stored at the Trojan ISFSI.

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2. Occupancy Rate for Off-Site Dose Calculations Description of Variation SRP Chapter 5.0," Shielding Evaluation," notes that off-site radiation exposure should be calculated using a full year occupancy (8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />). The Trojan calculations are based on 2000 -

hours occupancy. However, the Standard Review Plan notes that lower number may be acceptable ifjustified.

Basis for Variation from SRP 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> is a conservative occupancy rate for members of the general public based on the location, site arrangement, and historical usage at the Trojan site. The Trojan ISFSI is located within the boundaries of the Trojan Nuclear Plant site; The portions of the controlled area boundary that are accessible to the general public are located in rugged steep terrain bordering q

the Columbia River. There are no dwellings,' structures, or recreation areas adjacent to the controlled area boundary and members of the public would rarely approach the site along those ,

portions of the controlled area. The ISFSI controlled area also encompasses portions of the developed area of the Trojan Nuclear Plant site, including portions of the main office building.

Since members of the public may occasionally occupy this area,2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> was selected as a conservative bounding value based on continued occupancy during normal working hours (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> a week for 50 weeks / year).

3. Potential Effects of High Temperature on Shielding Materials l r Description of Variation I I

l SRP Chapter 5.0," Shielding Evaluation," notes that the potential for changes in shielding density at high temperatures should be addressed. This topic is not explicitly addressed in the Trojan ISFSI SAR.

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Basis for Variation from SRP The gamma shielding provided by the steel and concrete basket and cask will not be significantly affected by increased temperatures during fuel storage. The storage temperatures of the concrete casks are not expected to be sufficient to drive off the water of hydration that provides the necessary hydrogen for shielding of neutrons. The concrete casks require the water of hydration in order to maintain structural integrity. These casks are designed for a structural life of at least 40 years and the cask shielding properties are expected to remain within specifications over the lifetime of the casks. Any unexpected degradation of the concrete integrity would be identified during routine inspections and changes to cask shielding characteristics would be identified during routine radiation dose surveys and/or analysis of TLD monitoring data.

PGE addressed the potential effects of elevated temperatures on the neutron shielding material utilized in the Transfer Cask in response to NRC's RAl-2, Question 7-2. PGE's response notes that the shielding calculations were revised to conservatively bound the potential loss of hydrogen from the shielding material due to elevated temperatures.

4. Periodie Monitoring of the Efficacy of Neutron Absorber Materials Description of Variation SRP Chapters 6.0, " Criticality Evaluation," and 9.0, " Acceptance Tests and Maintenance Program," note that periodic monitoring / testing of neutron absorber materials is required by 10 CFR 72.124(b). The sealed design of the Trojan ISFSI does not allow direct monitoring or testing of the neutron poison material during the storage period.

Basis for Variation from SRP Although the TranStor storage system incorporates a fixed neutron poison material in the storage basket, the Trojan ISFSI criticality analyses do not take credit for this material. In addition, PGE has shown in response to NRC RAI-1, Question 1-21 that the neutron absorber material will not be significantly affected by the expected fuel storage temperatures or the expected radiation exposure. The use of this material meets the guidance of SRP Section 6.V.3.b (p. 6-4) for demonstrating the continued efficacy of the material by design and material properties in lieu of a surveillance or monitoring program.

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Nonetheless, PGE has also requested an exemption from this requirement based on the acceptable criticality analyses without credit for the neutron absorbing material (Ref: VPN-023-97 dated March 20,1997).

5. Methodology for Off-Site Radiological Dose Evaluation Description of Variation SRP Chapter 7, " Confinement Evaluation," recommends that the analysis of the hypothetical failure of the confinement banier be modeled as an " instantaneous release" and include the isotopes listed in SRP Table 7-1. The Trojan analysis is modeled as a 2-hour release and does not consider all of the isotopes listed in Table 7-1.

Basis for Variation from SRP The assumption of a two (2) hour release with concurrent exposure by the receptor for two hours results in the same dose as for an " instantaneous" release. The assumption of a two (2) hour

release and exposure period is consistent with other accident analyses and with the assumptions in the Trojan ISFSI Emergency Plan. Sine the calculated doses are the same, this variation from the Standard Review Plan guidance has no substantive impact and no change in the SAR analyses is necessary.

The Trojan ISFSI dose calculation did not assume that all the isotopes noted on SRP Table 7-1 would be released in the case of a hypothetical accident. The Trojan ISFSI fuel storage baskets

' are stored at approximately atmospheric pressure. Therefore, a breach in the confinement barrier would not result in the needed motive force to release nonvolatile elements. Therefore, the calculation considered the dose from "Kr and 3H.

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6. Evaluation of Radiological Consequences of Basket Leakage Description of Vardon l SRP Chapter 7 sta,es that the maximum acceptable basket leakage rate should be evaluated for radiological consequences. This evaluation is not contained in the Trojan ISFSI SAR or other licensing basis documents.

Basis for Variation from SRP The Trojan ISFSI uses a metal fuel storage basket with redundant seal welded closures.

Therefore, there will be little or no leakage during normal operation of the ISFSI. The maximum allowable basket release rate as specified in the proposed Trojan ISFSI Technical Specifications is IE-4 sec/sec. The Trojan fission product accident atmospheric release calculation assumes all the fuel rods in a basket fail with release of 30% of the "Kr and 3H in each rod. This inventory of radioactive material is then released over a two hour period. This is equivalent to a leak rate of approximately 8.28E2 sec/sec. (5.96E6 secl7200 sec.). The estimated radiation dose from a basket leaking at the maximum allowable rate can be found using the ratio of the accident leakage rate to the maximum allowable leakage rate and multiplying the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose by 1000 to obtain an annual dose (assuming 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year occupancy). The result is then divided by 100 since only a maximum of 1% of the fuel pins are assumed to fail during normal operations. The results of this estimation are presented below:

Distance Wholebody Skin (meters) (mrem / year) (mrem / year) 100 2.4 48 200 0.7 14 300 --

7.0 330 --

5.8 500 0.14 2.8 1000 0.04 0.9 i

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l 7 . 12rocedures for Unloading Storage Baskets l

Description of Variation l

SRP Chapter 8.0," Operating Procedures," recommends that procedures be provided for unloading spent nuclear fuel from a storage cask. The design of the Trojan ISFSI does not rely on the capability to unload a storage basket once a basket has been drained and vacuum drying operations have commenced and the SAR does not describe procedures for such an evolution.

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Basis for Variation from SRP l

The TranStor storage system is part of a modular system that provides for storage and shipping ]

of the spent nuclear fuel without the need for repackaging the fuel or reliance on the Trojan ,

Nuclear Plant Fuel Pool. The Trojan ISFSI is provided with a dry transfer facility (Transfer I Station) and shielded Transfer Cask that allow a sealed metal storage basket to be removed from the concrete storage cask and loaded into a shipping cask for transportation off-site.

In the unlikely event of a failure of a storage basket confinement boundary during storage, the l Trojan ISFSI design provides for the use of a metal overpack to restore the confinement boundary. A leaking storage basket can be removed from the concrete storage cask and placed in an overpack using the dry transfer facility without the need for unloading the spent nuclear fuel from the failed storage basket.

The Standard Review Plan does not contain explicit review criteria for dry transfer facilities such as the Trojan ISFSI Transfer Station. The Standard Review Plan notes that such guidance will be developed at a later date.

