ML20235A907

Forwards SER Accepting Proposals to Extend Surveillance Test Intervals & Allowable out-of-svc Time for Reactor Protection Sys/Anticipatory Reactor Trip Sys as Described in Suppl 1 to BAW-10167.CRGR Review Should Be Waived

ML20235A907
Person / Time
Issue date:	09/30/1988
From:	Sniezek J Office of Nuclear Reactor Regulation
To:	Jordan E Committee To Review Generic Requirements
Shared Package
ML20235A909	List: ML20235A907 ML20235A936 ML20235M383
References
GL-83-28, NUDOCS 8810140165
Download: ML20235A907 (69)

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UNITED STATES *

[y),e ' q NUCLEAR REGULATORY COMMISSION

-J aC WASHINGTON, D. C. 205SS jj i M(

'c:, ~ f September 30, 1988 M ' '

MEMORANDUM FOR: Edward L. Jordan, Chairman Committee to Review Generic Requirements FROM: James H. Sniezek, Deputy Director Office of Nuclear Reactor Regulation

SUBJECT:

REQUEST FOR CRGR REVIEW 0F B&W TOPICAL REPORT BAW-10167 AND SUPPLEMENT 1, " JUSTIFICATION FOR INCREASING THE REACTOR TRIP SYSTEM ON-LINE TEST INTERVALS" Enclosed is a safety evaluation prepared b extend surveillance test intervals (STIs) y NRR for acceptance of proposals to ano allowable out of service time (A0Ts) for Reactor Protection System / Anticipatory Reactor Trip System (RPS/ ARTS) instrument strings in B&W reactors as described and justified in BAW-10167 and Supplement 1.

We propose that CRGR review of these proposed extensions be waived.

BAW-10167 and Supplement 1 propose extending the STIs for RPS/ ARTS instrument strings (defined as the RPS equipment from the sensor through the bistable) from one to six months (with a staggered test schedule) and justifies this proposal using reliability and probablistic risk analysis methodology similar to that previously reviewed by CRGR for BWR plants at the July 8, 1987 meeting (No. 118). The staff SER would grant this extension such that individual utilities could voluntarily request changes to their technical specifications.

While the methodology and nature of the request are similar to those previously approved for GE and Westinghouse, this six month staggered interval would be unique to B&W in that a cuarterly test interval was requested and approved for GE and Westinghouse reactors.

However, we have reviewed the B&W request on its merits and believe it is justified.

The topical report also proposes extending indefinitely the current one hour time allowed for an inoperable channel to be in bypass prior to placing it in the tripped condition. This would allow repairs to proceed with the system in a two-out-of three configuration rather than a one of three which is more vulnerable to inadvertent scrams. The staff believes that an extension is warranted based on the risk and reliability analysis provided and that one hour is too short an interval in which to diagnose and repair instrument string components.

We believe that a reasonable time is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Contact:

D. Lasher (SICB/ DEST) x20787  ;

68 lof Ubl6 XA h) y

Edward L. Jordan Based on the CRGR charter, all staff approvals of topical reports should be reviewed by the CRGR. However, since the topical report and its supplement do not present any new methodology or require a staff position significantly different from that previously established for BWR ano Westinghouse PWR plants, we believe that CRGR review is not necessary. Since the BWOG request does go beyond previous generic requests, we would propose a briefing of the Committee  ;

or the CRGR staff. However, should you find that a CRGR review is necessary, '

please inform us and an appropriate CRGR package will be prepared. If CRGR review is not necessary, please advise us so the acceptance letter and SER can I be issued to the B&W Owners Group.  !

i 1

5iniezek, eput l -

ames .

i ec Office of Nuclear Reactor Regulation

Enclosures:

1. Draft Letter to B&W Owners Group

2. Safety Evaluation Report cc w/ enclosures:

T. Murley l

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/# " % ,4, UNITED STATES ENCLOSURE 1

[ h ., o i NUCLEAR REGULATORY COMMISSION

.' ,c WASHINGTON, D C. 20555

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Mr. C.W. Smythe Chairman BWOG/T.S. Comittee GPU Nuclear Company P.O. Box 480 Middletown, PA 17057

Subject:

NRC EVALUATION OF BWOG TOPICAL REPORT BAW 10167 AND S W LEMENT 1,

" JUSTIFICATION FOR INCREASING THE REACTOR TRIP SYSTEM ON-LINE TEST INTERVAL."

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The purpose of this letter is to provide the staff's evaluation of B&W Topical  !

Report 10167 and Supplement 1, prepared by Babcock and Wilcox for the B&W Owners Group (BWOG) Technical Specification Committee. This topical report was submitted by letter dated June 18, 1986 in response to the Technical Specification Improvement Program. Supplement I was transmitted on March 14, t 1988 in response to staff questions.

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This topical report presents justification in the form of risk analysis for extending the surveillance test intervals (STI) for RPS instrument strings from one month to six mcnths on a staggered test schedule. In addition, it proposed removing the current one hour allowable out of service time (A0T). In January 1988 the staff verbally presented questions regarding the report and its '

applicability to the generic resolution of Generic Letter 83-28, Item 4.5.3 to the author of the topical report, Mr. Enzinna. These questions were answereo and further information was provided in Supplement 1 to the topical report published in February 1988. Following the receipt of Supplement 1 on March 14, 1988, the staff met with BWOG representatives on June 21, 1988 to further discuss the review of the topical report and the proposals made in them.

The staff finds the topical report and its supplement acceptable for justifying the proposed extensions in STI and A0T for the RPS/ ARTS instrument strings with I one exception as discussed in the enclosed SER. The exception is the proposal to remove entirely the A0T for the instrument strings which would allow one of the four channels to be in bypass indefinitely. The staff agrees that one hour is too short a time interval in which to initiate any reasonable repair action and that your analysis justifies a significant extension. However, the staff continues to believe that a limit is required since the protection system has never been reviewed as a three channel system from a regulatory point of view.

Accordingly, the staff believes that an A0T of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable.

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Mr. C.W. Smythe The staff agrees that the STIs for the RPS instrument strings for those plants with a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or less A0T can be extended to the requested six month staggered test intervals contingent on the licensee in each case confirming that instrument drift occurring over the proposed STI would not cause the setpoint values to exceed those assumed in the safety analysis and specified in the Technical Specifications. Licensees must confirm that they have reviewed drift infonnation for each instrument channel involved and have determined that drift occurring in that channel over the period of the extended STI will not cause the setpoint value to exceed the allowable value as calculated for that channel by their setpoint methodology. (Instrument d*1ft is defined as the portion between the upper leave-alone zone and the allowable value.) Each licensee should maintain onsite records of the as-found and as-left values showing actual calculations and supporting data that are available for planned future staff audits. The onsite records should consist of monthly information taken over an extended period of time (approximately 2-3 years) and the plant-specific setpoint methodology used to derive the safety margins.

The staff further finds this topical report and its supplement generically acceptable for addressing the concerns of Generic Letter 83-28, Item 4.5.3.

Licensees may reference the analysis in this report and its supplement as part of the resolution of this item by confirming that the analysis applies to their plant.

Should you have any questions regarding the matters discussed above or the content of the enclosed SER, please contact Don Lasher of my staff on (301) 492-0787.

Sincerely, Gary M. Holahan, Acting Director Division of Reactor Projects III IV, Y and Special Projects x

Enclosures:

As stated l

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ENCLOSURE 2 SAFETY EVALUATION REPGRT E&WOG TOPICAL REPORT BAW-10167 JUSTIFICATION FOR INCREASING THE RTS ON-LINE TEST INTERVALS 1

1.0 INTRODUCTION

i The B&W Owners Group (BWOG) as part of their response to the Technical Specifi-  !

cation Improvement Program (TSIP) submitted Topical Report BAW-10167, "Justifi- l cation fcr Increasing the Reactor Trip System On-line Test Intervals" to justify increasing the Surveillance Testing Intervals (STI) from the current one month interval to a six month (with a staggered test schedule) interval. In addition, i the current allowable out-of-service time (A0T) limits for each instrument channel would be modified as discussed in the evaluation and conclusion of this 1

report. These changes apply only to the Reactor Protection System (RPS) and Anticipatory Reactor Trip System (ARTS) instrument strings which are defined as including all components downstream of the process sensors to and including the bistable. The BWOG analyses did not include manual reactor trip actions nor did they include the diverse trip system installed in response to 10CFR50.62  !

l (ATWS Rule). We were assisted in our review of the BWOG report by Idaho .

Naticnal Engineering Laboratory (INEL). The results of that review are reported in detail in EGC-RE0-7718 and summarized in this report.

This safety evaluation report evaluates the system modelling of the Reactor Trip System ano addresses both the acceptability of the probabilistic analyses and the acceptability of the extended STl/A0Ts. The acceptability of the probabilistic analyses was determined through review by our contractors, INEL, and by the Risk Application Branch. The acceptability of the extended STI/A0Ts is also based on the results of these reviews.

We further reviewed these analyses to determine their suitability for addressing the concerns of Item 4.5.3 of Generic Letter 83-28 (Salem ATWS). We found the

original analysis in BAW-10167 only partially addressed these concerns. Specific deficiencies were identified to the BWOG and they responded in Supplement I to BAW-10167 dated February, 1988. The results of our review of this adoitional information are presented in this report.

2.0 EVALUATION l

l This report includes input from the review of the tcpical report by our contractor, INEL; review of the INEL evaluation of the reliability data, probabilistic analyses and risk quantification results by the Risk Applications Branch; review and evaluation of the modelling of the two generic B&W reactor trip systems by the Instrumentation and Control Systems Branch; and review of the acceptability i of the risk analysis results as a basis for resolving the concerns of Generic Letter 83-28, Item 4.5.3.

In modelling the RTS for B&W plants, the analysis treated the differences in RTS configuration between plants by considering two models, one for the Oconee class of plants which includes Arkansas Nuclear One-1, Crystal River 3, Oconee 1, 2, 3, Rancho Seco and TMI-1; and the Davis Besse class which includes Davis Besse and, by extension, Bellefonte 1, 2 and WNP-1 whose reactor trip systems will be similar to that for Davis Besse. Within each of the two classes of plants, the individual plant to plant differences in instrumentation are found to be small particularly with regard to the components of the RPS/ ARTS instrument strings.

Since the differences in other equipment in the RTS are also small, the generic approach of modelling each of the two types of RTS is acceptable.

The most conservative success criteria for reactor trip used by the NRC was defined in SECY 83-293 as insertion of one-half the control rods into the core in a checkerboard pattern to shutdown the reactor. The BWOG requirement was more conservative in that it requireo full insertion of all the rods, both safety and regulating, to shutdown the reactor. This conservative approach could introduce uncertainties into the analysis results by potentially masking out contributions from RPS/ ARTS instrument string failures which could invalidate the BWOG conclusions. The sensitivity analysis to investigate the effect of

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1 this concern was performed during their review by the contractor which found that though trip breaker failures were the dominant effect, the effect of increased RPS/ ARTS instrument string STI/A0Ts en the ATWS induced risk continued to be small. Therefore, this concern appears to have little effect on the validity of the BWOG conclusions.

Reliability Block Diagrams (RBBs) were used by B&W to represent the haroware configurations for the modelling. Separate RBDs were drawn up for the Oconee and Davis Besse configurations to represent the RTS configuration differences l

between the Oconee class plants and the Davis Besse class plants. These RBDs l

model the hardware configuration. The RBDs are setup on a success criteria basis and model the RTS using removal of both main and secondary rod holding power as constituting successful operation of the RTS. These " super blocks" are then successively broken down into more detailed blocks in order to model the RTS in greater detail.

The level of detail included in the RBDs was stated to be dependent on the level of detail included in the reliability data available and included the major active components in the various RTS subsystems. The trip circuit breakers including their associated undervoltage and shunt trip attachments were modelled as part of the CRDCS subsystem. The RPS/ ARTS subsystem models I contained the sensors, instrument strings, main trip relays and output logic relays which was deemed sufficient modelling detail to show the insensitivity of RTS risk to changes in the STI/A0Ts for the instrument strings.

The Loss of Main Feedwater (LMFW) event was the transient modelled in the analysis. Other potential transients were considered by BWOG to be bounded by the LMFW event, however, through qualitative analyses perfomed by INEL and verified by the staff, it was determined that their results would be bounded by the LMFW results. Each RTS input parameter was included in the RBD models even though only those which would be the initiators of RTS action for the Loss of Main Feedwater event were quantified in the probabilistic analysis.

