ML20198S208
| ML20198S208 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1997 |
| From: | NRC |
| To: | |
| References | |
| NUREG-0750, NUREG-0750-V46-N02, NUREG-750, NUREG-750-V46-N2, NUDOCS 9801260109 | |
| Download: ML20198S208 (34) | |
Text
{{#Wiki_filter:! i i l ' NUREG-0750 ) Vol. 46, No. 2 l Pages 21 -48
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NUCLEAR REGULATORY
- COMMISSION ISSUANCES '
i ! August 1997 gnREGy i h 0, h
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U S. NUCLEAR REGULATORY COMMISSION .
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t I f Available from Superintendent of Documents U.S. Government Printing Office ' P.O. Box 37082 i Washington, DC 20402-9328 A year's subscription consists of 12 softbound issues, i 4 indexes, and 2-4 hardbound editions for this publication. Single copies of this publication are available from National Technical Information Service Springfield, VA 22181 l I Errors in this publication may be reported to the
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Office of the Chiei Information Officer U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 (301 -415 - 6844)
l l i ) : NUREG-0750 !' l i Vol. 46 No. 2 Pages 21-40 i ! l NUCLEAR REGULATORY l 4 COMMISSION ISSUANCES : l l i August 1997 l l This report includes the issuances received during the specified i period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administ,ative Law Judges (ALJ), the Directors' i Decisions (DD), and the Decisions on Petitions for Rulemaking (DPRM) The summaries and headnotes preceding the opinione reported hereir' are not to be deemed a part of those opinions or have any l independent legal signific nce. U S.' NUCLEAR REGULATORY COM (SSION 4 Prepared by the , Office of the Chief information Officer U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 1 (301 - 415- 6844) l
- p. ,
COMMISSIONERS Shirley A. Jackson, Chairman Greta J. Dicus Nils J. Diaz Edward McGaffigan, Jr. B. Paul Cotter, Jr., Chief Administrative Judge, Atomic Safety & Licensing Board Panel
CONTENTS e !-
. lsmaances of the Nucles/ Regulatory C="*
ATLAS CORPORATION (Moab. Utah Facility) Docket 40 3453-MLA 21 MEMORANDUM AND ORDER, CLI 97 8, AuFust 4,1997 . . .. . . . , L IN111RN A110NAl, URANIUM (USA) CORPORA *I* 0N (White Mesa Uranium Mill) Docket 40-8631 MLA (Alternate fiwi Material) 23 - MEMORANDUM AND ORDER, CL197 9 August 7,1997 . . . . . . . . RALPH L TETRICK
- (Denial of Application for Reactor Operator License)-
' Dociet $5 207245P 26 MEMORANDUM AND ORDER, CLI 9710, August 7,1997 . . . . . .
lesuance of IMrector's Deciolon - NOR111ERN STA1115 POWER ,t;COk ANY (Praitic Island Nuclear Generat.ng ) Prairie Island Independent Spent l'uct Storage Installation) Dockcts $0 282,50-306,7210 DIRl!CIDR'S D11 CISION UNDER 10 C.F.R. U.206. 35 DD-97 lH. August 29,1997......... l
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,- t Cite as 46 NRC 21 (1997) CL1-97 8 , ~ ,
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION COMMISSIONERS: + Shirley Ann Jackson, Chairman Greta J. Dicus Nils J. Dia - Edwstd McGaffigan, Jr. In the Matter of Docket No. 40 3453-MLA : ATLAS CORPORATION (Moab, Utah Facl!ity) August 4,1997 1he Commission concluded that the Petitioner for intervention had failed to demonstrate standing. The Commission therefore denied his appeal of the Presiding Officer's order denying his petition to intervene. COMMISSION PROCEEDINGS: APPELLATE REVIEW RULES OF PRACTICE: STANDING The Commission ordinarily defers to Licensing Board standing deterinina, tions, and sees no legal error or abt.;c of discretion in the P;esiding Officer's