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8. Procedures for Vacuum Drylag and Backfilling Description of Variation l

The Standard Review Plan guidance on cask draining and drying operations in Chapter 8.0,

" Operating Procedures," references PNL-6365 with regard to vacuum drying operations. This document is not directly referenced in the Trojan ISFSI licensing basis documents. PNL-6365 and the Standard Review Plan recommend that the cover gas in each cask be sampled after l backfilling to verify purity. This is not included in current ISFSI procedures.

Basis for Variation from SRP l

Trojan ISFSI procedures require the basket be evacuated and backfilled twice with helium consistent with the guidance in the Standard Review Plan and PNL-6365. PGE will use 99.999%

Grade 5 helium as a cover gas which is higher quality than the 99.995% weld grade tested in l PNL-6365.

Sampling of helium after the final backfill is not necessary based on the following I considerations:

The vacuum drying system is designed to preclude in-leakage. No inadvertent operation can result in in-leakage of oxidizing gases into the basket.

The purity level of the helium gas is controlled by quality inspection prior to receipt.

The vacuum drying system is operated by trained and certified operators and in accordance with approved procedures.

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9. Periodic Inspection and Maintenance

' Description of Variation SRP Chapter 9.0, " Acceptance' Tests and Maintenance Program," recommends that the SAR discuss the periodic inspection oflifling and rotating trunnion load-bearing surfaces and the periodic testing of support systems (e.g., vacuum drying, leak testing, etc.). These activities are .

not discussed in the Trojan ISFSI SAR or other licensing basis documents.

Basis for Variation from SRP Although not addressed in the SAR, PGE identified the need to evaluate needed periodic

inspections such as these as part of the review oflessons learned from other facilities. These specific recommendations are being tracked as part of the PGE's Lessons Learned Program for the ISFSI and will be incorporated into ISFSI procedures as appropriate.

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Description of Variation

SRP Chapter 14.0," Decommissioning," recommenos that the SAR contain a quantitative

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estimate of the activity of specific isotopes that must be disposed of as radioactive waste -

following shipment of the spent fuel off-site (i.e., activation of Concrete Cask liner and concrete).' The Trojan ISFSI SAR does not give estimated quantities of specific isotopes expected to remain after transfer of the spent nuclear fuel.

L Basis for Variation from SRP l

The TranStor storage casks are part of a modular storage and shipping system. The sealed

!- metal baskets containing the spent nuclear fuel are designed to be transferred intact to a licensed

. shipping cask and removed from the site for eventual disposal at a permanent spent nuclear fuel repository. Therefore, the structural components of the system located in closest proximity to the spent nuclear fuel, the sealed metal baskets, will be transferred off-site for disposal along with

' the fuel itself. Activation of the remaining concrete storage casks and concrete pad is expected to be minimal. PGE has not performed analyses to estimate the amounts of specific isotopes l

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expected as a result of activation. Because of the modular design of the system, however, such activity levels are expected to be low and any remaining radioactive waste will be packaged and disposed of as low-level radioactive waste in accordance with the provisions of 10 CFR 60.55.

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IV. Comparison of Trojan ISFSI with Standard Review Plan Guidance This section contains a discussion of the confonnance of the Trojan ISFSI to each chapter of the Standard Review Plan. The review of each Standard Review Plan Chapter contains a discussion of the Trojan ISFSI's degree of conformance to objective criteria or guidance contained in  ;

Standard Review Plan Section IV," Acceptance Criteria," and Section V," Review Procedures." l 1

SRP Chapter 1.0 -- General Description l As stated in the Standard Review Plan, the objective of Chapter 1.0 is to ensure that the applicant has provided a non-proprietary description of the facility that is adequate to familiarize the reviewers and other interested parties with the pertinent features of the storage system.

Section IV, " Acceptance Criteria," and Section V, " Review Procedures," in Chapter 1.0 are divided into identically numbered and titled subsections. Therefore, the requirements of the j various subsections of Section IV and Section V are discussed concurrently below. '

IV. Acceptance Criteria /V. Review Procedures 1

1. DCSS Description and Operational Features A broad overview of the Trojan ISFSI storage system is presented in SAR Sections 1.2 and 1.3.

The functions of components are discussed in Section 3.3.1 and the general functions of the storage system is discussed in Section 3.4. A more detailed description of the storage system components is provided in SAR Section 4.2.4 and 4.7.3. SAR Section 4.2.4 addresses the design life of the storage system and the use of valves and disconnects, lids, seals, and bolt closures. The weights of various components are provided in SAR Table 4.2-4 and materials of construction are discussed in SAR Sections 4.2.4,4.2.5, and 4.7.3. (Note: Some specific information related to the dimensions and materials of various components is considered to be proprietary

! information by SNC. This information has been provided to the NRC as proprietary information in accordance }vith 10 CFR 2.790.) The primary confinement boundary of the storage system is described in SAR Sections 3.3.2,4.2.3, and 4.2.4.

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Radiation shielding features of the storage system are discussed in SAR Sections 3.3.5 and 7.3.

Potential radioactive effluents are discussed in SAR Section 7.1.

The overall number aad arrangement of storage casks on the ISFSI storage pad are discussed in

, SAR Sections 1.3 and 7.1.2.

Lifling devices for the storage system components are discussed in SAR Sections 4.7.3 and 4.7.4.

It should be noted that the initial loading of spent fuel into the storage system baskets and the transfer of these loaded baskets into the concrete storage casks is performed in the Fuel Building of the Trojan Nuclear Plant (TNP). As such, the fuel handling aspects of these operations will be performed under the provisions of the TNP 10 CFR 50 license. PGE submitted detailed information relative to the lifting and handling of the fuel within the TNP Fuel Building in conjunction with a license amendment request to the 10 CFR 50 license (LCA-237). PGE also provided a copy of LCA-237 on the Trojan ISFSI Docket (Ref. VPN-006-98 dated 1/19/98).

A list of components classified as "important to safety"is provided in Section 3.3.3.1.

Additional detail relative to the conformance of the PGE Quality Assurance Plan to the guidance contained in the Standard Review Plan is contained in the review of SRP Chapter 13, " Quality Assurance," below.

2. Drmvings Illustrations of the storage system are presented in SAR Figures 1.3- 1, 2.1 -3, 4.2-1 through 4.2-8, and 4.7-6. Additional non-pmprietary drawings of key storage system components are included in SAR Appendix A " Drawings."

3. DCSS Contents The fuel and other radioactive wastes expected to be stored in the Trojan ISFSI are characterized in SAR Sections 1.2,1.3, and 1.3.1. The specific physical, thennal, and radiological characteristics of the material to be stored are presented in Sections 3.1.1.1 through 3.1.1.4.

PGE plans to store some degraded fuel at the Trojan ISFSI. The potential criticality of degraded fuel is addressed in SAR Sections 3.3.4 and 4.2.7. Retrievability and off-site transfer operations are addressed in SAR Section 5.1.1.6.

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4. Discussion ofOreanizational Roles As described in SAR Section 1.4, PGE is the principal owner and is responsible for fabrication, construction, operation, maintenance and surveillance of the Trojan ISFSI. Sierra Nuclear I Corporation is responsible for the design of the fuel storage system and for fabrication of some ,

auxiliary equipment. PGE's organizational responsibilities related to ISFSI construction, fuel  !

loading, and ISFSI operation are further discussed in SAR Sections 9.1.1 and 9.1.2. l l

5. Quality Assurance l PGE's Quality Assurance Program is described in SAR Chapter 11.0," Quality Assurance."

PGE's Quality Assurance Program is contained in PGE-8010," Trojan Nuclear Plant Nuclear Quality Assurance Program." PGE-8010 complies with 10 CFR 50, Appendix B," Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plams." in addition to 10 j CFR 50 activitics, PGE applies PGE-8010 to activities covered by Appendix H to 10 CFR 71, j

" Quality Assurance for Packaging and Transportation of Radioactive Material," and Subpart G to 10 CFR 72," Quality Assurance for Independent Storage of Speni Nuclear Fuel and High-Level Radioactive Waste."