1 Several dependent failures involving the RPS main trip relays and output logic relays together with dependent failures of the trip' circuit breakers were not included in the BWOG models. The reason given for their exclusion was that the available failure data indicated a very small probability of failure associated with them. A better understanding of their importance with respect I to overall RTS availability, particularly in those situations where the instrument strings are bypassed, would be to include these dependent events in the models. Accordingly, these failures were included in fault tree models developed during our contractor's review to independently verify the BWOG results.

The RBDs modelling the RTS assumed that several diverse plant parameters would be perturbed to indicate the existence of an anticipated transient and the requirement for an automatic plant trip. For the LMFW event, two diverse RPS parameters, RCS high pressure and temperature, together.with either turbine trip or main feedwater pump trip were represented in failure space as a three input "AND" gate. For the dominant transient initiators reported in the Oconee PRA, a reasonable assumption is that the probabilistically significant initiators will cause an RPS/ ARTS instrument channel to initiate a reactor trip. However, accident scenarios such as loss of condenser vacuum or a loss of main feedwater caused by factors other than direct loss of MFW pumps can be' postulated which will not cause an ARTS parameter to be perturbed. For these situations it is reasonable to consider a case where only the two RPS parameters initiate reactor trip. This case was not included in the BWOG sensitivity studies but was investigated as part of the INEL verification review ano the results showed no significant change in core melt probability due to the increased STI for the instrument strings. This further supports the BWOG findings.

The BWOG RBDs do not contain explicit modelling for human error. The BWOG explicitly eliminated RTS recovery through manual trip by not including. manual reactor trip in their model because including recovery in the model would tend to mask the importance of the RPS/ ARTS instrument strings and hence reduce the sensitivity of RTS availability to increased STI/A0Ts. Human error associated with miscalibrating/ testing of the RPS/ ARTS instrument strings is implicitly incorporated into the f ailure probability estimated for the common cause events

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identified in the RBDs. The results of the review of the failure data indicated that the data included both human and hardware faults and that in many cases it was not possible to separate the human contribution from the haroware  ;

contribution to the failure rate. Therefore, in the BWOG analysis, the human i caused contributions were combined with the hardware contributions into a basic failure rate. In the INEL audit calculations, human-induced connon-cause failures of the instrument strings such as instrument miscalibration were included in the calculation but did not appear in the dominant cutsets because they were not significant. For the reactor trip modules (RTMs) and reactor trip j

breakers (RTBs), human caused failures that contribute to RTS unavailability derive mainly from failures of the RTB UV trip device which in the past required frequent readjustment caused by degeneration of the lubricant in the trip shaft bearings. The failure data does not separate failures caused by misadjustment from other failures, therefore it is not possible to develop a failure rate for human caused misadjustment of the RTBs.

The test and maintenance model was stated to be capable of evaluating various '

test intervals and maintenance outage durations for the entire RTS or for each subsystem of the RTS. However, in the original report it was only used to examine test and maintenance cutage durations for the RTS and ARTS instrument strings. In Supplement 1 the reactor trip modules and reactor trip breakers are examined. The instrument strings are tested in bypass, thus the modelling includes the contribution of the bypassed channel to reduced redundancy and hence reduced RTS availability for the duration of the test and any subsequent repair activity, if needed. In addition, the modelling includes the contribution to reliability derived from the detection of latent failures during testing which is traded off against the reduced redundancy. Also included was the effect of staggered testing on human-caused connon-mode failures. The effects of reduced redundancy on overall RTS unavailability was evaluated to examine the impact of testing the instrument strings under bypassed or tripped conditions.

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I e The fault tree logic models developed in the INEL review were evaluated to generate cutsets for these two conditions. The results show that in the tripped state the instrument string unavailability is insensitive to increased STI/A0Ts both with and without ARTS. In the bypassed state, the instrument strings show less i

than 2% increase in unavailability with ARTS and a larger increase without ARTS which indicates combinations of RPS instrument failures tend to increase in importance. The reactor trip breakers, the SCR electronic trip and the reactor trip modules are tested by tripping the appropriate components. Since the functions are not bypassed, there is no effect on RTS unavailability due to reduced redundancy during the testing but only a small increase in the frequency of spurious trips. The model includes monthly, staggered testing of these i components which contributes to RTS availability by detecting latent failures which may accumulate between tests.  !

1 1

Failure data used to estimate failure rates were taken from sources such as NUREG/CR 3289, IEEE Standard 500-1984, The Oconee PRA (NSAC-60) and NUREG/

CR 1740 which are all standard sources of data for independent (random) failure rates. Common cause failure rates were derived for instrumentation and control components from random failure rate data by multiplying the ran em rate by a beta factor which adjusts the failure rate to account for common cause faults.

The beta factors were obtained from NUREG/CR 3289. Breaker and UV device common cause failure rates were calculated directly from both B&W and CE operating experience data. CE operating experience data, reported in CE NPSD 277 was included because both B&W and CE reactors use GE Type AK-2-25 breakers as reactor trip breakers and in addition, the CE plants automatically initiate the shunt trip attachment upon demand for reactor trip. These sources also provided random failure data for the reactor trip breakers and UV trip devices.

The BWOG estimates of RTS unavailability were calculated for both models using B&W's PACRAT code which calculates reliability characteristics for a system model as a function of time. The BWOG estimates also include uncertainty analyses, however, these analyses assigned error factors of three for independent failures and ten for dependent (comon cause) failures. Data sources such as WASH 1400,

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I NUREG/CR 3289 and the Oconee FRA all indicate that the use of such small error factors was not realistic or conservative. Audit calculations to test the effect of BWOG's use of the small error factors and the validity of the BWOG 1 results were performed by INEL using the FRANTIC III code which produces time dependent estimates of RTS unavailability similiar to those produced by the PACRAT code.

The BWOG analysis estimates overall core damage frequency considering the changes in RTS unavailability, and spurious reactor trip rate. The results of this BWOG risk trade-off indicated no net change in either RTS unavailability or core damage frequency between the case of monthly testing as compared to the case of semi-annual staggered testing. However, a net decrease of 42% was found in the spurious trip core damage frequency when the STI was increased from monthly to '

semiannual. The exclusive use of hourly failure rates, the assumption of three i diverse parameters for RPS initiation regardless of transient initiator, and the use cf the LMFW transient as being representative and bounding for other initiators, were assumptions made by the BWOG in the course of their analysis that indicated sensitivity analyses were needed. As part of the INEL review, a set of 14 cases were analyzed to determine the effect of these assumptions. The base case (BWOG analysis) was compared to 13 variations in parameters. The results cf this sensitivity study indicated that for the Oconee class of j

plants, the reactor trip breakers were the dominant contributors to RTS 3

unavailability and that the instrument strings were insensitive to the proposed  !

increase in STI from one to six months and the removal of the one hour A0T time limit. For the Davis Besse class of plant, the BWOG results showed a much l lower RTS unavailability and the RTS availability was relatively insensitive to 4 changes in the instrument string STI/A0Ts. The INEL audit analysis showed the  ;

i Davis Besse model was not dominated by RTB failures but included comon cause i

failures of various sensors and random instrument string failures. These 3

results generally agreed with the BWOG results. '

i The BWOG proposed removing the limitations on A0Ts for the instrument strings l from the Technical Specifications entirely ano their analysis indicated no t significant increase in plant risk would result from doing so. The INEL j sensitivity study also showed no significant increase in plant risk would result {

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t from changes in A0T of IC8 and 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />. A value of 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> was used by BWOG as the mean time to repair (MTTR) for an instrument string in their model to represent the effect of an unlimited A0T. The staff believes that an extension to the allowable time a channel may be in bypass is clearly warranted, since one hour is too short a time for any reasonable repair action to be initiated and the effect of the bypass condition on risk is very small. A survey of technical specifications for the various B&W plants showed that two plants have A0Ts of one hour, one plant has a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> A0T, another has a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> A0T, and four plants have no A0T .specified in their Technical Specifica-tions. Therefore, allowing sufficient time for virtually all repairs to be accomplished without having to place the channels in trip is appropriate. On '

the other hand, the requested indefinite allowed cutage time would permit the RTS to operate in a two-out-of-three logic configuration which was 9t analyzed or reviewed as part of the plant's licensing basis. We believe this logic configuration would not meet IEEE Std 279 for all cases since the instrument strings may share penetrations, sensing lines and ultimate power sources. As a result, it is our judgement that an extension of the one hour A0T to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable based on the risk analysis, the IEEE 279 considerations and on l

previcus staff analyses that were performed for CE plants that conclude that 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is sufficient to make repairs.

3.0 EVALUATION OF GENERIC LETTER 83-28, ITEM 4.5.3 CONCERNS This BWOG analysis originally did not fully address all the five items and the remainder of the RTS as requested by item 4.5.3 of Generic Letter 83-28. Further discussion with BWOG and B&W indicated that the analysis had included the parts of the RTS other than the instrumentation strings ano the effects of the five items could be readily addressed. It was agreed that B&W would address the five items and report the results of the analysis of the remainder of the RTS in a supplement to BAW-10167. Supplement I was issued in February 1988 and our evaluation of the results reported in it follows.

Uncertainties in both component and comon cause failure rates were discussed earlier in the report as part of the discussion of the BWOG unavailability analyses.

As stated there, the uncertainties assigned to the independent and dependent

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failures were only factors of three and ten respectively which were considerably smaller than those recomended in standard sources of failure data. As part of the INEL review, audit calculations to test the validity of the BWOG results were performed.

Reduced redundancy during testing was addressed for the instrument strings, RTMs SCR trip, and RTBs as discussed in the evaluation of the test and maintenance model.

Operator errors during testing were not addressed separately but were combinea in the BWOG analyses with hardware faults into a basic failure rate. Thus the effect of these errors is included in the results of the analyses even though they were not separately identified ana calculated.

Component " wear out" caused by testing was addressed in Supplement 1 to BAW 10167 and extensive investigation of the breakers, which were assumed to be the '

RTS components most subject to wear-related failures was made. Other RTS components including relays and electronic components were also investigated.

Included in the investigation were " age-related" failures, primarily those involving progressive stiffening of the lubricant in the trip shaft bearings.

The effect cn RTS unavailability of the current STI and the proposed extensions of the STI for the RTMs and RTBs was more fully examined in Supplement 1.

Calculations of RTS unavailability for the RTMs and RTBs for STIs of one, four and six months were made with the STI for the instrument strings held constant at one month to investigate the sensitivity of the RTMs and RTBs to increased STIs. These calculations were made for both models and for the Davis Besse class, the results show a negligible increase in RTS unavailability for increases in STI to six months. For the Oconee class, the results indicated an approximate factor of four increase in RTS unavailability for an increase of STI from one to six months. The difference in results is due to RTS configura-tion differences between the Davis Besse and Oconee class plants. These results were said to be conservative because the beneficial effects of staggerea testing which will be done were not included in these BW0G calculations. INEL audit l

calculations to confirm these results were not performed, however the results appear reasonable and do show the current one month STI to provide the highest RTS availability.

4.0 CONCLUSION

S Our review found that, in general, the modelling of the RTS presented in the topical report, BAW-10167, was adequate to support the risk analyses performed to justify extension of the STI/A0Ts for the RPS/ ARTS instrument strings as requested in the report. During our review of this topical report, several concerns with the modelling of the RTS were identified as discussed in the Evaluation section of this report. Our conclusions on these concerns follow.

We concur that modelling of the RTS -for the Arkansas Nuclear One-1, Crystal River 3, Oconee 1, 2, 3, Rancho Seco and TH1-1 plants by the Oconee generic model is reasonable given these plants were all furnished with the same RTS configuration and this is acceptable. Similarly, since Davis Besse is the only k

current plant using its RTS configuration, we find this use acceptable. (

I (Bellefonte 1, 2 and WNP-1 may also use the Davis Besse RTS configuration, however, this concern can be resolved individually for these plants as part of their licensing review.)

It is also reasonable to assume that within each of the two RTS configurations, the plant-plant differences in instrumentation are small since the same configuration is used and the instrumentation components are all furnished by the same venaor (Bailey Peter). Therefore, we find this assumption acceptable.