**fusal to grant standing to a Petitioner to intervene, given his failure to offer .ove than general responses to the Presiding Officer's reasonable and clearly .e slated requests for more specific information about ble proximity-based
- w. ding claims. The four opportunities that the Petitioner had to specify his
' aims were entirely adequate. MEhlORANDUM AND ORDER The Commission denies the appeal filed by Mr. John F. Darke June 2,1997, and affirms the Presiding Officer's order, LBl%? 9,45 NRC 414 (1997). We - 21
= ,
do 66 on the grounds set forth in LilP-97-9. In that order, the Presiding Officer , _,( rejected Mr. Darke's request for a formal hearing, concluded that Mr. Darke had not met his burden to estrblish standing to intervene, and dismissed the-. ! proceeding. We ordinarily defer to Licensing Doard standing determinations. See Dniet Atom!c flectric Co. (Yankee Nuclear Power Station), CLI.96-7,43 Nf f 23$. 248 (1996). Ilere, we see no legal erne or abuse of discretion in the P.: tiding Officer's refusal to grant standing to Mr. Duke, given his failure to offer more than general responses to the IYesiding Officer's reasonable and clearly articulated requests for more specific information about Mr. Darke's proxirrity-based standing claims. The four opportunities that Mr. Darke had to specify his _ , , claims were entirely adequate. ! IT IS SO ORDERED. Ihr the Commission P JOllN C. IlOYl.E Secretary of the Commission Dated at Rockville, Maryland, this 4th day of August 1997. 22
h i Cite as 46 NRC 23 (1997) - CLI-97 9 j,.s , UNITED STATES OF AMERICA . NUCLEAR REGULATORY COMMISSION : COMMISSIONERS: 1 Shirley Ann Jackson, Chairman ! Greta J. Dicus Nils J. Diaz Edward McGaffigan, Jr. in the Matter of Docket No. 40-8681 MLA (Alternate Food Material) INTERNATIONAL URANIUM (USA) CORPORATION
-(White Mesa Uranium Mill) August 7,1997 In response to a letter that included (1) an appeal of a Presiding Officer's decision denying Petitioners standing, (2) moving for reconsideration of the decision, and (3) moving to reopen the record, the Commisshn instructs the Presiding Officer to pass upon the two motions. The Commission concludes that the Presiding Officer's greater familiarity with the prior proceeding and pleadings in this case rendered him better equipped than the Commission to make prompt initial rulings on' the merits of the two motions.
COhlhilSSION PROCEEDINGS: APPELLATE REVIEW RULES OF PRACTICE: APPELLATE FILINGS; APPELLATE REVIEW (INTERVENTION DENIALS); LICENSING llOARDS; hlOTION FOR RECONSIDERATION; hlOTION TO REOPEN RECORD; RECONSIDERATION PETITIONS
'the Commission disapproves of the practice of simultaneously seeking reconsideration of a Presiding Officer's decision and filing an appeal of the same ruling, because taking that approach would call for rulings on the same issues at the same time from both a trial and appellate forum.
23 m- _ - _ , ~ , - . ,
ljCENSING BOARDS: JURISDICTION ^ ^ ADJUDICATORY BOARDS: - JURISDICTION , RULES OF PfMCTICE: APPEI LATE FILINGS; APPELLATE REVIEW (INTERVENTION DFNIALS); JURISDICTION (LICENSING BOARDS, PRESIDING OFFICER)! LICENSING llOARDS; MOTION-FOR RECONSIDERATION; MOTION TO REOPEN RECORD; RECONSIDERATION PETITIONS Because the Presiding Officer's greater familiarity with the prior proceedmg and pleadings in this case renders him better equipped than the Commission to make prompt initial rulings on the merits of the motions for reconsideration and reopening of the record, the Commission instructs him to pass upon th<ne motions, notwithstanding the pendency of the appeal. MEMORANDUM AND ORDER On July 30, 1997, three Petitioners jointly submitted a letter to Chairman
.lackson' styled as an " appeal" of the Presiding Officer's order (LBP-9712,46 s NRC 1 (1997)) rejecting their claims of standing, he same letter also asked the Presiding Officer to reconsider his decision and to reopen the record.