SAR Section 11.0 also incorporates by reference SNC's Quality Assurance Plan which has previously been reviewed and approved by the NRC in conjunction with a previously licensed storage system.

6. Consideration of10 CFR Part 71 Reanirements Recardine Transportation As described in SAR Section 1.5, the Trojan ISFSI utilizes the storage portion of SNC's TranStor modular fuel storage and shipping system. SNC has separately applied for a Certificate of Compliance for the TranStor* Shipping Cask system (NRC Docket No. 71-9268) and the application is currently under review by the NRC. PGE intends to register as a user of the TranStor Shipping Cask once a Certificate of Compliance is issued by the NRC.

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SRP Chapter 2.0 -- Principal Design Criteria l

This SRP Chapter provides guidance to ensure that the principal design criteria for structures, systems and components important to safety comply with the relevant requirements of 10 CFR 72.

IV. Acceptance Criteria l

1. Structures. Systems and Components Imnortant to Safety Chapter 1 of the SAR provides a general overview of the storage system. SAR Chapter 3,

" Principal Design Criteria." identifies structures, systems and components important to safety.

Additional detail for individual components is mntained in SAR Chapter 4, " Installation Design."

)

i

2. Desien Basesfor Structures. Systems. and Components important to Saferv i

The principal design criteria for the ISFSI are detailed in SAR Chapter 3, " Principal Design Criteria." Chapter 4," Installation Design," contains specific information relative to how each  !

major component complies with these criteria.

The material to be stored at the Trojan ISFSI is limited to the spent nuclear fuel and related radioactive materials from the Trojan Nuclear Plant. The specific materials to be stored at the Trojan ISFSI are described in SAR Section 3.1.1," Materials to be Stored."

Applicable external conditions such as temperature extremes, earthquakes, tornados, etc. are detailed in SAR Chapter 3.

3. Desien Criteria for Safety Protection Systems Specific design criteria are described in SAR Chapter 3 with additional information related to specific components contained in SAR Chapter 4. Radiation protection features of the design are discussed in SAR Chapter 7 and general operating procedures are outlined in Chapter 5.

t Decommissioning of the ISFSI is discussed in SAR Section 9.8, " Decommissioning."

18 ,

l V. Review Procedures ,

J. Structures. Systems and Components hnnortant to Saferv l

The SAR identifies those components that are important to safety in Section 3.3.3.1,

" Equipment." The criteria used for classification ofitems as important to safety is discussed in SAR Section 3.4," Classification of Structures, Systems and Components." The basis for excluding certain structures, systems or components from this classification is found various S AR Sections discussing individual ISFSI components (e.g., classification of the concrete storage pad is discussed in SAR Section 4.2.1, " Structural Specification," and the classification of various lifling and handling equipment is discussed in Section 4.7.1, " Structural Specifications")

1

2. Design Basesfor Structures. Systems. and Components hnportant to Safety The material to be stored at the Trojan ISFSI is limited to the spent nuclear fuel and related l radioactive materials from the Trojan Nuclear Plant. The specific materials to be stored at the Trojan ISFSI are described in SAR Section 3.1.1," Materials to be Stored." This material includes both intact and damaged fuel assemblies, fuel pellets and fuel debris, control assemblies, burnable poison assemblies, and source assemblies. Failed fuel assemblies and fuel debris are

)

stored in a failed fuel can to contain this material in place to ensure subcritical conditions are maintained and that the thermal analyses remain valid.

l 1

Assumptions of fuel pin failures and fission product releases are consistent with the guidance in  !

the Standard Review Plan.  !

Accident conditions considered in SARChapter 8," Accident Analysis" are consistent with the l l

Standard Review Plan guidance and include:

• Cask Drops, Cask Tipover,

. Fire,

• Fuel Rod Ruptures, ,

Leakage of the Confinement Boundary, Explosive Overpressure,

=

Air Flow Blockage, 19 l

• Flood, ,

• - Tornado,

• ' Earthquake, -

-*. , Burial under Debris (volcano ash), and

(

• Lightning .

. 3. Design Criteriafor Saferv Protection Syster.u l

The Trojan ISFSI uses a passive design that does not rely on active safety protection systems.

i. 'The capability of the design to provide adequate confinement, cooling, suberiticality, fuel j retrievability and radiaiton protection is described in the review of applicable SRP Chapters.

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[

! The requirement for continuous monitoring in accordance with 10 CFR 72.122(h)(4) is met via i routine surveillance.~ This is discussed in further detail under SRP Chapter 7.0 below, p

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SRP Chapter 3.0 -- Structural Evaluation This SRP Chapter contains guidance related to the structural performance of the ISFSI components under normal, off-normal, and accident conditions and natural phenomena. The storage system must maintain confinement, suberiticality, radiation shielding and fuel retrievability under all credible loads.

IV. Acceptance Criteria

1. Confinement Cask As described in SAR Chapter 4," Installation Design," the steel storage baskets are designed to j Section III of the ASME Code (confinement boundary to Subsection NC, internals to Subsection NG) with exceptions noted in Table 4.2-1a.

2. Reinforced Concrete Structures Important to Safety. but not within the Scope ofACI-359.

The concrete storage casks are designed to the requirements of ACI-349 and ANSI 57.9 with exceptions noted in SAR Table 4.2-2a.  !

3. Other Reinforced Concrete Structures Subject to Approval 1

i The concrete storage and transfer pads are designed to ACI-318.

4. Other System Comnonents Important to Safety The load combinations for the storage cask are dermed in accordance with ANSI 57.9. The transfer cask and lifling yoke are designed in accordance with ANSI N14.6.

5. Other Components Subject to NRC Approval j None,

( 21 l

V. Review Procedure

1. Confinement Cask The steel storage basket has been evaluated for the l'uil range ofload combinations specified in ANSI 57.9 and shown not to deform in a manner that would jeopardize subcritical conditions or prevent retrievability of the fuel. Although there is no credible mechanism that could result in a cask tipover during storage conditions, this event was analyzed as a bounding condition in accordance with the guidance of the Standard Review Plan.

The principal radiation shielding during storage is provided by the metal storage basket and the outer concrete cask. No credible events result in significant degradation of this shielding and any degradation of the concrete cask would be readily observable.

The confinement boundary is fabricated of stainless steel and is not subject to brittle fracture.

The potential for metal fatigue failure is discussed in SAR Section 4.2.5.3.6," Basket Fatigue Evaluation." Fatigue effects on the basket are addressed using the criteria contained in ASME Section III, NC-3219.2. Fatigue analysis need not be performed provided the criteria of Condition A are met.

The Trojan ISFSI confinement boundary uses redundant closure welds. The welds are specified on design drawings in accordance with AWS standards. Materials of construction are per ASME Section III with deviations noted on Table 4.2-la. Weights and centers of gravity of key ISFSI components are given in SAR Table 4.2-4, " Nominal Weights and Centers of Gravity."

The SAR identifies applicable load combinations for normal, off-normal, and accident conditions. Consistent with the guidance in the Standard Review Plan, accidents evaluated in the SAR include a hypothetical cask tipover, postulated explosions, fire, flood (the Trojan ISFSI is  ;

located above the maximum credible flood level), tornados, and postulated earthquakes.