A concern that the rather conservative success criteria used by BWOG in their modelling might introduce unacceptable uncertainties into the analysis results by masking out the RPS/ ARTS instrument strina failures was investigated through a sensitivity analysis performed curing our review. This indicated that though trip breaker failure dominates, the effect of increased RPS/ ARTS instrument string STI/A0Ts on the ATWS induced risk continued to be small. Therefore, we find this concern to be insignificant, t

We find the use of the RBDs as employed in the topical report to model the two RTS configurations acceptable. We also find that the invel of detail employed in the models is acceptable for showing the insensitivity of the ATWS induced risk to changes in the STI/A0Ts for the instrument strings and that the current test intervals for the RTMs and RTBs provide acceptable RTS availability.

We find the use of the LMFW as a representative transient to model the RTS for this analysis acceptable as also the assumption that the sensors and instrument strings useo in the analysis of the LMFW event were representative of sensors and instrument strings that would respond to other transients.

We noted in our evaluation that several depenoent failures involving RTS subsystems other than the RPS/ ARTS instrument strings were not included in the model. These were included in fault trees developed in the INEL review ano it i

was concluded that their effect was negligibly small.

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The BWOG RBDs did not model the effect on RTS availability of events that might l not perturb the ARTS sensors when only the two RPS parameters would be available I to trip the reactor. The INEL review found that even though a small increase inRTSunavailabilitywouldresult,itwasnothignificant. Therefore, we find that the BWOG neglect of this concern does not affect the applicability of the RTS unavailability results to the justification analyses.

l In our review of the RTS modelling we concluded that the treatment of human failures was reasonable and acceptable. The results of the INEL audit calculations showed little or no effect on RTS availability contributed by human-causeo errors from the instrument strings. For the RTMs and RTBs a qualitative evaluation shows that even though RTB failures are the dominant contributors to RTS availability, the major possibility for human caused failures is in adjustment of the UV trip devices. It was not possible to develop a human caused failure rate from failure data and the results of this evaluation indicated that it would have little significance. Therefore we conclude that human caused errors are not significant contributors to RTS l

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1 unavailability for the instrument strings. We also conclude that the treatment 1 1

of testing and maintenance concerns in the RTS modelling is conservative and, therefore, acceptable. l l

l We conclude that the failure data was obtained and developed from standard l l

sources using reasonable methods and is therefore acceptable.

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We conclude, on the basis of the INEL review and audit calculations, that the f

extended STIs proposed by BWOG in BAW-10167 for the instrument strings do not contribute an unacceptable increase in the net risk ano therefore, the increase in the test interval from one month to six months is acceptable. The STIs for  !

the RTMs and RTBs remain at monthly intervals and the tests are performed on a f staggered basis. Thus they will not contribute to the increase in overall risk.

I Even though the BWOG analyses and the INEL sensitivity study showed no

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appreciable increase in plant risk due to increasing the A0Ts, the staff  !

continues to believe that a limit on the allowable time a channel may be in bypass is required.

The staff also believes that one hour is too short a time for any reasonable repair action to be initiated. Accordingly, the staff finds 1

an A0T of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> acceptable. Plants that would volunteer for the extended test interval would need to confirm that the RTS channel A01 is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or less .

or moaify their specifications accordingly as part of their relaxation request.

This restriction is necessary since (1) the models presented in BAW-10167 depict e four channel system, (2) the protection system was originally reviewed and accepted as a four channel system, and (3) to meet the single failure require-ments of IEEE 279, the amount of time that a channel can be placed in bypass is limited to a period (e.g., less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) thct will offset the loss of conservatism incurred by placing the channel in bypass.

We also find that the effects of drift in both the sensors and the instrument strings were not considered sufficiently in this analysis, it is individual to each specific plant, and therefore should be assessed and factored into the analysis on a plant specific basis. Each licensee should, therefore, confirm that they have reviewed dritt information including as found and as left values for each instrument channel involved and have detemined that drif t occurring

l in that channel over the period of the extended STI will not cause the setpoint value to exceed the allowable values as calculated for that channel by their setpoint methodology (Instrument drift is defined as the portion between the upper leave-alone zone and the allowable value). Each licensee should maintain onsite records showing the actual setpoint calculations and supporting data that are available for planned future staff audits. This data should consist of montHy infonnation taken over an extended period of time (approximately 2-3 years).

The RWOG analysis in BAW-10167 crigininally did not fully address the Generic Letter 83-28, Item 4.5.3 concerns but additional supplementary information was provided to address these concerns.

We have evaluated this information and ccnclude as follows.

With regard to uncertaintities in component and coninon cause failure rates, we  ;

believed the BWOG assigned uncertainty factors were too small. The INEL audit

)

calculations showed however that the model used to calculate the point estimates of uncertainty was insensitive to instrument string failure rates and this is

! not a concern.

1 Our evaluation and the BWOG analysis showed that reduced redundancy from placing an instrument string in bypass during testing was not a significant concern for the B&W RTS configurations.

Operator errors during testing were not separately addressed in the BWOG analyses and no data was available to calculate them, therefore they were not included in the INEL audit calculations. Since the major contributors are trip circuit breaker failures, test caused failures would have had only a minor effect on the overall RTS unavailability.

The concern about component " wear-out" caused by testing was investigated in the BWOG analysis and we conclude from our review of this information that it is not a significant concern.

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t I

We conclude that the analyses in BAW-10167 together with additional information l in BAW 10167 Supplement 1 are acceptable to resolve item 4.5.3 for the B&W plants.

Licensees may reference these reports in their response to this concern and confirm the applicability of this information to their plant, j i

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i PRELIMINARY REGULATORY ANALYSIS l Storage of Spent Nuclear Fuel in NRC Approved Storage Casks at Nuclear Power Reactor Sites (10CFRParts72and170) i

1. STATEMENT OF THE PROBLEM I

It is anticipated that electrical utilities that utilize nuclear reactors i for power will have a major need for additional storage of spent fuel, to supplement the reactor's water basin storage, starting in the 1990s. The Nuclear Waste Policy Act of 1982 (NWPA) recognizes this need for additional spent fuel storage capacity at nuclear power reactor. sites. Insection218(a), q the NWPA states that "the Secretary [of DOE] shall establish a demonstration l program, in cooperation with the private sector, for the dry storage of spent fuel at civilian nuclear power reactor sites, with the objective of establishing one or more technologies that the Commission may, by rule, approve for use at the sites of civilian nuclear power reactors without, to the maximum extent practicable, the need for additional site-specific approvals by the Commission". In section 133, the NWPA states that "the Commission shall, by rule, establish procedures for the licensing of any technology approved by the Commissionundersection218(a)". The Commission recognizes these needs by including the development of the basis for rulemaking that would enable use of dry spent fuel storage in casks without, to the extent practicable, site-specific license reviews in their planning guidance (NUREG-0885, "U.S.

l Nuclear Regulatory Commission Policy and Planning Guidance 1987", Issue 6, September 1987). Currently the regulations in Part 72 do not permit licensing spent fuel storage without extensive site reviews. This rulemaking would accomplish these directives by providing for issuance of a general license to the holders of nuclear power reactor licenses for the storage of spent fuel at

, the site in dry casks approved by the NRC.

I 1 Enclosure 3

2. OBJECTIVES 2.1 To provide for compliance with the directives in sections 133 and 218(a) of the NWPA that instruct the Commission to approve one or more I technologies for the dry storage of spent fuel at civilian nuclear power reactor sites without the need for additional site-specific approvals and to set forth procedures for licensing any approved technology.

l l

2.2 To establish conditions that provide adequate protection of public health and safety and that are not inimical to the common defense and security.

l

3. ALTERNATIVES l

3.1 No Action The NWPA directs that the Commission approve one or more technologies, ,

1 that have been developed and demonstrated by DOE, for the use of spent fuel '

storage at the sites of civilian nuclear power reactors without, to the maximum extent practicable, the need for additional site-specific review. It also directs that the Commission, by rule, set forth procedures for licensing the l technology. Regulations for accomplishing these needs are not in place, thus, l

some action is necessary.

3.2 Available Alternatives The procedural alternatives available to NRC staff are amendment of licenses, use of regulatory guides or branch technical positions, and use of l

the rulemaking process. The purpose of this action is to license storage of l spent fuel. 10 CFR Part 72 specifically addresses dry storage of spent fuel, under a materials license. The reactor is licensed to operate under 10 CFR Part 50 and amendment of the reactor operating license, which is a facility

! license, is not appropriate. Regulatory guides or branch positions do not l carry the force of law, so they are only appropriate for conveying information concerning staff procedures. The preferred course of action is to proceed with rulemaking to amend Part 72.

2 Enclosure 3

4. CONSEQUENCES 4.1 Benefits l The proposed action will bring NRC regulations into compliance with the NWPA with no adverse effect on the public health and ' safety, and minimal impact on nuclear power reactor licensees, spent fuel storage cask system vendors, and the NRC. NRC and Industry impacts are discussed in sections 4.2.1, 4.2.3, and 4.2.4 l

The proposed rule would ensure protection of public health and safety through the use of the Commission's inspection and enforcement authority. NRC inspectors would inspect activities related to storage of spent fus1 at the reactor site and verify that conditions important to safety are in compliance with the Commission's regulations. Personnel from the Office of Nuclear Material Safety and Safeguards (NMSS) will evaluate design and fabrication procedures for storage casks, as submitted in a safety analysis report by cask vendors, approve cask system designs, and issue a Certificate of Compliance.

The criteria for obtaining a Certificate of Compliance are set forth in proposed subpart L. In general terms, approvals can only be obtained after NRC is assured that designs are adequate for storage of the type of spent fuel specified and that a quality assurance program (QA) acceptable to the NRC will be applied to the cask design, fabrication, testing, and maintenance. 'Except for the formal submittal of a license application and its related documents, the procedure for storage of spent fuel under this rule would essentially be the same as that currently required for a specific license under Part 72. The rulemaking process ensures that the public will be involved in the development I of any final rule that may be promulgated.

4.2 Impacts 4.2.1 NRC o NMSS. Approximately one staff-year (2087 hours0.0242 days <br />0.58 hours <br />0.00345 weeks <br />7.941035e-4 months <br />) of effort and  ;

$200,000 in contractor work is currently required for reviews and evaluations l l 3 Enclosure 3

i related to issuance of a specific license under Part 72. The $200,000 includes technical assistance for such work as independent verification of design criteria applications and computation of design bases.

It is estimated that additional spent fuel storage capacity, that is capacity I above that available in reactor spent fuel storage pools, will be needed at I

between 35 and 45 reactor sites by about the year 2000. This would average about 4 license applications per year, under existing regulations in part 72, over the 10 year period analyzed. Using this estimated average, the total resource burden on NMSS would be abou_t 83,480 staff-hours and $8,000,000, if specific licenses were required under part 72. The proposed rule would eliminate the need for site-specific license reviews, and thus, these resource requirements.

The staff did not analyze resource requirements beyond the year 2000, because recent indications are that the Department of Energy (DOE) will be accepting spent fuel at a repository by the year 2003. Power reactors of recent vintage have spent fuel pool designs that have adequate space for the l life of the plant, provided DOE is accepting spent fuel for disposal by about 2005. However, if DOE does not start accepting spent fuel in the 2005 time frame or if a significant number of reactor licensees seek and obtain license period extensions (beyond the current limit of 40 years), additional dry storage capacity could be needed at an additional 15 to 30 sites. Thus, the requirement for additional specific licenses and resource requirements would increase j proportionately. l The proposed rule would require that only NRC-approved casks be used for storage of spent fuel. Subpart L of the proposed rule sets forth criteria for obtaining cask approvals. The major burden for reviews, evaluations, and j issuance of Certificates of Compliance will be assigned to NMSS personnel. It is estimated that NMSS staff resource requirements for approval of each cask would be about one-half a staff-year (1,000 staff-hours). Based on current submittals and information, the staff anticipates that 10 or more applications for spent fuel storage cask system design approvals will be submitted. For purposes of this analysis it is assumed that there will be 10 submittals. Thus, it would require about 10,000 staff-hours for reviews and evaluations related to storage cask approvals, or about 1,000 staff-hours per year averaged over the 4 Enclosure 3 l

10 year period analyzed. Technical assistance contract costs of about $200,000 per cask design would still be required.