He Commission disapproves of the practice of simultaneously seeking reconsideration of a Presiding Officer's decision and filing an appeal of the same ruling, llouston Ligt. ting and Power Co. (Allens Creek Nuclear Generating Station, Unit I), ALAll 630,13 NRC 84, 85 (1981), because taking that approach would call for rulings on the same issues at the same time from both a trial and appellate forum. IIere, the Presiding Officer's greater familiarity with the prior proceeding and pleadings in this case renders him better equipped than the Commission to make prompt initial rulings on the merits of the motions for reconsideration and reopening of the record. See Curators of the Uniwrsity of Alissouri, CL1951, 41 NRC 71,94 (1995). We therefore instruct him to pass upon the two motions on their merits expeditmusly, notwithstanding the pendency of the appeal. See Portland General Electric Co. (Trojan Nuclear Plant), ALAB 627,13 NRC 20, 21 n.6 (1981). We will take appropriate action on the appeal after the Presiding I Along with this order, we are norstng cryies or the uppe.nl letter on the Pteuding ofhcet and on the other pianes to the adjudwatson All plembags in Comnusuon nJ udwauons, 3 esen in mformal Subpart 1. proceedings, nest be accompanied by an appropemie cevuticate of serv 6ce See 10 C F R 662.120Ne) 1712. This neuer contaned no ceruhcate of servies and apparently was not actually served We cauthm the partws to pay heed to the ceruskase of servut requivements in the future. 24
t 4 Officer decides whether to grant or deny the requests for reconsideration ased' - reopening. t
'IT IS SO ORDERED.
For the Commission JOllN C.110YLE Secretary of the Commission Dated at Rockville, Maryland, this 7th day of August 1997. b
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Cite as 46 NRC 26 (t997) - CLi-9710_ p* i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION - COMMISSIONERS: Shirley Ann Jackson, Chairman Greta J. Dicus Nils J. Diaz Edward McGaffigan, Jr. In the Matter of - Docket No. 55-20726-SP
' RALPH L TETRICK
- (Denial of Application for Reactor
. Operator License) August 7,1997 i
The Commission grants the NRC Staff's Petition for Review and reverses j the Presiding Officer's decision requiring issuance of a Senior Reactor Operator (SRO) license. The Commission disagrees with the Presiding Officer's con-clusions that the NRC Staff should have anticipated the need to offer evidence and arguments on the issue whether the candidate's SRO examination score should be rounded up to a passing grade of 80%, and that the Staff's failure to anticipate this need precludes their raising such arguments and evidence on reconsideration. 'the Commission also disagreed with the Presiding Officer's decision to round up the SRO examination score, but agreed with him that the :
- candidate had incorrectly answered Question 63 of the examination.
TECHNICAL ISSUES DISCUSSED: SENIOR REACTOR OPERATOR EXAMINATION (ROUNDING OF GRADE) l . Although the Statf could reasonably have anticipated both that he might rule in the SRO candidate's favor on one of the exam questions and that such a ruling
- would raise his score to either a 79.59 (question deleted) or 79.80 (question graded in candidate's favor), the Staff need not have further anticipated that the .
, Presiding Officer would then round the revised score upwa.J to the next integer. I L -j 26 i
='
TECHNICAL ISSUES DISCUSSED: SENIOR REACTOR OPERATOR 4 EXAMINATION (ROUNDING OF GRADE)
- REGULATORY GUIDES: APPLICATION The version of NUREG 1021 in effect at the time the candida;c took his exam did not address rounding directly but did state that a successful applicant ; must answer correctly "at least 80 percent" of the questions on the written examination. The phrase "at least" on its face suggesh strongly that 80% is the minimal acceptable score and that rounding up lower scores is impermissible, TECHNICAL ISSUES DISCUSSED: - SFNIOR REACTOR OPERATOR EXAMINATION (ROUNDING OF GRADE)
AGENCY PRACTICE Agency practice is one indicator of how an agency interprets its regulations. , Given that the Staff itself set the 80% threshold in the first place, the Commission -
= is disinclined to disturb its consistently held view.
TECHNICAL ISSUES DISCUSSED: SENIOR REACTOR OPERATOR EXAMINATION (ROUNDING OF GRADE) REGULATORY GUIDES: APPLICATION He NRC's recent revision of NUREG-1021 to replace the minimum passing grade of "80 percent" with "80.00 percent" does not support an implication that the former term permitted rounding and therefore needed correction. Rather, the revision was akin both to the clarifying regulatory amendments that the Commission and other agencies regularly promulgate and to the clarifying legislation that Congress regularly enacts. ' LICENSING llOARDS: SCOPE OF REVIEW; REVIEW OF NRC STAFF'S ACTIONS; RESPONSililLITIES ADJUDICATORY llOARDS: ROLE REGULATORY GUIDES: APPLICATION SENIOR OPERATOR LICENSE: SCOPE OF INQUIRY
%c decision whether to round up near-passing scores requires a policy cl.aice. Either option is plausible. Ilere, in the adjudicatory setting, the Commission declines to set aside the NRC Staff's policy judgment, supported -
- by the language of NUREG 1021, to draw the pass fait line at 80% minimum, 27 F
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without rounding up.; When the Presiding Officer ordered rounding up_'on the ~
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- ground that the SRO written examinations are not so precise that tenths' of a'.