Finite element analyses described in the SAR are performed using the ANSYS computer code.  !

Closed form calculations were used for some simple analyses and as checks of the results of selected analyses.

I 22 I

r N --_ _ _

2. Reinforced Concrete Comvonents L The concrete casks are designed to the requirements of ACI-349. Temperature limits for the concrete casks are in accordance with alternative 2 described on page 3-21 of the Standard l Review Plan. Temperature limits are described in SAR Section 4.2.4.2.4, " Description of the Concrete Cask."

Applicable load combinations for the concrete casks are shown in SAR Table 4.2-10," Concrete Cask Load Combinations."

The concrete storage and transfer pads are designed to the requirements of ACI-318 with load combinations in accordance with ANSI 57.9.

3. Other System Components Important to Safety The Transfer Cask and Lifting Yoke are designed and fabricated in accordance with NUREG-l 0612 and ANSI N14.6. The Transfer Cask trunnions are fabricated in accordance with ANSI N14.6 requirements and are tested to 300% of their maximum design load.

l l_ The Transfer Station is designed and fabricated to the AISC Manual of Steel Construction.

4. Cher Components Subject to NRC Approval

! None.

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1

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23 l

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SRP Chapter 4.0 -- Thermal Evaluation This SRP Chapter provides guidance for an evaluation to ensure that the cask and fuel materials remain within allowable values or criteria for normal, off-normal, and accident conditions. This evaluation also serves to confirm that the temperature of the fuel cladding during the storage period will not result in cladding degradation that could lead to gross rupture.

IV. Acceptance Criteria The SAR demonstrates that fuel cladding (zircaloy) temperature at the beginning of dry cask storage are below the anticipated damage-threshold temperatures for normal conditions and a minimum of 20 years of cask storage. The storage temperature limit is selected to a very low probability (e.g.,0.5 percent per fuel rod) of cladding breach during long-term storage. Fuel cladding temperature for short-term accident conditions, short-term off-normal conditions, and fuel transfer operations (e.g., vacuum drying of the cask or dry transfer) remain well below the Standard Review Plan guidance of 570 C (1058 *F).

l l

The maximum internal pressure of the storage basket remains within its design pressures for normal, off- j normal, and accident conditions assuming rupture of 1 percent,10 percent, and 100 percent of the fuel rods, respectively. The pressure analysis is consistent with the recommended assumptions for pressure  !

calculations including release of 100 percent of the fill gas and 30 percent of the significant radioactive gases in the fuel rods. I l

Cask and fuel materials are maintained within their minimum and maximum temperature criteria for j normal, off-normal, and accident conditions in order to enable components to perform their intended j safety functions.

PGE has not performed a specific analysis to demonstrate that potential cladding damage resulting from creep cavitation is limited to 15 percent of the original cladding cross-sectional area during dry storage.

As discussed in Section III of this document, however, this is not considered a credible failure mechanism for the spent fuel to be stored at the Trojan ISFSI.

The cask system is passively cooled and the thermal performance of the cask is within the allowable design criteria specified in the SAR Chapter 3, " Principal Design Criteria," for normal, off-nor mal, and accident conditions.

i 24 I

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V. Review Procedures

1. Spent Fuel Cladding Analysis for the limiting Trojan ISFSI fuel was performed using the methodology of PNL-6364 as recommended by the Standard Review Plan and accepted by the NRC for previous cask designs. The thermal analyses described in S AR Section 4.2.6 demonstrate that the cladding  !

temperatures for the Trojan ISFSI will be below the calculated fuel cladding limiting temperature l I

as shown on SAR Table 4.2-12.

I1 Specific information related to Trojan peak fuel internal pressures and temperatures was also provided via PGE letter VPN-057-97, dated July 31,1997, and is addressed in response to NRC'c RAl-2, Question 3-5.

Explicit analyses of diffusion controlled cavity growth (DCCG) has not been performed as recommended by the Standard Review Plan. As discussed in the summary of variations from Standard Review Plan guidance in Section III of this review, DCCG is not considered a credible failure mechanism given the cladding stress values and storage temperatures at the Trojan ISFSI.

I Short term cladding temperatures are limited to below 1058 F as recommended by the Standard l l

Review Plan Conservative calculations of cladding temperatures during vacuum drying operations are addressed in LCA-237 and are reflected in SAR Table 4.2-12. These calculations indicate that the calculated peak cladding temperatures during vacuum drying operations and I other short-term operations or off-normal events are well below the short term limit of 1058 F.

The proposed Trojan ISFSI Technical Specifications limit the air outlet temperature to ensure that storage system components do not exceed applicable temperature limits. The Standard Review Plan also recommends that the Technical Specifications limit the maximum time fuel can be submerged in a basket that has not been evacuated and sealed. The proposed Technical Specifications include such limitations.

2. CaskSystem ThermalDesien

\

As described in SAR Section 4.2.6, the Trojan storage cask system relies solely on passive h- at removal. An air outlet temperature monitoring system is provided as discussed in SAR Section 25 I

1

( _ _.. _ . . _ _ _ . . . . _ _ _ _ . _ _ _ _ _ _ . . _ . _ _ . _ . _ _ . . _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _

5.1.3.4, however, this system is not classified as important to safety since air outlet temperature l monitoring can be performed manually.

3. TirermalLoad Specificationhimbient Temperature  !

The Trojan storage cask system is designed for a maximum of 26KWt. The proposed Technical Specifications conservatively limit the maximum heat load in a single basket to 24KWt. The )

actual planned maximum basket loading is less than 18KWt. Ambient temperatures selected for use in thermal analyses are based on historical meteorological data for the Trojan site and vicinity.

SA.R Section 5.1.1.2 describes the methodology used to preclude boiling of the water in a basket during fuel loading operations.

Thermal calculations have not been performed to support potential reflooding of a dry basket.

The TranStor storage system is part of a modular system that provides for storage and shipping of the spent nuclear fuel without the need for repackaging the fuel or reliance on the Trojan Nuclear Plant Fuel Pool. The Trojan ISFSI is provided with a dry transfer facility (Transfer l

Station) and shielded Transfer Cask that allow a scaled metal storage basket to be removed from the concrete storage cask and loaded into a shipping cask for transportation off-site.

In the unlikely event of a failure of a storage basket confinement boundary during storage, the Trojan ISFSI design provides for the use of a metal overpack to restore the confinement boundary. A leaking storage basket can be removed from the concrete storage cask and placed in l an overpack using the dry transfer facility without the need for unloading the spent nuclear fuel from the failed storage basket.

The Standard Review Plan does not contain explicit review criteria for dry transfer facilities such as the Trojan ISFSI Transfer Station. The Standard Review Plan notes that such guidance will be developed at a later date.

4. ModelSpecification 1 f The thennal analyses models are described in SAR Section 4.2.6.2,4.2 6.3, and 4.2.6.4. Thermal j interaction of the casks array is addressed in SAR Section 4.2.6.4.2. Temperature limits for j various materials are discussed in SAR Section 4.2.6.

26

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! Appropriate cask boundary conditions are specified for the cask during normal, off normal and accident events.

1

The effects of postulated fires are addressed in SAR Section 8.2.14.2.2 -

5. Thermal Anah' sis

' Thermal analyses for the Trojan ISFSI storage system were performed using ANSYS finite element analyses. Input and output files have been provided to the NRC staff when requested in conjunction with the submittal of various proprietary SNC calculations.