Cask designs that have been approved for transportation of spent fuel may also be considered for spent fuel storage. If the cask vendor has a Certificate of Compliance issued under Part 71 for the cask, the procedure for approval for spent fuel storage would entail an analysis showing that the spent fuel could be stored in the cask safely for 20 years. This could be a rather simple analysis and no technical assistance costs are anticipated. Further, since many of the cask designs currently approved for transportation would be l uneconomical for spent fuel storage, few submittals are expected. Thus, no l separate analysis for this type of approval is made.

t o NRR. The anticipated involvement of the Office of Nuclear Reactor Regulation (NRR) personnel under the proposed rule would be significantly different from their current involvement. Currently, NMSS personnel coordinate with reactor project managers to ensure that independent spent fuel storage installation (ISFSI) operations and reactor operations are compatible. Under the proposed rule, it is anticipated that NRR personnel will be responsible for inspections related to initial spent fuel cask fabrication. The regions, under NRR programs, will conduct physical protection inspections.

It is estimated that about 7,000 metric tons of heavy metal (MTHM), as spent fuel, exceeding spent fuel storage pool capacity will be removed from reactors over the 10 years analyzed. It is assumed that this spent fuel will be stored in casks and that about 10 MTHM can be stored in a cask, thus, a total of about 700 storage casks will be required. It has been estimated that the maximum cumulative aaditional requirements (beyond the capacity of reactor spent fuel pools) for spent fuel storage will be between 12,000 and 20,000 MTHM and will occur between the years 2012 and 2016. Thus, estimates made in this analysis are probably conservative and the number of casks required could be significantly higher. Experience indicates that extrapolations into the future are subject to large uncertainties, so reevaluations will be made during the continuing rulemaking procedure.

5 Enclosure 3 l

i i

NRR will conduct inspection of cask manufacturers during their initial cask production. Subsequent NRR inspections will be conducted on a reactive basis in response to identified areas of concern. It is estimated that the total number of initial inspections will be ten. An inspection trip is estimated to require about 40 staff-hours and cost about $2,000. The reason for the large estimated expense is that some inspection trips will require foreign travel. In addition, there would be technical assistance contract costs estimated to be about $10,000 per inspection. It is anticipated that there will also be about 80 staff-hours required for inspection preparation and report writing. Thus, about 1,200 staff-hours and $120,000 in expenses would be required for an initial inspection of the ten cask fabricators. If three initial inspections and one or two reactive inspections per year are conducted, a resource burden of about 0.3 FTE would result. r I

l o Total NRC Resource Requirements. If storage of spent fuel in an I ISFSI were to be licensed under existing regulations, the major resource requirements would be the estimated 83,480 NMSS staff-hours and the $8,000,000 in contracted technical assistance required for reviews and evaluations related to issuing an estimated 40 specific licenses which would be required under i existing Part 72. This would be about 8,350 staff-hours and $800,000 per year, averaged over the 10 year period analyzed. Other NRC resource requirements would not have been zero, but would be relatively small.

Resources requirements for NMSS activities under the proposed rule are estimated to be about 1,000 staff-hours per year for storage cask system design reviews and approvals. In addition, there would be the 1,200 staff-hours and

$120,000 expenses per year required by NRR related to initial cask fabrication inspections, and the 800 staff-hours per year required by the regions for onsite inspections. Thus, NRC staff resource requirements are estimated to l amount to about 3,000 staff-hours per year under the proposed rule, which would compare to an estimated 8,350 staff-hours if specific licenses were to be approved under existing regulations. The $200,000 for contractor assistance J would continue to be part of the initial approval of cask designs and would amount to $2,000,000 over the 10 year period analyzed. These contractor assistance costs would be about a quarter of those estimated to be required if specific licenses were to be required. Resources also are required for 6 Enclosure 3

l inspections related to safeguards and physical security, but these resources are expected to remain about the same as those currently required and are not i considered separately'. The total estimated resources for training are 1,200 staff-hours and $15,000 in expenses. This leads to the conclusion that total NRC resource requirements under the proposed rule would be lower than if 3 l

specific licenses were issued. In any' case, fee schedules in 10 CFR 170 are 1

being revised to ensure that costs related to the rule are fully recovered.  !

4.2.2 Other Government Agencies.

No other Government agency, except TVA, is licensed to operate a nuclear power reactor. The impacts estimated for nuclear power reactor 1,icensees would apply to TVA.

4.2.3 Nuclear Power Reactor Licensees Currently nuclear power reactor licensees must submit an application for a specific license under 10 CFR Part 72 to store spent fuel in an independent spent fuel storage installation on the reactor site. Licensing fees in cases similar to those covered under the general license amount to between $250,000 j to $300,000. These licensing fees would be eliminated under this proposed l rule.

It is estimated that the annual reporting burden for a specific license under Part 72 is currently about 1,309 hours0.00358 days <br />0.0858 hours <br />5.109127e-4 weeks <br />1.175745e-4 months <br /> and the recordkeeping burden about 5,165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br />. The proposed rule would eliminate the annual reporting burden for affected reactor licensees. Costs related to printing and distribution of license documents required by Part 72 (e.g., safety analysis report,, '

environmental report) would also be eliminated. However, the recordkeeping burden would remain about the same. Records related to operating and organizational activities would still have to maintained by the reactor licensee and would be subject to inspection by, but need not be submitted to, the NRC. The proposed rule would not alter reactor operating requirements under Part 50. The proposed rule would simplify the procedures under which a nuclear power reactor licensee could store spent fuel. A draft regulatory 7 Enclosure 3

. guide entitled, " Standard Format and Content for a Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks," was issued for public comment in April 1986 under task number CE-301-4. This guide and the public comments received on it were considered in the development of proposed subpart K.

(Single copies of the draft guide may be obtained from W.R. Pearson, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC, 20555 (Telephone: (301) 492-3764).

Incremental costs for design, construction, operation, and decommissioning independent spent fuel storage installations under this proposed rule have not been separately estimated. If a reactor licensee has a need for storage of spent fuel beyond the capacity of the reactor storage pool, the licensee could choose between whether to apply for a specific license under 10 CFR ,

Part 72 or to store spent fuel under the general license provided by this rule. In either case there would be costs related to the design, construction, operation, and decommissioning. Licensees would decide on the procedure that provided the better solution for their purposes.

4.2.4 Cask Vendors.

Cask vendors have submitted six topical safety analysis reports to NMSS for approval for use of casks for spent fuel storage. Four of these topical reports have been approved and these cask designs are being approved in this rulemaking.

Costs for approval of a topical safety analysis report, which is the present means of getting dry spent fuel storage cask designs approved for use, are currently limited to $20,000. This is significantly less than NRC actual costs. Under the proposed rule NRC would recover full costs for approval of cask designs. This could amount to about $250,000 to $300,000 per design, including cost of review and approval procedures, contractor work, and cask fabrication inspections.

No incremental costs related to reporting requirements are expected as a result of this action. The criteria for approval of spent fuel storage casks, as set forth in subpart L, are not significantly different from the design, 8 Enclosure 3

, fabrication, and quality assurance criteria that are currently used. A draft regulatory guide entitled, " Standard Format and Content for a Topical Safety Analysis Report for a Dry Spent Fuel Storage Cask," was issued for public comment in April 1986 under task number CE-306-4. The guide and the public comments received on it were considered in development of subpart L. (Single copies of the draft guide may be obtained from W.R. Pearson, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC, 20555 (Telephone: (301) 492-3764)).

4.2.5 Public.

No incremental cost impact on the public is expected. As shown in the preceding cost analyses, no significant increase in the cost of doing business is expected as a result of this action. It is anticipated that c'osts to NRC l and power reactor licensees will be less than that required to obtain a specific license for the same type of storage. However, these incremental cost I reductions would be insignificant compared to the overall costs. Costs to spent fuel storage cask vendors is expected to increase, but since only about  !

10 submittals for cask design approvals are expected. So the total economic l impact on the public as a result of this action is not expected to be significant. Since the power reactor licensee must comply with the requirements of the Commission's regulations, no reduction in public health and safety is anticipated. In fact, risk to public health and safety could be reduced, because if shipments of the spent fuel are significantly delayed, the

l. radioactivity of the spent fuel would be lower at the time of shipment.

4.3 Impacts on other Requirements.

4.3.1 Other Rulemakings A final rule amending Part 72 was made effective on Monday September 19, 1988 (53 FR 31651). It primarily concerned licensing the storage of spent fuel and high-level radioactive waste in a monitored retrievable storage facility, which would be constructed and operated by the DOE and licensed by the Commission. It did not address requirements for licensing mandated by sections 133 and 218(a) of the NWPA.

9 Enclosure 3

4.3.2 Finding of no Significant Environmental Impact The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. The rule is mainly administrative in nature and would not change safety requirements, which could have significant environmental impacts. The proposed rule would provide for power reactor licensees to store spent fuel in casks approved by the NRC at reactor sites without additional site-specific approvals by the Conmission. It would set forth conditions of a general license for the storage of spent fuel and procedures and criteria for obtaining storage cask approval. , The

environmental assessment and finding of no significant impact on which this l determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW., Washington, DC (Lower Level). Single Copies of the environmental assessment and finding of no significant environmental impact are available from W. R. Pearson, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC 20555; Telephone

(301) 492-3764.

l 4.3.3 Paper Work Reduction Statement l

This proposed rule amends information collection requirements that are subject to the Paper Work Reduction Act of 1980 (44 U.S.C. 3501 et seq.).

l This rule has been submitted to the Office of Management and Budget for review and approval of the paper work requirements. l 4.3.4 Regulatory Flexibility Act Certification In accordance with the Regulatory Flexibility Act of 1980 (SU.S.C 605(b)), the Commission certifies that this rule, if promulgated, will not have l

a significant economic impact on a substantial number of small entities. This rule would affect only licensees owning and operating nuclear power reactors.

The owners of nuclear power reactors do not fall within the scope of the l definition of "small entities" in the Regulatory Flexibility Act or the small l 10 Enclosure 3 1

1

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Business size standards set forth in regulations issued by the Small Business Administration at 13 CFR Part 121.

4.3.5 Backfit Analysis The NRC has determined that a backfit analysis is not required, because these amendments do not involve any provisions that would impose backfits as defined in 650.109(c}(11 I

4.4 Constraints.

No legal, institutional, or policy constraints are anticipated.

5. DECISION RATIONALE. '

An assessment of the benefits and impacts of the alternatives leads to the conclusion that the requirements of the proposed rule are commensurate with the Commission's responsibilities for public health and safety and the common defense and security. No other available alternative is believed to be as satisfactory, thus, this action is recommended.

6. IMPLEMENTATION.

This proposed rule will be published in the Federal Register allowing 45 j days for public comment. Since rulemaking is mandated by the NWPA and the I incremental impacts of this rule are minor, no implementation problems are anticipated.

I 11 Enclosure 3

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NUCLEAR REGULATORY COMMISSION 10 CFR Parts 72 and 170 Storage of Spent Nuclear Fuel in NRC-Approved Storage Casks at Nuclear Power Reactor Sites AGENCY: Nuclear Regulatory Commission.

l ACTION: Proposed rule.

SUMMARY

The Commission is proposing to amend its regulations in 10 CFR Part 72 to provide, as directed by the Nuclear Waste Policy Act of 1982, for the storage of spent fuel at the sites of power reactors without, to l the maximum extent practicable, the need for additional site-specific l

approvals. Holders of power reactor operating licenses would be permitted to store spent fuel, in casks approved by NRC, under a general license.

The proposed rule contains criteria for obtaining an NRC Certificate of Compliance for spent fuel storage casks.

DATE: Submit comments by (45 days following publication). Comments l received after this date will be considered if it is practical to do so, but the Commission is able to assure consideration only for comments received on or before this date.

ADDRESSES: Mail written comments to Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, 20555 ATTN: Docketing Service Branch.

Deliver comments to One White Flint North, 11555 Rockville Pike, Rockville, MD between 7:30 a.m. and 4:15 p.m. weekdays.

11/08/88 1 Enclosure 1 l .

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, Copies of NUREG-0459, 0575, 0709, 1092, 1140, and NUREG/CR-1223, reports which are referenced in this notice and the environmental assess-ment, may be purchased through the U.S. Government Printing Office by calling (202) 275-2060 or by writing to the U.S. Government Printing Office, P.O.' Box 37082, Washington, DC 20013-7082. Copies of DOE /RL-87-11, refer-enced in the environmental assessment, and the NUREG reports listed above may be purchased from the National Technical Information Service, U.S.

Department of Commerce, Springfield, Virginia 22161. Copies of the NUREG reports listed above, the environmental assessment and finding of no sign-ificant environmental impact, and comments received on the proposed rule are available for inspection and copying for a fee at the NRC Public Document Room, 2120 L Street NW., Washington, DC, Lower Level.

FOR FURTHER INFORMATION CONTACT: William R. Pearson, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC 20555.