J percent have any meaning and essentially reduced the passing score from 80% Lt o 79.5%, he stepped into a Staff area of responsibility. ; i MEMORANDUM'AND ORDER ! On Ibbruary 28,1997, the Presiding Officer issued an hiitial Decision in this ; proceeding concluding that Ralph L Tetrick, who is currently a reactor operator - at the Turkey Point Nuclear Generating Plant (Units 3 and 4), had answered correctly seventy-eight out of ninety-eight valid questions on his Senior Reactor Operator (SRO) written examination, As a result of this ruling, the Presiding l Officer revised Mr. Tetrick's score upwards to 79.59%. He Presiding Officer then rounded Mr. Tetrick's revised score upwards still further - to the nearest -
~ integer,80 -, the eby giving him a passing grade on the written examination.
; LDP-97 2,45 NRC 51, reconsiderarlon denied, LBP.97-6,45 NRC 130 (1997).-
5%e Presiding Officer accordingly ordered issuance of an SRO license to Mr. , 1Tetrick.1The NRC Staff has filed a petition for review seeking Commission reversal of the Presiding Officer's decision. -
- Mi. Tetrick, in addition to supporting the Presiding Officer's ruling on the
-" rounding" issue, also asserts _ as an alternative ground for affinnance that he .
should be given credit for a^ correct answer to Question 63 of the written . SRO examination.' (%e Presiding Officer had found that Mr. Tetrick's answer - -l was incorrect, Sec 45 NRC at 53-5$.) Recently, because of new information submitted to the Commission, we remanded the Question 63 issue for further consideration by the Presiding Officer. CLI 97 5,45 NRC 355 (1997). On remand, the Presiding Officer issued a Memorandum and Order again concluding
- that Mr, Tetrick's answer to Question 63 was incorrect. LBP 97 il,45 NRC -
441 (1997).- thr the reasons set forth below, we agree with the Staff's positions regardinF
' both the rounding issue and Question 63. We therefore grant the Staff's petition for review and reverse the Presiding Officer's decision requiring the Staff to
. issue Mr. Tetrick an SRO license.2
)See Funkte Anmur ENetric Co (Yankee Nuclear Power sununt 011%7. 43 NRC 235. 247 a 6 0996)(*the prevading party below Inwy) argue any ground that woukt defend the ulurnate roult reactied by de Board -
inclu bag argurnents that the Board had rejected'). 2 in our v6ew, our dnpoution of this cme would not beneet from requanng ruH bnenng } ' 28 s P
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HACKGROUND Pursuant to Part 55 of our regulations, an applicant for an SRO license nest pass both a written and an operating examination. De passing score for the written examination is 80%. " Operator Licensing Examiner Standards," NUREG 1021. Mr. Tetrick passed the operating exam but received an initial score of only 78% on his 100-question written test, taken on Jur.e 14, 1996. On July 30,1996, he sought an informal Staff review of his score on the latter exam, challer.ging the grading of four questions. On September 12,1996, the Stalf upheld the grading of three contested questions be? agreed with Mr. Tetrick that the fourth was invalid and should be deleted. The Staff therefore , raised Mr, Tetrick's score to 78.8% (seventy-eight out of ninety nine). On September 25, 1996, Mr. Tetrick sought a hearing before a Presiding Officer. Mr. Tetrick continued to challenge the grading of the remaining three questions, and also contested the scoring of another question. Ibilowing an informal hearing under 10 C.F.R. Part 2, Subpart L, the Presiding Officer issued LDP-97 2, ruling that one of the challenged questions (Number 96) was - ambiguous and should be stricken from the written examination, but holding that Mr. Tetrick's answer to the other three challenged questions (Numbers 63, 84, and 90) were indeed incorrect. 