6. SunnlementalInformation

. Supplemental information related to the thermal analysis of the storage system has been provided in response to various requests for additional information.-

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27 L : _ - _- _ - - - _ - - - - - - - - - - - - - - - - - - - - - . - . - - - - - - - - - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - .

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SRP Chapter 5.0 -- Shielding Evaluation This SRP Chapter seeks to ensure that the proposed shielding features of the storage system provide adequate protection against direct radiation from the cask contents and that calculated radiation levels are within regulatory limits for normal, off-normal, and accident conditions.

IV. Acceptance Criteria The Trojan ISFSI conforms to the acceptance criteria contained in this SRP Chapter as described

-l in the " Review Procedure" section below. Specifically:

I a

the ISFSI controlled area boundary is at least 100 meters a

the ISFSI complies with the radiation dose requirements of 10 CFR 72.104(a)

=

dose rates are consistent with ALARA principals a

the design basis accident radiation dose does not exceed 5 rem to the whole body or any organ, and the shielding features ensure compliance with occupational and radiation dose limits for individual members of the public as prescribed in 10 CFR 20, Subparts C and D.

V. Review Procedures l

i

1. Shieldine Design Description l

The shielding design of the storage system is described in SAR Section 7.1.2. The design maximum radiation dose rates for the storage casks are 200 mrem /hr on the cask top and 100 i mrem /hr on the cask sides. These design values are well within the range of past accepted values l cited in the Standard Review Plan.

2. Radiation Source Definition The radiation source terms for shielding analyses are described in SAR Section 7.2. Shielding analyses were performed for two design basis cases,40,000 mwd - 5 year cooled fuel, and 45,000 mwd - 6 year cooled fuel. The entire Trojan spent fuel inventory is bounded by these

. two cases with respect to burnup level and cooling time. For each case, initial enrichment levels 28 L__-____-- - - - - _ _ _ - - - _ - _ - - - - _ - - _ _ - - - _ _ _ - - - -

i are assumed which bound the lowest actual enrichment levels, since this yields the maximum

! gamma and neutron source terms for fuel of a given burnup level.

l w

l As described in SAR Section 7.2, the gamma and neutron source terms are defined as a function of energy. The source term defm' ition also includes the quantities of various radionuclides necessary for the input to accident release calculations.

3. Shieldine ModelSpecification l

I The model used for shielding calculation is described in SAR Section 7.3. The models used to calculate the Trojan ISFSI normal and accident doses provide the assumptions, input data and methods with sufficient information to allow a reviewer to understand the results and check the method used. The model homogenizes only the fuel region; the top and bottom fuel nozzles are modeled separately. The Trojan ISFSI source model uses a 1.1 axial peaking factor and the non-fuel top and bottom assemblies are modeled for dose calculations. l The modeled receptor locations are appropriate for determining potential occupational and/or l

public doses. General area dose rates for detemlining worker exposure and dose versus distance from a single cask and from the array of casks on the pad were also calculated.

PGE addressed the potential effects of elevated temperatures on the neutron shielding material utilized in the Transfer Cask in response to NRC's RAI-2, Question 7-2. PGE's response notes  ;

that the shielding calculations were revised to conservatively bound the potential loss of hydrogen from the shielding material due to elevated temperatures.

4. Shieldine Analvsis The computer codes used for shielding analyses are discussed in SAR Section 7.3. These are industry standard codes such as those listed in the Standard Review Plan.

Details of the various shielding analyses are contained in proprietary SNC calculations which PGE has provided to the NRC for review.

l 29

5 Stapplementalinformation Supplemental information related to the shielding analyses has been provided to the NRC in response to various requests for additional information.

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SRP Chapter 6.0 -- Criticality Evaluation f This SRP Chapter directs the NRC reviewer to ensure that the applicant has demonstrated that the spent fuel remains suberitical under normal, off-normal, and accident conditions involving handling, packaging, transfer, and storage.

IV. Acceptance Criteria The Standard Review Plan guidance states that the multiplication factor (km) should not exceed i 0.95 under all credible conditions, including all biases and uncertainties at a 95-percent confidence level. At least two unlikely, independent events should be required before an accidental criticality is deemed possible. The Standard Review Plan also recommends that criticality safety should be based on favorable geometry and/or fixed neutron poisons and not take credit for fuel burnup or bumable neutron absorbers. i The Trojan ISFSI design complies with these acceptance criteria as discussed below.

V. Review Procedures

1. Criticality Design Criteria and Features i

Storage system criticality during dry storage is principally discussed in SAR Section 4.2.7. The dry storage criticality analysis relies solely on the favorable geometry provided by the storage system design. Although fixed neutron absorbing material is included in the TranStor basket design, the Trojan ISFSI criticality does not credit this material. The storage system is designed to maintain subcritical conditions for credible accidents. In order to invalidate criticality assumptions, the storage basket volume would have to be filled with water. In order for this condition to occur, the ISFSI location would require a flood in excess of the design basis flood l scenario coincident with confinement barrier failure. Neither of these conditions is considered credible based on the accident analysis presented in SAR Chapter 8.

PGE has also evaluated the potential for criticality during fuel loading operations in the Trojan Fuel Building. This analysis is described in Revision 2 to PGE's license change application (LCA) 237 to the Trojan Nuclear Plant's 10 CFR 50 license submitted on January 19,1998 31

(VPN-006-98). This analysis does credit the fixed neutron absorbing material within the storage baskbt and demonstrates that adequate subcritial conditions exist even assuming the loaded fuel basket is flooded with unborated water, i

2. FuelSpecification As noted in SAR Section 4.2.7, the Trojan ISFSI criticality analysis is performed assuming the ,

maximum fuel enrichment that conservatively bounds the fuel to be stored at the facility (the

)'

maximum initial fuel enrichment of the fuel available at the facility) and does not credit the negative reactivity effects of fuel burnup. The analysis also conservatively addresses the impact of degraded fuel or fuel debris on the criticality analysis.

3. ModelSpecification Details of the criticality analyses described above are contained in SNC proprietary calculations - l listed below: l l

-* PGE01-10.02.02-01, Revision 1, Dry TranStor Basket Criticality

• . PGE01-10.02.02-03, Revision 1, TranStor Basket Criticality. Analysis

' Current revisions of these calculations were submitted to the N'RC in conjunction with Revision -

2 to LCA-237.

4. Criticality Analvsis SAR Section 4.2.7 describes the criticality analysis for dry storage conditions. This analysis was performed using the KENO-Va module of the SCALE-4.1 computer code. This same computer I

code was used to perform the criticality evaluation for fuel loading operations. The details of these analyses are contained in the SNC proprietary calculations noted above.

5. SupplementalInformation PGE has provided additional supplemental information as requested by the NRC staff during l review of the Trojan ISFSI license application.

l 32  !

l i SRP Chapter 7.0 -- Confinement Evaluation l \

This SRP Chapter provides guidance for evaluating the confinement features of the storage system. The confinement features ensure that radiological releases to the environment will be within regulatory limits and that the spent fuel cladding will be protected from potential degradation during storage.

1 IV. Acceptance Criteria l

The Trojan ISFSI uses a metal storage basket as the principal confinement boundary. The basket is designed and fabricated to appropriate codes and standards as described in the review of SRP l I

Chapter 3.0 above. The storage basket is sealed by redundant welded closures. The integrity of the basket closure welds is verified by helium leak testing and the maximum permissible leakage rate is specified in the proposed Trojan ISFSI Technical Specifications. The fuel is maintained in an inert environment within the basket during storage to prevent potential degradation of the fuel cladding. ,

1 V. Review Procedures

1. Confinement Desien Characteristics The sealed metal storage basket provides the principal confinement boundary. The basket is sealed by redundant seal welds as described in S AR Section 4.2.4.2.1. A basket overpack can be used in the unlikely event of a failure of the storage basket confinement. The overpack is also provided with redundant seal welds. The overpack is described in 4.2.4.2.3.