Telephone: (301)492-3764.

SUPPLEMENTARY INFORMATION:

Background

Section 218(a) of the Nuclear Waste Policy Act of 1982 (NWPA) includes the following directive, "The Secretary [of DOE] shall estab-lish a demonstration program in cooperation with the private sector, for the dry storage of spent nuclear fuel at civilian nuclear power reactor sites, with the objective of establishing one or more technologies that the [ Nuclear Regulatory] Commission may, by rule, approve for use at the sites of civilian nuclear power reactors without, to the maximum extent 11/08/88 2 Enclosure 1

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b. practicable, the need for additional site-specific approvals by the Commission." Section 133 of the NWPA states, in part, that "the Commission shall, by rule, establish procedures for the licensing of any technology approved by the Commission under Section 218(a) for use at the site of any civilian nuclear power reactor."

l Discussion l This proposed rule would allow power reactor licensees to store spent fuel at the reactor site without additional site-specific re, views. A gen-eral license would be issued to holders of power reactor licenses for the storage of spent fuel in dry. casks approved by the NRC. The reactor licensee ,

l would have to show that there are no changes required in the facility tech-nical specifications or unreviewed safety questions related to activities involving storage of spent fuel under the general license. The licensee 1

would also have to show conformance with conditions of the Certificate of Complian:e issued for the cask by NRC. The licensee would have to estab-lish and maintain records showing compliance, which would have to be made available for inspection by the Commission.

This rule would not limit storage of spent fuel to that which is generated at the reactor site. Transfers of spent fuel from one reactor i

site to another are authorized under the receiving site's facil,ity operat-ing license pursuant to 10 CFR Part 50. The holder of a reactor operating license would apply for a license amendment, under S 50.90 (unless already authorized in the operating license), for the receipt and handling of the spent fuel from another reactor. In addition, the reactor licensee would be expected to request amendment of the Price-Anderson indemnification 11/08/88 3 Enclosure 1

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agreement to provide for coverage of the transferred spent fuel. 10 CFR i

Part 72 is not germane to such transfers of spent fuel. If the spent i

fuel has been previously transferred and is currently stored in the reactor spent fuel pool, the only consideration under the general license would be whether or not the spent fuel meets conditions of the cask's Certificate of Compliance.

Although experience with storage of spent fuel under water is greater than with dry storage in casks, experience with storage of spent I

fuel in dry casks is extensive and widespread. The Canadians have been storing dry CANDU-type spent fuel at Whiteshell in vertical concrete casks called silos since 1975. Although the storage of spent fuel at Whiteshell does not involve light-water-reactor (LWR) fuel, it has con-tributed to the knowledge and experience of dry spent fuel storage in l

l concrete casks. Dry cask storage has been demonstrated in West Germany.

There has also been experience with dry spent fuel storage in the United States. The Department of Energy (D0E) and its predecessors have kept non-LWR spent fuel in dry storage in vaults and dry wells since the 1960s.

An NRC survey of the dry storage of spent fuel, in the United States and elsewhere, was presented in NUREG/CR-1223, " Dry Storage of Spent Fuel -

A Preliminary Survey of Existing Technology and Experience" (April 1980).

NUREG/CR-1223, at Section IV.C, contains a description of DOE demonstra-tion of dry LWR spent fuel storage in sealed storage casks (SSC) and dry wells. The storage of LWR spent fuel in SSC, which is an above ground, steel-lined, reinforced concrete cylinder or cask, started in 1979. The DOE demonstration program has continued and has been expanded to include dry storage in metal casks and storage of consolidated fuel rods as well as storage of spent fuel assemblies. Programs have been conducted by DOE i

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in cooperation with Virginia Power at its Surry plant, with Carolina Power and Light at its H.B. Robinson 2 plant, and with General Electric at its Morris plant for dry storage of LWR spent fuel. Also dry storage of LWR l

spent fuel assemblies continues at the Idaho National Engineering Lab-oratory, along with demonstration of their disassembly and storage of the consolidated fuel rods.

The NRC staff has obtained substantive information from the DOE development programs. It has also gained experience from the issuance of licenses for the onsite storage of spent fuel in nodular cast iron casks at the Surry site of Virginia Power and in stainless steel canisters stored inside concrete modules at the H.B. Robinson 2 site of Carolina Power and Light. The safety of dry storage of spent fuel was considered during l

l development of the Commission's original regulations in 10 CFR Part 72,

" Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Iiistallation (ISFSI)," which was promulgated on )

November 12, 1980 (45 FR 74693). A final rule entitled, " Licensing l

Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste," which replaced the regulations issued on 11/12/80, was publi:hed in the Federal Register on August 19, 1988 (53 FR 31651) and became effective September 19, 1988. This final rule mainly provides for licensing the storage of spent fuel and high-level waste at a monitored retrievable storage (MRS) facility, and does not cover the mandates of Sections 133 and 218(a) of the NWPA. However, it did specifically address the safety of dry storage of spent fuel.

11/08/88 5 Enclosure 1

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Activities related to loading and unloading spent fuel casks are routine procedures at power reactors. The procedures for dry storage of spent fuel in casks would be an extension of these procedures. Over the last several years the staff has reviewed and approved four spent fuel storage cask designs. Requests for approval of cask designs are currently submitted in the form of topical safety analysis reports (TSARS). Four dry storage cask TSARS have been approved for referencing, which means that an ISFSI license applicant may reference appropriate parts of the report in licensing proceedings for the storage of spent fuel. This greatly reduces an ISFSI license applicant's time, effort, and cost. The same reliance on an approved safety analysis is being made for on-site dry cask storage.

Separate topical safety analysis reports have been received for design of casks fabricated using nodular cast iron, thick-walled ferritic steel, concrete, and stainless steel and lead. Four cask design topical reports are under active review at the present time. Four spent fuel storage cask topical safety analysis reports have been approved for referencing, and approval of their design for spent fuel storage under a general license is being included in this rulemaking. Casks approved for storage in the future will be routinely added to the listing in proposed 672.214 through rulemaking procedures. Their Certificates of Compliance will be exhibited in NUREG report issued by the NMSS staff, which will be updated as appropriate. Because this type of rulemaking would neither constitute a significant question of policy nor amend 10 CFR Parts 0, 2, 7, 8, 9 Subpart C, or 110, the Commission concludes that additions to 672.214 may be made under the rulemaking authority delegated to the Executive Director for Operations.

11/08/88 6 Enclosure 1

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During review of storage cask designs to be approved in this rule- q making, the NMSS staff has considered the compatibility of their designs with transportation to and disposal at DOE facilities and will continue to do so in the future. Currently, there is limited knowledge concerning specific design criteria to facilitate handling of spent fuel between the time it is put into casks at the reactor site and the time it will be handled for interim storage at a monitored retrievable storage facility )

4 (MRS) or disposal at a repository. However, the staff will remain in

{

contact with DOE and will ensure, to the extent practicable, that cask J designs incorporate the latest design criteria available at the time that j t

the design is approved.

The NRC experience in the review of cask design and fabrication and licensing of spent fuel storage installations on the site of operating l reactors has been documented in part by publication of two draft regula-tory guides. In April of 1986, two draft regulatory guides entitled

" Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks" (Task number CE-301) and " Standard I

Format and Content for a Topical Safety Analysis Report for a Dry Spent Fuel Storage Cask" (Task number CE-306) were issued for public comment.

Single copies of these draft guides may be obtained from W.R. Pearson,  !

Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC 20555 (Telephone: (301) 492-3764).

The passive nature of dry storage of spent fuel in casks provides operational benefits attractive to potential users. One benefit is that there is no need to provide operating systems to purify and circulate cooling water or other fluid. Another benefit is that the potential for corrosion of the fuel cladding and reaction with the fuel is reduced, 11/08/88 7 Enclosure 1

I

[7590-01]

especially if an inert atmosphere is maintained inside the cask. Also, because cooling of the spent fuel is a passive activity, active mechanisms, such as pumps and fans, are not required. Although Part 72 allows storage of any spent fuel over one year old, (i.e., one year since the fuel was involved in a sustained nuclear chain reaction), it is anticipated that most spent fuel stored in casks will be five years old or more. Because )

of the passive nature of cask cooling, the storage capacity of a cask is significantly increased as the spent fuel is aged, especially for fuel ll that is five years old or more. It is probable that reactor licensees l

1 will remove the older fuel from their storage pools to take advantage of (

, j this additional cask storage capacity. Spent fuel storage casks are 1

massive (on the order of 100 tons), of simple design, passive in nature, j and will be manufactured under a strict quality assurance program.

The Commission believes that, with provisions for proper quality assurance ensured under the Commission's inspection and enforcement j authority, dry storage of spent fuel in casks provides adequate protec-tion to public health and safety and the environment.

I 1

Proposed Rule The General License Under this proposed rule, a general license would be issued to holders of nuclear power reactor licenses to store spent fuel at reactor sites in casks approved by the NRC. The Commission will rely on dry storage of spent fuel in casks for confinement of radioactive material to provide adequate protection of public health and safety and the environment. It will rely on its inspection and enforcement authority to ensure compli-ance with conditions of the general license and cask certificates. A 11/08/88 8 Enclosure 1

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[7590-01]

power reactor license holder would have to notify the Commission before storing spent fuel under the general license for the first time and reg-ister use of each cask as the spent fuel is stored. The Commission would make use of the notification of first use to initiate on-site inspection actions. The registration of each cask will be used to establish indepen-dent records related to use of casks. These records will be used to detect potential cask safety problems, to keep track of cask repairs, to keep track of defects and damage that may result in a significant reduc-tion in safety, and to keep track of the date by which spent fuel must be removed from the cask unless the cask model has been reapproved. (As explained later, a cask service life is initially limited to 20 years.)

A separate record would also be established for each cask by the cask vendor. This record would be transferred to, and be maintained by, cask users who would show any maintenance to the cask and lists its contents.

When a cask model has been reapproved, users of this model would be noti-l fied by the NRC. They would also be advised of any changes in conditions for use of the cask.

The reactor license holder would have to show that the storage of spent fuel will be in compliance with the conditions of the cask Certifi-cate of Compliance, including assurance that site parameters and other design bases are within the envelope of the values analyzed in the cask safety analysis report. An evaluation would also be made to show that there will be no changes necessary to the facility technical specifica-tions and no unresolved safety questions in activities involving the storage casks. Procedures and criteria in 10 CFR 50.59 would be used for these evaluations. These types of evaluation are currently done for specific licenses issued under Part 72. Issues related to systems and 11/08/88 9 Enclosure 1

[7590-01]

l

~

components used both for reactor operations and spent fuel storage j l

activities would be included. Most concerns to date have been related to {

I control of heavy loads and have been accommodated. If there is a safety )

problem or a change in technical specifications required, and the reactor l license holder wishes to store spent fuel under the general license, the problem must be resolved before storage, including submittal of an application for license amendment under Part 50 if necessary.

The reactor license holder would commit to establishing and maintain-ing a quality assurance program, an emergency plan, a training program, and a radiation control program for activities related to spent fuel storage under the general license. Similar plans and programs are in effect for reactor operations. The appropriate existing plans or pro-grams could be modified or amended to cover activities related to the spent fuel storage, if the reactor licensee chose to do this. These d

plans and programs would be raade available for examination by the NRC l inspection staff.

The reactor license holder should make a commitment to conduct spent l 1

fuel storage activities in accordance with written and approved procedures.  !

Procedures for safe handling of the spent fuel should be established by a thorough study of what is to be accomplished and approved by two inde-pendent competent groups within the licensee's organization. The reactor licensee has made this commitment for reactor operations. The same or a similar approval system may be used for this storage of spent fuel.

Instances in which significant reductions in the safety effective-ness of or defects in casks are discovered must be reported. Initial notifications would have to be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and written reports would follow within 30 days. The 24-hour notification is necessary, 11/08/88 10 Enclosure 1

[7590-01]

because defects or damage from accidents may present a hazard to public health and safety which could be mitigated by assistance from Federal l Agencies, including the Commission. A written report is needed so that the staff can determine whether or not there are generic health and safety implications. The 30-day period is allowed so that the licensee can review and analyze the event and present a clear and complete history.