The storage basket is vacuum dried to remove any residual moisture and backfilled with helium prior to seal welding (Ref: SAR Section 5.1.1.2). The maximum allowable helium leakage rate specified in the proposed Trojan ISFSI Technical Specifications is sufficiently small to ensure that the inert atmosphere will be maintained during the storage period (Ref: PGE Response to RAI-2, Question 10-2(d)).

33

2. Confinement Monitoring Capabilitv The Trojan ISFSI storage baskets are completely closed by welding and, therefore, no monitoring system is required by the guidance in the Standard Review Plan.

3. Nuclides with Potential for Release The Trojan ISFSI dose calculation did not assume that all the isotopes noted on SRP Table 7-1 would be released in the case of a hypothetical accident. The Trojan ISFSI fuel storage baskets are stored at approximately atmospheric pressure. Therefore, a breach in the confinement barrier would not result in the needed motive force to release nonvolatile elements. Therefore, the calculation considered the dose from "Kr and 3H.

4. Confinement Analysis PGE has conservatively evaluated the non-mechanistic failure of a basket confinement boundary coincident with the complete failure of 100% of the fuel pins. This conservative analysis served to define the controlled area boundary for the Trojan ISFSI. This analysis is described in SAR Section 8.2.1.2.

5. SunnlementalInformation Supplemental information related to the confinement system has been provided as requested by the NRC in response to requests for additional information.

l I.

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SPR Chapter 8.0 -- Operating Procedures SRP Chapter 8 provides guidance to ensure that the applicant has provided acceptable operating sequences, guidelines, and generic procedures for key ISFSI operations.

IV. Acceptance Criteria PGE's application is for a site specific license in accordance with 10 CFR 72. As such, detailed operating procedures for the Trojan ISFSI will be developed under the provisions of the proposed ISFSI Technical Specifications. Procedures for the initial fuel loading will be developed and I controlled under the provisions of the Trojan Nuclear Plant's 10 CFR 50 license.

Chapter 5," Operations," of the ISFSI SAR describes the methods and sequences operational controls which personnel performing spent fuel loading and storage activities will implement to i

assure that operations utilize the passive safety features of the Trojan ISFSI design as described in the SAR. Fuel loading and basket sealing operations (including nondestructive examination and pressure testing) will be perfomled within the Fuel Building in order to utilize the existing systems and equipment for heavy lifts, radiation monitoring and controls, decontamination and any necessary auxiliary support (i.e., electrical, crane, service air, etc.). Fuel handling and cask loading operations in the Fuel Building will be performed in accordance with Portland General Electric Company's 10 CFR 50 license for the Trojan Nuclear Plant. However, spr. ific restrictions related to the basket loading operations are also included in the proposed ISFSI Technical Specifications. Information related to the fuel handling and loading operations is also l provided in a license change application to the TNP's 10 CFR 50 license (LCA-237).  !

1 V. Review Procedures 1

1. CaskLoading l j Cask loading operations are primarily described in SAR Chapter 5, " Operations," and in LCA-237 to the TNP 10 CFR 50 license. Radiation protection and ALARA programs are discussed in  !

SAR Chapter 7," Radiation Protection." The SAR and proposed ISFSI Technical Specifications address the major elements listed in the Standard Review Plan, including:

35 l

fuel specifications,

= ALARA,

= off-site releases, a

draining and drying, a

filling and pressurization, and a

welding and sealing.

The Standard Review Plan guidance on cask draining and drying operations relies heavily on PNL-6365. Although this document is not directly referenced in the Trojan ISFSI licensing basis documents, the draining and drying operations described in the SAR generally comply with the recommendations of PNL-6365 and the Standard Review Plan with one noted exception. The Standard Review Plan recommends that the cover gas in each cask be sampled after backfilling to verify purity. The control of helium purity at the Trojan ISFSI is controlled by quality inspection prior to receipt. Trojan has committed to using 99.999% Grade 5 helium vs. 99.995%

weld grade as tested in PNL-6365. Sampling of helium after the final backfill is not necessary based on the following considerations:

• The vacuum drying system is designed to preclude inadvertent in-leakage, No inadvertent operation can allow oxidizing gases into the basket.

The purity level of the helium gas is controlled by inspection prior to receipt.

. The vacuum drying system is operated by trained and certified operators using approved procedures.

2. Cask Handling and Storage Operatiwn Cask storage operations are primarily described in SAR Chapter 5, " Operations." Additional l

information related to handling of the fuel storage basket within t'.ie TNP fuel building is provided in LCA 237 to the TNP 10 CFR 50 license.

i

3. Cask Unioadine The Trojan ISFS1 utilizes the TranStor storage system. The TranStor storage system is part of a modular system that provides for storage and shipping of the spent nuclear fuel without the need for repackaging the fuel or reliance on the Trojan Nuclear Plant Fuel Pool. The Trojan 36 w__-___-____ . _ - - _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ _ _ _ - _ . _ _ _ _ _ ____ _ _ _ _ _ _ _ _ _ _

ISFSI is provided with a dry transfer facility (Transfer Station) and shielded Transfer Cask that allow a sealed metal storage basket to be removed from the concrete storage cask and loaded into a shipping cask for transportation off-site.

In the unlikely event of a failure of a storage basket confinement boundary or other unforseen problems during storage, the Trojan ISFSI design provides for the use of a metal overpack to restore the confinement boundary (Ref: S AR Section 4.2.3 and 4.2.4). 'A leaking storage basket can be removed from the concrete storage cask and placed in an overpack using the dry transfer facility without the need for unloading the spent nuclear fuel from the failed storage basket.

The Standard Review Plan does not contain explicit review criteria for dry transfer facilities such as the Trojan ISFSI Transfer Station. The Standard Review Plan notes that the NRC will develop such guidance at a later date.

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SRP Chapter 9.0 -- Acceptance Tests and Maintenance Program This SRP Chapter seeks to ensure that the applicant's SAR includes appropriate acceptance tests and maintenance programs for the storage system.

IV. Acceptance Criteria The goveming codes and standards for the Trojan ISFSI components are identified in SAR Section 4.2.1," Structural Specification." The cited codes and standards are consistent with those listed in the Standard Review Plan. Deviations from these codes and standards are clearly identified in the SAR consistent with the Standard Review Plan guidance (.Ref: SAR Table 4.2-1a, "ASME Code Deviations," and Table 4.2-2a, " Concrete Cask Code Deviations").

V. Review Procedures

1. Acceptance Tests Required visual and nondestructive examination inspections are discussed in SAR Chapter 4,

" Installation Design," and are specified on appropriate design documents and drawings. As

. noted in the SAR, the NDE of the confinement boundary closure welds is not performed using volumetric examinations as specified in the goveming code. These redundant closure welds are instead examined using s combination of hydrostatic testing, helium leak rate testing, and dye penetrant examinations. These attemative examinations are consistent with the guidance provided in the Standard Review Plan. The NDE of the closure welds is described in SAR ,

Section 3.3.2.2,"PWR Basket Closure Welds." Although not specifically required by goveming codes or the guidance in the Standard Review Plan, PGE will perform separate helium leak rate testing on each of the redundant basket closure welds. l l

The effectiveness of the cask shielding system is verified during startup testing for each storage cask and acceptance criteria are included in the proposed ISFSI Technical Specifications. 'Ihe l neutron shielding material used in the transfer cask will be tested prior to fabrication.