When the power reactor operating license expiration date approaches, l the holder of the license must take some actions. Under 10 CFR 50.54(bb) the reactor license holder must submit.a program in writing to the Com-mission, no later than five years prior to the license expiration date, showing how the reactor licensee intends to manage and provide funding for the management of all irradiated fuel on the reactor site. This i program would have to include the spent fuel stored under the general license proposed in this rulemaking. The reactor licensee will also have to decide whether to request termination of the reactor oper-l ating license under 10 CFR 50.82. If the reactor license holder decides to apply for termination of the license, the plan submitted with the application must show how the spent fuel stored under this general license l will be removed from the site. The plan would include an explanation of when and how the spent fuel will be moved, unloaded, and shipped prior to  !

starting decommissioning of the equipment needed for these activities.

In part, the environmental assessment for this rulemaking relies on findings from the waste conficcace proceedings, in which the Commission concluded they had confidence that there would be no significant environ-mental impacts from the storage of spent fuel for a period of 30 years beyond the expiration date of reactor licenses. Thus, an application for reactor license termination that proposes a decommissioning period beyond 11/08/88 11 Enclosure 1

I

[7590-01]

this 30 year period would have to contain a discussion of the environ-mental impacts from storage of the spent fuel beyond the period analyzed by the Commission. The general license would terminate automatically when the spent fuel is removed from storage.

Cask Certification Cask vendors will submit a safety analysis report (SAR) showing how cask designs and fabrication can provide adequate protection to public health and safety. In the process of evaluating design bases in the SAR, I certain assumptions must be made in order to arrive at practical solu-tions. One assumption is that the spent fuel will be stored in the cask for 20 years. Thus, the NRC initially approves casks for only 20 years of storage, after which they would have to be reapproved. This does not mean that after 20 years the cask becomes an unsafe container, it simply means that evaluations were not performed for a period greater than 20 years. The service life of a cask is 20 years from the time spent fuel is initially loaded into the cask. As a result of the limited i

service life, casks in use will have varying storage lifetimes remaining. l For instance, 20 years after a cask model has been approved for storage there could be casks of this model in use with from zero to 20 years of service life remaining.

The holder of the cask Certificate of Compliance (cask vendor) should apply for reapproval of a storage cask. Submittal of an application would be made 17 years after the initial cask approval date, which is three years prior to the expiration date of the cask certificate, to allow time for the NRC staff to reevaluate the cask safety and reissue the cask certificate. However, if the holder of a cask certificate goes out of 11/08/88 12 Enclosure 1

i l [7590-01]

business or will not submit an application for reapproval in a timely manner for any reason, the Commission would be notified and in turn would notify the cask users. In any case, cask users would have to take action to ensure that spent fuel is stored in casks approved by the NRC. Several options would be available to licensees. If the cask were reapproved under submittals by the vendor, the Commission would notify all users and the only action necessary for the users would be to update the cask records, i If the cask vendor does not apply for reapproval, for whatever reason, the licensee would be notified by the Commission. The licensee would then have to arrange for reapproval or remove casks from service as their i service life expires. This could mean removal of the spent fuel and storing it elsewhere.

The cask will be relied on to provide safe confinement of radio-active material independent of the operations in which it is involved or .

l regardless of its location, so long as conditions comply with the Certi-1 ficate of Compliance. The cask approval program will be analogous to l that now conducted for casks approved for shipping spent fuel under 10 CFR Part 71. Records will be established by vendors and maintained by users to provide historical information on all casks, so that if there is a safety problem with a particular cask, a cask fabrication process, or with a cask model the NRC could issue notices to cask vendors I and users to initiate corrective actions. '

The Commission believes that a prudent concern for overall' activities related to the back-end of the LWR fuel cycle dictates that consideration should be given to the compatibility of spent fuel storage cask designs with the transportation of the spent fuel to its ultimate disposition at a Department of Energy (00E) monitored retrievable storage facility or 11/08/88 13 Enclosure 1

1

[7590-01] l l

geologic repository. Cask designers should be aware of DOE developments and plans for transportation of spent fuel offsite and should design spent fuel storage casks, to the extent that is practicable given the l l

information that is available at the time that the cask is designed, for j compatibility with future disposition of the spent fuel. 1 The four cask designs that would be approved in this rulemaking i

comply to the extent practicable at this time. The Commission notes {

that the vendors of these casks have indicated their intent to pursue certification for these casks as shipping casks for offsite transporta-tion under 10 CFR Part 71. However, spent fuel can be safely off-loaded from storage casks at reactor sites, if necessary, at the end of the stor-age period. In the interest of overall fuel cycle efficiency, however, I

the Commission encourages storage design developments to avoid this j 1

eventuality for all spent fuel storage.

The scope of this rule is to allow holders of nuclear power reactor licenses to store spent nuclear fuel at reactor sites under a general l

license using certified dry storage casks, because use of these casks is essentially independent of site characteristics. The Commission has evaluated and approved, in specific licenses issued under 10 CFR Part 72, other types of dry storage modules. These methods may be approved in the future for use under a general license.

NRC costs related to cask approval reviews and evaluations, quality assurance program approvals, and cask fabrication inspections would be fully recovered. The schedule of fees in 10 CFR 170.31 and 170.32 would be revised to recover these costs. Inspection of plant and site-related activities would be performed by resident inspectors. NRC costs related 11/08/88 14 Enclosure 1

[7590-01]

to onsite inspections would also be recovered under 10 CFR Part 170.

Safeguards Spent fuel removed from light water reactors contains low enriched uranium, fission products, plutonium, and other transuranium elements j (transuranic). Owing to the special nuclear material in spent fuel, safeguards for an independent spent fuel storage installation must pro-tect against theft and radiological sabotage and must provide for mate-rial accountability. The requirements for physical protection are set forth in proposed S 72.212. No specific requirements for material con-trol and accounting are being added because existing requirements in Parts 72 and 50 are adequate. <

i The theft issue arises mainly from the plutonium component of the spent fuel. Plutonium, when separated from other substances, can be used 1

in the construction of nuclear explosive devices and therefore must be provided with a high level of physical protection. However, the plutonium contained in spent fuel is not readily separable from the highly radioactive fission products and other transuranic and for that reason is not considered a highly attractive material for theft. Moreover, the massive construction of casks significantly complicates theft scenarios.

For these reasons no specific safeguards measures to protect against theft are proposed other than maintaining accounting records and conducting periodic inventories of the special nuclear material contained in the spent fuel. .

Safeguards measures to protect against sabotage should be consistent with the probability and consequences of radiological sabotage. The term

" radiological sabotage" is defined in 10 CFR Part 73 and means any delib-erate act directed against a plant or transport vehicle and cask in which 1

I 11/08/88 l 15 Enclosure 1

[7590-01]

an activity licensed under NRC regulations is conducted, or against a component of a plant or transport vehicle and cask which could directly or indirectly endanger the public health and safety by exposure to radiation.

In assessing the probability and consequences of radiological sabo-tage, the NRC considers: (1) the threat to storage facilities; (2)'the response of typical storage casks or vaults and their contained spent {

fuel to postulated acts of radiological sabotage; and (3) the public health consequences of acts of radiological sabotage.

The NRC has carried out studies to develop information about possible adversary groups which might pose a threat to licensed nuclear facilities.

The results of these studies are published in NUREG-0459, " Generic Adver-sary Characteristics - Summary Report" (March 1979) and NUREG-0703,

" Potential Threat to Licensed Nuclear Activities from Insiders" (July 1980). Actions against facilities were found to be limited to a number i of low consequence activities and harassments, such as hoax bomb threats, vandalism, radiopharmaceutical thefts, and firearms discharges. The list of actions is updated annually in a NUREG-0525, " Safeguards Summary Event List" (July 1987). None of the actions have affected spent fuel contain-ment and, thus, have not caused any radiological health hazards. I In addition, the NRC staff regularly consults with law enforcement agencies and intelligence gathering agencies to obtain their views concerning the possible existence of adversary groups interested in radiological sabotage of commercial nuclear facilities. None of the information the staff has collected confirms the presence of an identi-fiable domestic threat to dry storage facilities or to other components of nuclear facilities.

11/08/88 16 Enclosure 1

[7590-01]

Despite the absence of an identified domestic threat, the NRC has considered it prudent to study the response of loaded casks to a range of sabotage scenarios. The study is classified. However, an overview of the study and the conclusions are provided in the following paragraphs.

Being highly radioactive, spent fuel requires heavy shielding for safe storage. Typical movable storage casks are of metal or concrete, weigh 100 tons, and.have wall thickness from 10 to 16 inches of metal or i

30 inches of concrete. The structural materials and dimensions enable the casks and vaults to withstand attack by small arms fire, pyrotechnics, mechanical aids, high velocity objects, and most forms of explosives without release of spent fuel.

After considering various technical approaches to radiological sabotage, the NRC concluded that radiological sabotage, to be successful, would have to be carried out with the aid of explosives. Explosive configurations included air blast, platter, breaching, and shaped charges.

The NRC further concluded that (1) release of spent fuel by sabotage of a storage cask would require skillful use of explosives; (2) large amounts of explosives (tens to hundreds of pounds) would be needed in credible l scenarios; (3) saboteurs interested in most effective use of explosives or larger releases would need to gain temporary control of the storage installation; (4) pound for pound, shaped charges would be more effective j than other explosive configurations; and (5) small shaped charges or sophisticated antitank weapons might penetrate a storage cask, but the hole produced and the subsequent release would be smaller than for the case of a large, manually placed shaped charge.

l 11/08/88 17 Enclosure 1

[7590-01]

The consequences to the public health and safety would stem almost l exclusively from the fraction of the release that is composed of respir-able particles. In an NRC study, an experiment was carried out to evaluate the effects of a very severe, perfectly executed explosive sabo-tage scenario against a simulated storage cask containing spent fuel j assemblies. The amount of fuel disrupted was measured. The fraction of disrupted material of respirable dimensions (0.005%) had been determined in a previous experiment. From this information, an estimate of the air-borne, respirable release was made, and the dose as a function of range i

and other variables was calculated. In a typical situation, for an indi-vidual at the boundary of the reactor site (taken as 100 meters from the location of the release) and in the center of the airborne plume, the whole-body dose was calculated to be 1 rem and the 50 year dose commit-ment (to the lung, which is the most sensitive organ) was calculated to be 2 rem. Doses higher or lower can be obtained depending on the vari-ables used in the calculation. Variables include the meteorological conditions, the age and burn-up of the fuel, the heat-induced buoyancy of the airborne release, the range to the affected individual, and the explosive scenario assumed. I Although the experiment and calculations carried out lead to a con-clusion of low public health consequences, there are limitations that must be taken into account. In particular, consequence modeling assump-tions more severe than those in the foregoing calculation are possible if 1

unconstrained sabotage resources or protracted loss of control of the I storage site are allowed. For that reason protection requirements are proposed to provide for (1) early detection of malevolent moves against I 11/08/88 18 Enclosure 1 am__a..m_ __m._..a.

. [7590-01] {

. the storage site, and (2) a means to quickly summons response resources to assure against protracted loss of control of the site.

The proposed requirements comprise a subset of the overall protec-tion requirements currently in force at every operating nuclear power reactor. Inasmuch as the security force at each reactor is thoroughly familiar with requirements similar to those proposed and has years of experience in carrying them out, the NRC concludes that the requirements can be successfully imposed through a general license for storage of spent fuel in NRC-approved casks without the need for advanced NRC review and approval of a physical security plan or other site-speci,fic document before the reactor licensee implements the requirements.

Material control and accounting (MC&A) requirements are designed to protect against the undetected loss of the special nuclear material in spent fuel by maintaining vigilance over the material, tracking its move-ment and location, monitoring its inventory status, maintaining records l of transactions and movements, and issuing reports of its status at the time of physical inventory. Similar requirements for MC&A have been applied to power reactors, to spent fuel storage at independent spent fuel storage installations, and to operations at certain other classes of fuel cycle facilities without requiring the licensee to submit a plan to document how compliance will be achieved. In these situations the requirements have been found to be sufficient. For these reasons, it is concluded that the MC&A requirements for the dry storage'of spent fuel at power reactors can be handled under a general license.

A minor editorial change to S 72.30(b) is also proposed to make clear that a decommissioning funding plan is an integral part of an applicant's proposed decommissioning plan.

11/08/88 19 Enclosure 1

1 l

l .

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Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Sub-part A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. The rule is mainly administrative in nature and would not change safety requirements, which could have significant environmental impacts. The proposed rule would provide for power reactor , licensees to store spent fuel in casks approved by NRC at reactor sites without addi-tional site-specific approvals by the Commission. It would set forth l conditions of a general license for the spent fuel storage and procedures and criteria for obtaining storage cask approval. The environmental l assessment and finding of no significant impact on which this determina-tion is based are available for inspection at the NRC Public Document Room, 2120 L Street NW., Washington, DC, Lower Level. Single copies of the environmental assessment and the finding of no significant impact are available from W.R. Pearson, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC 20555; Telephone:

(301) 492-3764.