The storage baskets also contain a fixed neutron absorber. However, the fixed neutron absorber is not credited for criticality control during storage at the Trojan ISFSI. In addition, PGE has i i

38 L_____-_________________-__-_-____-_. _ _ ___ - _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _

shown in response to NRC RAl-1, Question 1-21 that the neutron absorber material will not be I significantly affected by the expected fuel storage temperatures or the expected radiation exposure. The use of this material meets the guidance of SRP Section 6.V.3.b (p. 6-4) for demonstrating the continued efficacy of the material by design and material properties in lieu of a surveillance or monitoring program.

Nonetheless, PGE has also requested an exemption from this requirement based on the acceptable criticality analyses without credit for the neutron absorbing material (Ref: PGE letter VPN-023-91 dated March 20,1997).

The thermal performance of each cask is also verified during startup testing.

The ISFSI pre-operational and startup testing program is described in SAR Section 9.2, " Pre-Operational and Startup Testing." PGE has requested an exemption from the requirement of 10 CFR 72.82(e) concerning the time frame for submitting a pre-operational test report (30 days prior to fuel loading). PGE is seeking this exemption since the 30-day time requirement could unnecessarily delay fuel loading operations and the NRC is expected to directly observe the pre-operational testing (Ref: PGE Letter VPN-012-98 dated February 10,1998). j

2. Maintenance Program The Trojan ISFSI is designed to require minimal maintenance. Periodic visual inspections, temperature monitoring and radiation level monitoring are addressed in the proposed ISFSI Technical Specifications.

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l SRP Chapter 10.0 -- Radiation Protection

This SRP Chapter provides guidance for evaluation of the radiation protection capabilities of the proposed cask system.

Section IV, " Acceptance Criteria," and Section V, " Review Procedures," in Chapter 10.0 of the Standard Review Plan are divided into identically numbered and titled subsections. Therefore,

l the requirements of the various subsections of Section IV and Section V are discussed concurrently below.

. IV. ' Acceptance Criteria /V. Review Procedures 1 Design Criteria

- As described in ISFSI SAR Section 7.1, the Trojan ISFSI design is based on guidance contained l: ' in Regulatory Guides 8.8 and 8.10. The Trojan ISFSI calculated radiation doses include both I direct and potential releases from normal operations, off-normal events, and accidents. The controlled area boundary is based on committed doses associated with a maximum hypothetical l

event that is non-mechanistic and considered beyond the design basis. This analysis shows compliance with the dose limits of 10 CFR 72.104(a) and 106(b). (

Reference:

SAR Sections  !

7.6.2,8.2.1 and 8.2.4.3) l

2. OccupationalExposure i.

L The Trojan ISFSI occupational doses are based on 10 CFR 20 limits as required. PGE does not

. employ minors to work in occupations receiving radiation exposure. The Trojan Radiation i l

l Protection Program includes dose limits for occupationally exposed personnel, minors, fertile ,

l females, and declared pregnant workers and members of the public. Anticipated doses associated L with cask loading and monitoring are evaluated in Chapter 7 of the ISFSI SAR (Ref: SAR Sections 7.4 and 7.5.3.2).

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3. Public Exposure The SAR evaluates potential doses to the public from normal operations and accident conditions against the limits of 10 CFR 72.104(a) and 106(b). The distance for control of exposure is greater than 100 meters as required (325 meters). (Ref: SAR Section 7.6.2)

The SAR contains specific dose analyses to prove that anticipated radiation levels are less than specified regulatory limits. No other fuel cycle facility is located in the vicinity of the Trojan ISFSI (following temiination of the Trojan Nuclear Plant 10 CFR 50 license) to add to the radiation dose a member of the public would receive. Trojan Nuclear Plant decommissioning activities do not add a substantial dose to a member of the public as defined in the Trojan Offsite Dose Calculation Manual and as reported in the Trojan Annual Environmental Report. (Ref:

SAR Section 7.6.2) j i

1 Only airborne releases are considered in calculating ISFSI accident radiation exposures. The storage system uses a dry cask system and there are no expected liquid radioactive effluent

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releases associated with the Trojan ISFSI. Both fission products contained in the fuel and potential surface activity on the basket exterior have been evaluated as described in SAR Sections 8.1.3.1 and 8.2.1.

4. ALARA The radiation protection features and controls described in the SAR were developed using the existing Trojan Radiation Protection Program, which includes a comprehensive ALARA j program that directs the evaluation of engineering and operational controls to reduce radiation doses. The Trojan ALARA program includes detailed reviews of proposed design changes to ensure the changes will not increase projected operational doses and that the design change is installed with ALARA in mind.

The Trojan Radiation Protection Program includes a written policy that imphaents management's commitment to dose reduction for occupational and public exposures. The Trojan ALARA program specifies PGE's commitment to Regulatory Guides 8.8 and 8.10. (Ref: SAR Sections 7.1.1, 7.1.2, and 7.1.3).

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SRP Chapter 11.0 - Accident Analyses The Standard Review Plan directs the NRC reviewer to evaluate the applicant's identification and analysis of potential hazards to ensure that all credible accidents are identified and analyzed for adequate safety performance of the storage system and conformance with applicable regulatory requirements.

IV. Acceptance Criteria

1. Dose Limitsfor OIT-NormalEvents SAR Section 7.5.3.2 notes that procedures for personnel radiation protection are prepared consistent with the requirements of 10 CFR 20. The Trojan ISFSI also complies with the regulatory requirements for off-site radiation exposure to members of the general public due to direct radiation as described in SAR Section 7.6. There are no anticipated discharges of radioactive materials during normal ISFSI operations.

Radiation releases as a result of postuhted off-normal events are described in SAR Section 8.1.3.

The resultant radiation dose is well within regulatory limits.

2. Dose Limitsfor Desien-Basis Accidents SAR Section 8.2.1 discusses the potential radiological iinpact of a hypothetical failure of all fuel pins within a storage basket and the simultaneous complete failure of the basket confinement barrier. The resulting calculated dose complies with the limits delineated in the Standard Review Plan.

3. Criticality L SAR Section 4.2.7 discusses the potential for criticality during fuel storage. The calculated value of k,is well below the value of 0.95 specified in the Standard Review Plan. This analysis was conducted for dry storage conditions. Water moderation due to flooding of the ISFSI is not a credible event for the Trojan ISFSI due to the elevation of the site above postulated flood levels.

An evaluation of potential criticality during fuel loading events in the TNP Fuel Building was 1

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submitted to the NRC in conjunction with a license change application (LCA-237) to the TNP 10 CFR 50 license. This information was also provided on the 10 CFR 72 ISFSI docket (VPN-006-98 dated 1/19/98). The fuel loading criticality evaluation also showed confom1ance to the guidance of the Standard Review Plan.

4. Confinement The accident analyses presented in Chapter 8 of the SAR demonstrate that the confinement system will remain intact during all credible postulated events. A postulated non-mechanistic failure of the confinement boundary is evaluated to bound the potential impact of accidental radioactive releases from the ISFSI.

5. Retrievability Storage systems must be designed to allow ready retrieval of spent fuel for further processing or disposal. Spent nuclear fuel is stored within seal welded closure baskets. The ISFSI is designed to allow retrieval of the baskets and placement into a transportation cask. SAR Section 5.1.1.5

' discusses operations associated with retrieval and shipment of spent nuclear fuel.