Paperwork Reduction Act Statement This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). This rule has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements.

11/08/88 20 Enclosure 1

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Public reporting burden for this collection of information is estimated to average 1,200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per submittal, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. The submittal referenced above is the submittal of a 1

safety analysis report (SAR) by spent fuel storage cask vendors. Review l

and approval of an SAR is necessary in order to obtain a Certificate of Compliance for a cask design from NRC. A Certificate of Compliance is required for each cask before it can be used for spent fuel storage under the general license in this rule. The only response from power i

reactor licensees required under this rule would be registration. An informational letter would be submitted to NRC, thus, there would be no significant reporting burden. Send comments regarding this burden I estimate or any other aspect of this t .lection of information, includ-ing suggestions for reducing this burden, to the Records and Reports Management Branch, Mail Stop P-530, Division of Information Support Services, Office of Administration and Resources Management, U.S.

Nuclear Regulatory Commission, Washington, DC 20555; and to the Office of Information and Regulatory Affairs, Office of Management and Budget, Washington, DC 20503.

Regulatory Analysis The Commission has prepared a preliminary regulatory analysis on this proposed rule. The analysis examines the benefits and impacts considered by the Commission. The Preliminary Regulatory Analysis is available for inspection in the NRC Public Document Room, 1717 H Street NW., Washington, DC. Single copies may be obtained from W.R. Pearson, 1

11/08/88 21 Enclosure 1

[7590-01]

Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC 20555; Telephone: (301)492-3764.

The Commission requests public comments on the preliminary regula-tory analysis. Comments on the preliminary regulatory analysis may be submitted to the NRC as indicated under the ADDRESSES heading.

Regulatory Flexibility Act Certification As required by the Regulatory Flexibility Act of 1980 (5 U.S.C.

605(b)), the Commission certifies that this rule, if adopted, will not have a significant economic impact on a substantial number of small entities. This proposed rule affects only licensees owning and operating nuclear power reactors. The owners of nuclear power plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121.

Backfit Analysis i

The NRC has determined that the backfit rule, 10 CFR 50.109, does j not apply to this proposed rule, and, thus, a backfit analysis is not i required for this proposed rule, because these amendments do not involve any provisions which would impose backfits as defined in S 50.109(a)(1). l i

11/08/88 22 Enclosure 1

[7590-01]

l

, List of Subjects Part 72: Manpower training programs, Nuclear materials, Occupational safety and health, Reporting and recordkeeping requirements, Security measures, Spent fuel.

Part 170: Byproduct material, Nuclear materials, Nuclear power plants and reactors; Penalty, Source material, Special nuclear material.

f For reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of j 1974, as amended, the Nuclear Waste Policy Act of 1982, and 5 U.S.C. 552

]

and 553, the NRC is proposing to adopt the following revisions to 10 CFR Part 72 and conforming amendments to 10 CFR Part 170.

l PART 72 - Licensing Requirements for the Independent Storage i

I of Spent Nuclear Fuel and High-Level Radioactive Waste

1. The authority citation for Part 72 is revised to read as follows:

Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953, 954, l 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233, 2234, 2236, 1

2237, 2238, 2282); sec. 274, Pub. L.86-373, 73 Stat. 688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); Pub. L.95-601, j sec. 10, 92 Stat. 2951 (42 U.S.C. 5851); sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135, 137, 141, 11/08/88 23 Enclosure 1

[7590-01]

I 1

. Pub. L.97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148, Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153, 10155,10157,10161,10168).

Section 72.44(g) also issued under secs. 142(b) and 148(c), (d),

Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b),

10168(c)(d)). Section 72.46 also issued under sec. 189, 68 Stat. 955 (42 U.S.C. 2239); sec. 134, Pub. L.97-425, 96 Stat. 2230 (42 U.S.C.

10154). Section 72.96(d) also issued under sec. 145(g), Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also issued under j secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L.97-425, 96, Stat. 2202, )

2203, 2204, 2222, 2244 (42 U.S.C. 10101, 10137(a), 10161(h)). Sub- I parts K and L are also issued under sec. 133, 96 Stat. 2230 (42 U.S.C. l 10153) and 218(a), 96 Stat. 2252 (42 U.S.C. 10198).

For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C.

2273); SS 72.6, 72.22, 72.24, 72.26, 72.28(d), 72.30, 72.32, 72.44(a),

(b)(1),(4),(5),(c),(d)(1),(2),(e),(f),72.48(a),72.50(a),

72.52(b),72.72(b),(c),72.74(a),(b),72.76,72.78,72.104,72.106, <

72.120,72.122,72.124,72.126,72.128,72.130,72.140(b),(c),72.148, 72.154, 72.156, 72.160, 72.166, 72.168, 72.170, 72.172, 72.176, 72.180, 72.184, 72.186 are issued under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b)); SS 72.10(a), (e), 72.22, 72.24, 72.26, 72.28, 72.30, 72.32,72.44(a),(b)(1),(4),(5),(c).,(d)(1),(2),(e),(f),72.48(a),

72.50(a),72.52(b),72.90(a)-(d),(f),72.92,72.94,72.98,72.100, 72.102(c),(d),(f),72.104,72.106,72.120,72.122,72.124,72.126, 72.128,72.130,72.140(b),(c),72.142,72.144,72.146,72.148,72.150, 72.152, 72.154, 72.156, 72.158, 72.160, 72.162, 72.164, 72.166, 72.168, 72.170, 72.172, 72.176, 72.180, 72.182, 72.184, 72.186, 72.190, 72.192, 11/08/88 24 Enclosure 1

. [7590-01] i I

. 72.194 are issued under sec. 1611, 68 Stat. 949, as amended (42 U.S.C.

2201(i)); and SS 72.10(e), 72.11, 72.16, 72.22, 72.24, 72.26, 72.28, 72.30, 72.32, 72.44(b)(3), (c)(5), (ci)(3), (e), (f), 72.48(b), (c),

1 72.50(b),72.54(a),(b),(c),72.56,72.70,72.72,72.74(a),(b),

72.76(a),72.78(a),72.80,72.82,72.92(b),72.94(b),72.140(b),(c),

(d),72.144(a),72.146,72.148,72.150,72.152,72.154(a),(b),72.156, 72.160, 72.162, 72.168, 72.170, 72.172, 72.174, 72.176, 72.180, 72.184, 72.186, 72.192, 72.212(b), 72.216, 72.218, 72.230, 72.234(e) and-(g) are issued under sec. 1610, 68 Stat. 950, as amended (42 U.S.C. 2201(o)).

l

2. In S 72.30, paragraph (b) is revised to read as follows:

S 72.30 Decommissioning planning, including financing and recordkeeping.

(b) [The-decommissioning-funding plan-must-contain] The proposed decommissioning plan must also include a decommissioning funding plan containing information on how reasonable assurance will be provided that funds will be available to decommission the ISFSI or MRS. This informa-tion must include a cost estimate for decommissioning and a description of the method of assuring funds for decommissioning from paragraph (c) of this section, including means of adjusting cost estimates and asso- l l

ciated funding levels periodically over the life of the ISFSI or MRS.

3. New Subpart K and Subpart L are added to read as follows:

1 1

11/08/88 25 Enclosure 1 1 .

1

._-__-___-._-___w

i i

. [7590-01] l l

. Subpart K - General License for Storage of Spent Fuel at Power Reactor Sites Sec.

l l

72.210 General license issued.

72.212 Conditions of general license issued under S 72.210.

72.214 List of approved spent fuel storage casks.

72.216 Reports.

72.218 Termination of the licenses.

72.220 Violations. ,

Subpart L - Approval of Spent Fuel Storage Casks  !

72.230 Procedures for spent fuel storage cask submittals.

.I 72.232 Inspection and tests.

i 72.234 Conditions of approval.

72.236 Specific criteria for spent fuel storage cask approval.

72.238 Issuance of an NRC Certificate of Compliance.

72.240 Conditions for spent fuel storage cask reapproval.

Subpart K - General License for Storage of Spent Fuel at Power Reactor Sites .

S 72.210 General license issued.

A general license is hereby issued for the storage of spent fuel in an independent spent fuel storage installation at power reactor sites to 11/08/88 26 Enclosure 1

[7590-01]

persons authorized to operate nuclear power reactors under Part 50 of this chapter.

S 72.212 Conditions of general license issued under S 72.210.

(a)(1) The general license is limited to storage of spent fuel in casks approved under the provisions of this part.

(2) The general license for each cask fabricated under a Certificate of Compliance shall terminate 20 years after the date that 1

the cask is first used to store spent fuel, unless the cask model is reapproved. In the event that a cask vendor does not apply for a cask l model reapproval under 6 72.240, any user or user representa'tive may apply for cask reapproval.

(

(b) The general licensee shall:

(1)(i) Notify the Nuclear Regulatory Commission under 672.4 at i l

least 90 days prior to first storage of spent fuel under the general license. The notice may be in the form of a letter, but must contain the licensees name, address, reactor license number (s), and the name and means of contacting a person for additional information. A copy of the submittal must be sent to the Administrator of the appropriate Nuclear Regulatory Commission regional office listed in Appendix D to Part 20.

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(ii) Register use of each cask with the Nuclear Regulatory Commis- i sion no later than 30 days after using the cask to store spent " fuel.

This registration may be accomplished by submitting an NRC Form-xxx or by a letter containing the following information: the licensee's name  ;

and address, the licensee's reactor license number (s), the name and title of a person who can be contacted for additional information, the l

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. cask certificate or model number, and the cask identification number. Sub-mittals must be in accordance with the instructions contained in S 72.4 of this part. A copy of each submittal must be sent to the Administrator of the appropriate Nuclear Regulatory Commission regional office listed in Appendix D to Part 20.

(2) Perform written evaluations showing that conditions set forth in the Certificate of Compliance are met for the anticipated total number of casks to be used for storage. The licensee shall also show that cask storage pads and areas are designed to adequately support the static load of the stored casks. Evaluations must show that the requirements of 6 72.104 of this part are met. A copy of this record must be retained for 3 years.

(3) Determine, using precedures and criteria in S 50.59 of this chapter, whether activities under this general license involve any unreviewed safety questions or changes in the facility technical specifi-cations, including activities related to spent fuel storage casks. If any Nuclear Regulatory Commission approval is required, the procedure set forth in Part 50 of this chapter for this type of approval must be followed. A copy of the evaluation must be retained by the licensee for tnree years after initial storage of spent fuel under the general license.

(4) Protect the spent fuel against the design basis threat of radio-logical sabotage in accordance with the licensee's physical security plan approved in accordance with 6 73.55 of this chapter, with the followina additional conditions and exceptions:

(i) The physical security organization and program must be expanded and modified as necessary to assure that activities conducted under this general licensee do not decrease the effectiveness of the protection of vital equipment in accordance with 6 73.55 of this chapter.

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i (ii) Storage of spent fuel must be within a protected area, in accordance with 6 73.55(c) of this chapter, but need not be within a separate vital area. Existing protected areas may be expanded or new protected areas added for the purpose of storage of spent fuel in accordance with this general license.

(iii) Notwithstanding any requirements of the licensee's approved i

security plan, the observational capability required by S 73.55(h)(6) '

of this chapter may be provided by a guard or watchman in lieu of closed circuit television for protection of spent fuel under the provisions of this general license.

(iv) For the purposes of this general license, the licensee is  !

exempt from 6 73.55(h)(4)(iii)(A) and (5) of this chapter.

l (5) Establish and maintain as current records an emergency plan, a l quality assurance program, a training program, and a radiation protec-tion program for activities related to storage of spent fuel under the general license until the general license is terminated.

(6) Maintain a copy of the Certificate of Compliance and documents l

referenced in the certificate for each model of cask used for storage of spent fuel, until use of the cask model is discontinued. The licensee  !

shall comply with the terms and conditions of the certificate.

(7)(i) Maintain the record provided by the cask supplier for each l

cask that shows:

(A) The NRC Certificate of Compliance number; (B) The name and address of the cask vendor / lessor; (C) The listing of spent fuel stored in the cask; and (D) Any maintenance performed on the cask.