6. Instrumentation The Trojan ISFSI uses a passive cooling design that does not rely on instrumentation or control systems to remain ftmetional under accident conditions.

V. Review Procedures

1. Cause ofEvent SAR Chapter 8 includes a subsection discussing the postulated cause of each analyzed event.

Note that some events are non-mechanistic events that are analyzed for regulatory purposes only and do not have a credible cause or initiating event.

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2. Detection ofEvent SAR Chapter 8 includes a subsection discussing the detection of each off-normal event analyzed i in SAR Section 8.1. A specific subsection discussing the detection of events is not included in the discussion of accidents in Section 8.2 Many of the postulated accidents analyzed in SAR Section 8.2 are non-mechanistic, beyond design bases events that are analyzed for regulatory purposes only. Such events are not postulated to occur and no discussion of specific methods of j detection is needed (e.g., failure of all fuel pins simultaneous with confinement barrier failure, storage cask tipover, and basket pressurization). Accidents such as natural phenomena (i.e.,

tornados, earthquakes, and volcanos) would be readily apparent. Discussion of the detection of other postulated accidents, such as overpack and off-site shipping events in SAR Section 8.2.13, is included within the body of the SAR evaluation.

3. Summarv ofEvent Conscauences and Regidatorv Comnliance

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SAR Chapter 8 demonstrates that the consequences of all postulated accidents are within the applicable regulatory limits for radioactive releases. Bounding non-mechanistic events are analyzed to demonstrate that these regulatcry limits are met for any credible accident.

4. Corrective Course ofAction SAR Chapter 8 includes a subsection discussing corrective actions for each off-normal event analyzed in SAR Section 8.1. A specific subsection discussing the corrective actions is not included in the discussion of accidents in Section 8.2. Many of the postulated accidents analyzed in SAR Section 8.2 are non-mechanistic, beyond design bases events that are analyzed for regulatory purposes only. Such events are not postulated to occur and no discussion of specific l corrective actions is needed (e.g., failure of all fuel pins simultaneous with confinement barrier failure, storage cask tipover, and basket pressurization). The analyses of a number of other accidents show acceptable results such that no specific corrective action is required (e.g.,

earthquakes). Where appropriate, corrective actions are discussed in the body of the evaluation  ;

of the postulated accidents (e.g., the repair of concrete spalling due to a postulated tornado missile is discussed in SAR Section 8.2.4.3).

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SRP Chapter 12.0 - Conditions for Cask Use -- Operating Controls and i l

Limits or Technical Specifications  !

l The Standard Review Plan directs the NRC reviewer to evaluate the Technical Specifications proposed by the applicant to ensure that the Technical Specification contain sufficient operating limits and controls, monitoring instruments and control settings, surveillance requirements, j design features, and administrative controls to ensure safe operation of the storage system.

IV. Acceptance Criteria /V. Review Procedures The Standard Review Plan does not contain explicit guidance or acceptance criteria relative to the format or content of Technical Specifications for a site-specific ISFSI license application q such as the Trojan ISFSI. PGE submitted proposed Technical Specifications for the Trojan )

ISFSI as part of the original license application. Discussions with the NRC staffindicate that the I NRC has adopted the Technical Specifications issued in conjunction with the recent North Anna ISFSI license as a standard for the format and content ofISFSI Technical Specifications. As a  ;

result of these discussions, PGE has significantly revised the proposed Technical Specifications for the Trojan ISFSI to conform as closely as practicable to the recently approved North Anna ISFSI Technical Specifications. These revised Technical Specifications are being submitted under separate cover.

These revised Trojan ISFSI Technical Specifications contain:

1. Functional and Operating Limits restricting the materials to be stored at the ISFSI and limiting the storage cask air outlet temperature.

2. Limiting Conditions for Operation and Surveillance Requirements addressing:

- storage basket vacuum drying l

• storage basket backfilling with inert gas

. storage basket closure weld leakage limit

. storage cask air outlet temperature e storage cask surface radiation dose limits e storage cask surface contamination limits

. ISFSI perimeter radiation dose rate limits 45 I

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3. Design Features addressing the site characteristics and maximum storage capacity of the ISFSI

4. Administrative Controls addressing:

organizational responsibilities

. staffqualifications

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procedures and programs l

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I SRP Chapter 13.0 -- Quality Assurance This SRP Chapter directs the NRC reviewer to evaluate the applicant's Quality Assurance Program to ensure that adequate controls are in place over activities related to the design, fabrication, assembly, testing and use of storage system structures, systems, and components that are important to safety, i

IV. Acceptance Criteria /V. Review Procedures PGE's Quality Assurance Program is described in SAR Chapter 11.0," Quality Assurance."

PGE's Quality Assurance Program is contained in PGE-8010," Trojan Nuclear Plant Nuclear Quality Assurance Program." PGE-8010 complies with 10 CFR 50, Appendix B," Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants." In addition to 10 CFR 50 activities, PGE applies PGE-8010 to activities covered by Appendix H to 10 CFR 71,

" Quality Assurance for Packaging and Transportation of Radioactive Material," and Subpart G to 10 CFR 72," Quality Assurance for Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste."

PGE's letter of October 9,1995 (VPN-054-95) notified the NRC of PGE's intent to apply the Trojan 10 CFR 50, Appendix B program to ISFSI related activities as allowed by 10 CFR 72.140(d). PGE applies the Quality Assurance Program to activities affecting ISFSI equipment j classified as impmtant to safety. These components are identified in S AR Section 3.3.3.

I As noted in Appendix B of PGE-8010, PGE applies all eighteen criteria to ISFSI related activities along with additional requirements unique to the ISFSI that are consistent with 10 CFR 72, Subpart G. Scotion 2.2.1 of the Quality Assurance Plan addresses the application of a  !

" graded approach" to quality assurance requirements.

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i SRP Chapter 14.0 -- Decommissioning l

This SRP Chapter directs the NRC reviewer to ensure that the applicant has provided a conceptual decommissioning plan for NRC evaluation such that there is a reasonable assurance that the ISFSI can be decontaminated and decommissioned in a manner that adequately protects  ;

the health and safety of the public. l IV. Acceptance Criteria i

The Standard Review Plan notes that the decommissioning plan should fulfill the criteria of Regulatory Guide 1.86 and meet the requirements for classification and disposal of wastes contained in 10 CFR 61.55. The conformance of the Trojan ISFSI to these criteria is discussed under SRP Section V below. l l

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V. Review Procedures  !

The decommissioning plan for the Trojan ISFSI is described in SAR Section 9.8,"ISFSI Decommissioning Plan." As noted in the SAR, the ISFSI storage system is designed to minimize the decontamination efforts required for decommissioning. The spent fuel is confined within a sealed metal basket and steps are taken to minimize the potential surface contamination of these baskets pr;or to loading in the concrete storage casks. There are no normal radioactive effluents associated with operation of this system.

The TranStor storage casks are part of a modular storap and shipping system. The sealed metal baskets containing the spent nuclear fuel are designed to be transferred intact to a licensed shipping cask and removed from the site for eventual disposal at a permanent spent nuclear fuel repository. Therefore, the structural components of the system located in closest proximity to the spent nuclear fuel will be transferred off-site for disposal along with the fuel itself. The activation of the remaining concrete casks and concrete pad is expected to be minimal. PGE has not performed analyses to estimate the amounts of specific isotopes expected as a result of activation. Because of the modular design of the system, however, such activity levels are expected to be low and any remaining radioactive waste will be packaged and disposed of as low-level radioactive waste in accordance with the provisions of 10 CFR 60.55.

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