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(ii) This record must include sufficient information to furnish documentary evidence that any testing and maintenance of the cask has been conducted under a quality assurance program accepted by the Nuclear Regulatory Commission.

(iii) In the event that a cask is sold, leased, loaned, or otherwise transferred, this record must also be transferred to and must be accurately maintained by the new user. This record must be maintained by the current cask user during the period that the cask is used for storage of spent fuel and retained by the last user for 3 years following decommissioning of the cask.

(8) Conduct activities related to storage of spent fuel under this general license in accordance with written procedures.

(9) On reasonable notice the licensee shall make records available to the Commission for inspection.

S 72.214 List of approved spent fuel storage casks.

The following casks have been reviewed and evaluated by the Commission and are approved for storage of spent fuel under the conditions specified in their respective Certificates of Compliance.

Certificate Number: [To be completed later].

SAR Submitted by: General Nuclear Systems, Inc. .

SAR

Title:

" Topical Safety Analysis Report for the Castor V Cask Independent Spent Fuel Storage Installation (Dry Storage)"

Docket Number: 72-1000 Certification Expiration Date:

Model Number: CASTOR V/21 11/08/88 30 Enclosure 1

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Certificate Number: [To be completed later].

SAR Submitted by: Westinghouse Electric Corp.

SAR

Title:

" Topical Safety Analysis Report for the Westinghouse MC-10 Cask for an Independent Spent Fuel Storage Installation (Dry Storage)."

Docket Number: 72-1001 Certification Expiration Date:

Model Number: MC-10 Certificate Number: [To be completed later].

t SAR Submitted by: Nuclear Assurance Corp.

SAR

Title:

" Topical Safety Analysis Report for the NAC Storage /

Transport Cask for use at an Independent Spent Fuel Storage Installation."

Docket Number: 72-1002 Certification Expiration Date:

Model Number: Storage / Transport Certificate Number: [To be completed later].

SAR Submitted by: Nuclear Assurance Corp.

SAR

Title:

" Topical Safety Analysis Report for NAC Storage /

Transport Cask Containing Consolidated Fuel - Fo.r_

~

use at an Independent Spent Fuel Installation."

Docket Number: 72-1003 Certification Expiration Date: [To be completed later].

Model Number: NAC-C28 S/T 11/08/88 31 Enclosure 1

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1 6 72.216 Reports.

(a) The licensee shall make an initial report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to l

the Director, Office of Nuclear Materials Safety and Safeguards, and a resident inspector at the reactor site, of any:

(1) Defect with safety significance discovered in any cask; and (2) Instance in which there is a significant reduction in the 1

safety effectiveness of any cask during use.

(b) A written report, including a description of the means employed 1

to repair any defects or damage and prevent recurrence, must be submitted i in accordance with 6 72.4 within 30 days. A copy of the written report i r I mustbesenttotheAdminisgatoroftheappropriateNuclearRegulatory Commission regional office shown in Appendix D to Part 20 of this Chapter. I S 72.218 Termination of licenses.

(a) The notification regarding planning for the management of all spent fuel at the reactor required by S 50.54(bb) of this chapter must include a plan for removal of the spent fuel stored under this general license from the reactor site. The plan must show how the spent fuel l will be managed before starting to decommission systems and components needed for moving, unloading, and shipping this spent fuel.

(b) Spent fuel previously stored may continue to be stored under this general license after termination of the reactor license under S 50.82 of this chapter. An application for termination of the reactor operating license submitted under S 50.82 of this chapter must, however, contain a description of how the spent fuel stored under this general license will be removed from the reactor site. If the decommissioning mode selected under S 50.82 is likely to extend beyond 30 years after 11/08/88 32 Enclosure 1

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the expiration date of the reactor operating license, the licensee shall include in the application a discussion of incremental environmental impacts of the extended spent fuel storage.

(c) The reactor licensee shall send a copy of submittals under S 72.218(a) and (b) to the Administrator of the appropriate Nuclear Regulatory Commission regional office shown in Appendix D to Part 20 of this Chapter.

S 72.220 Violations.

Storage of spent fuel under a general license may be halted or terminated under S 72.84.

Subpart L - Approval of Spent Fuel Storage Casks S 72.230 Procedures for spent fuel storage cask submittals.

(a) An application on NRC Form-xxxx must be submitted in accordance with the instructions contained in S 72.4. A safety analysis report describing the proposed cask design and how the cask should be used to store spent fuel safely must be included with the application.

(b) Casks that have been certified for transportation of spent fuel under Part 71 of this chapter may be approved for storage of spent fuel under this subpart. An application on NRC Form-xxxx must be submitted in accordance with the instructions contained in S 72.4. A copy of the Certificate of Compliance issued by the NRC for the cask, and d'rawings and other documents referenced in the certificate, must be included with the application. A safety analysis report showing that the cask is suitable for storage of spent fuel for a period of at least 20 years must also be included.

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(c) Public inspection. An application for the approval of a cask for storage of spent fuel may be made available for public inspection under 6 72.20.

(d) Fees. (1) Fees for review and evaluation related to issuance of a spent fuel storage cask Certificate of Compliance are those shown in 6 170.31 of this chapter.

(2) Fees for inspections related to spent fuel storage on reactor sites and vendor inspection of dry storage casks are those shown in 6 170.32 of this chapter.

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l 6 72.232 Inspection and tests.

(a) The applicant shall permit, and make provisions for, the l Commission to inspect at reasonable times the premises and facilities at which a spent fuel storage cask is fabricated and tested.

(b) The applicant shall perform, and make provisions that permit the Commission to perform, tests that the Commission deems necessary or appropriate for the administration of the regulations in this part.

i (c) The applicant shall notify the Chief, Vendor Inspection Branch, Division of Reactor Inspection and Safeguards, Office of Nuclear Reactor l Regulation, and the Director, Division of Industrial and Medical Nuclear Safety, Office of Nuclear Materials Safety and Safeguards, U.S. Nuclear i

Regulatory Commission, Washington, DC, 20555, at least 45 days prior to '

starting fabrication of any spent fuel storage cask.

6 72.234 Conditions of approval.

(a) Design, fabrication, testing, and maintenance of a spent fuel storage cask must comply with the technical criteria in 6 72.236.

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(b) Design, fabrication, testing, and maintenance of spent fuel storage casks must be conducted under a quality assurance program that i

meets the requirements of Subpart G of this part. )

)

(c) Cask fabrication must not start prior to receipt of the Certifi-  !

cate of Compliance for the cask model.

(d) Cask model reapproval is required 20 years after the issuance of a Certificate of Compliance. Any applicant under S 72.230, who receives a Certificate of Compliance under S 72.238, shall notify the Commission if an application for cask reapproval will not be submitted.

(e)(1) The applicant shall ensure that a record is established and maintained for each cask fabricated under the NRC Certificate of Compliance.

(2) This record must include:

(i) The NRC Certificate of Compliance number; i (ii) The cask model number; l (iii) The cask identification number; 1 '

(iv) Date fabrication started; (v) Date fabrication completed; (vi) Certification that the cask was designed, fabricated, tested, and repaired in accordance with a quality assurance program accepted by NRC; (vii) Certification that inspections required by 6 72.236(j) were performed and found satisfactory; and (viii) The name and address of the cask user.

l (3) A copy of this record must be submitted to the Commission in 1

I accordance with instructions contained in S 72.4 and the original of the record supplied to the cask user. A current copy of a composite record of 11/08/88 35 Enclosure 1

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all casks, showing the above information, must be retained by the applicant for 20 years after the cask is shipped.

(f) The cr ocsite record required by S 72.234(e)(3) must be made available to tM fpnmission for inspection.

(g) The applicant shall ensure that written procedures and appropriate tests are established for use of the casks. A copy of these procedures and tests must be provided to each cask user.

l 6 72.236 Specific criteria for spent fuel storage cask approval.

(a) Technical specifications concerning the spent fuel' to be stored in the cask, such as the type of spent fuel (i.e., BWR, PWR, both),

enrichment of the unitradiated fuel, burn-up (i.e., megawatt-days /MTU),

cooling time of the spent fuel prior to storage in the cask, maximum heat designed to be dissipated (i.e., kw/ assembly, kw/ rod), the maximum spent fuel loading limit, and condition of the spent fuel (i.e., intact assembly or consolidated fuel rods), inerting atmosphere requirements, I must be provided. '

(b) Design bases and design criteria must be provided for struc-

, tural members and systems important to safety.

(c) The cask must be designed and fabricated so that the spent fuel is maintained in a subcritical condition under credible conditions.

(d) Radiation shielding and confinement features must be brovided to the extent required to meet the requirements in SS 72.104 and 72.106 of this part.

(e) Casks must be designed to provide redundant sealing of con-l finement systems.

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i (f) Casks must be designed to provide adequate heat removal capac-ity when the cask is stored without active cooling.

(g) Casks must be designed to store the spent fuel safely for a minimum of 20 years and permit maintenance as required.

(h) Casks must be compatible with wet or dry spent fuel loading and unloading facilities.

(i) Casks must be designed to facilitate decontamination to the extent practicable.

1 (j) Casks must be inspected to ascertain that there are no cracks, '

pinholes, uncontrolled voids, or other defects that could significantly reduce their confinement effectiveness.

(k) Casks must be conspicuously and durably marked with (1) A model number; (2) A unique identification number; and (3) An empty weight.

l (1) Casks and systems important to safety must be evaluated, by l j subjecting a sample or scale model to tests appropriate to the part being l

tested, or by other means acceptable to the Commission, demonstrating j that they will reasonably maintain confinement of radioactive material under normal, off-normal, and accident conditions.

(m) To the extent practicable, in the design of dry spent fuel ,

I storage casks, consideration should be given to the compatibility of the  ;

I dry storaoe cask systems and components with transportation and other

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activities related to the removal of the stored spent fuel from the i reactor site for ultimate disposition by the Department of Energy.

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11/08/88 37 Enclosure 1 l 1

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6 72.238 Issuance of an NRC Certificate of Compliance.

l A Certificate of Compliance for a cask model will be issued by NRC l l

on a finding that

(

(a) The criteria in 6 72.236(a) through (i) are met; and (b) The applicant certifies that each cask will be fabricated, inspected, and tested in accordance with 6 72.236(j) and (1).

l 6 72.240 Conditions for spent fuel storage cask reapproval.

(a) The holder of a cask model Certificate of Compliance, a user l l

of a cask model approved by NRC, and representatives of cask users may I apply for a cask model reapproval.

1 (b) Application for reapproval of a cask model must be submitted j 3 years prior to the date that the Certificate of Compliance for that I

{

model expires. The application must be accompanied by a safety analysis report (SAR). The new SAR may reference the SAR originally submitted for the cask model (c) A cask model will be reapproved if conditions in 6 72.238 are met, including demonstration that storage of spent fuel has not signif-icantly, adversely affected systems and components important to safety.

Part 170 - Fees for Facilities and Materials licenses and Other Regulatory Services Under the Atomic Energy Act .

I of 1954, as Amended

4. The authority citation of Part 170 continues to read as follows:

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AUTHORITY: 31 U.S.C. 9701, 96 Stat. 1051; sec. 301, Pub. L.92-314, 86 Stat. 222 (42 U.S.C. 2201w); sec. 201, 88 Stat.1242, as a:nended (42 U.S.C. 5841).

5. In S 170.31, a new category 13 is added to read as follows (note: Footnotes to the chart remain unchanged): ,

l S 170.31 Schedule of fees for materials licenses and other regulatory services.

t i Category of materials and type Fee 2 of feel

-13. Spent fuel storage cask Certificate of Compliance Application $150 Approvals:

Amendments, Revisions and Supplements Full Cost 3 Reapproval Full Cost 3

6. In S 170.32, Category 13 is added to read as follows:

(Note: Footnotes to the chart remain unchanged):

S 170.32 Schedule of fees for health and safety, and safeguards inspections for materials licenses.

11/08/88 39 Enclosure 1

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4 Type of Maximum l Category of licensees inspection 1 Fee 2 frequency 3 j l

13. A. Spent fuel Routine Full Per l storage cask Cost 4 inspection  !

Certificate of I Compliance Non-routine --do-- --do-- l B. Storage of spent Routine Full Per fuel under Cost 4 inspection  !

572.210 '-

Non-routine --do-- --do-- {

4

• A A A Dated at Rockville, Maryland, this day of , 1988.

For the Nuclear Regulatory Commission.

1 Samuel J. Chilk, Secretary of the Commission.

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11/08/88 40 Enclosure 1  ?

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