45 NRC at 53-58. His ruling had the effect of raising Mr. Tetrick's score to 79.59% (seventy-eight out of ninety-eight questions) Because the Presidmg Officer concluded that the written SRO tests were "not so precin that tenths of a percent have any meaning " he rounded Mr. Tetrick's revised score of 79.59 to the nearest integer, 80 thereby giving him a passing grade on the written examination. LBP-97-2, 45 NRC at 60. On March 10,1997, the Staff sought reconsideration of the Initial Decision. The Staff challenged the Presiding Officer's authority to round up Mr. Tetrick's score and submitted supportive evidence showing a Staff practice not to round scores upward to the nearest integer. On March 27,1997, the Presiding Officer denied the Staff's request on the ground that the Staff had improperly raised an argument based on evidence that the Staff could have (but had not) submitted during the hearing stage of the proceeding. According to the Presidmg Officer, the Staff should have anticipated the possibility that he would rule in Mr. Tetrick's favor regarding one of the four contested questions and that the rounding issue would therefore arise. In justifying his prior ruling regarding rounding, the Presiding Oft:cer explained that the Staff's recent amendment of NUREG 1021 to require a passing score of "801#1 percent" rather than simply "80 percent" was not yet in effect at the time Mr. Tetrick took his written exam, and that there was no other published guidance concerning either the number of signi6 cant digits in an examination 29
score, or whether and how the score should be rounded. LBP-97-6,45 NRC 130, 131 32 (1997). He Staff filed _with the Commissiori both a request for stay and a petition. for review of the Presiding Officer's rulings in LDP-97-2 and LDP-97-6 on the rounding issue. Responding to the Staff's petition for review, Mr. Tetrick anerted that, if the Commission were to review the Presiding Officer's decisions on the rounding issue, it should also examine whether the Presiding Officer was correct in ruling that Mr. Tetrick had incorrectly answered Question 63 of the written SRO examination. Shortly thereafter, the Staff submitted to the Commission a May 1,1997_ letter in which Mr. RJ. Ilovey, the utility's Vice-President at Turkey Point stated his behef that Mr. Tetrick's answer to Question 63 was a correct one, The Staff, however, continued to maintain otherwise. De Commission concluded in CLI 97-5 supra, that the Question 63 issue appeared to turn ultimately on the interpretation of language in a number of 1:chnical documents, some of which might not be in the record. He Commission therefore remanded the issue to the Presiding Officer and directed him to reconsider his prior ruling. The Commission also retained jurisdiction over the Staff's petition for review of the Presiding Officer's rulings on the rounding
~ issue: deferred ruling on that issue; and granted a temporary stay of LHP-97-2 and LBP-97-6.
' On remand, the Presiding Officer sought further information from the parties (May 27,1997 unpublished order) and, based on that information, issued LBP-97.ll, supra, reaffirming his earlier determination that Mr. Tetrick had incorrectly answered Question 63. De Presiding Officer reasoned that Mr.
. Ilovey's support of Mr. Tetrick's answer was based _ on the erroneous assumption that the question posited only one annunciator. De Presiding Officer also found that Mr. Tetrick's proposed verification of the two consistent a:munciaters was unnecessary, F i ven that they verified each other. In addition, the Presiding Officer was influenced by Mr. Tetrick's failure to respond directly to the
- questions regarding what specific steps Mri Tetrick would take to verify the validity of the alarms and what would persuade him not to take the required IMMEDIATE ACTION after he had taken those steps. 45 NRC at 445-47. De case is now back before the Commission to decide the Staff's petition for review challenging t! 3e Presiding Officer's decision that Mr. Tetrick should receive his SRO license. DISCUSSION We are faced with three issues in this proceeding: (1) whether the Presiding Officer erred in concluding that the Staff's failure to present its rounding 30
arguments at the hearing bars it from raising it on reconsideration; (2) if so, is the Staff's argument on rounding correct; and (3) is the Presiding Officer correct that h1r. Tetrick incorrectly answered Question 63. We answer all three
. questions "yes."
A. The " Rounding" Issues We cannot accept the Presiding Officer's conclusion that the Staff should have anticipated at the hearing that it would need to present its evidence and argurnents on the rounding issue. Although we agree with the Presiding Officer that the Staff could reasonably have anticipated both that he might rule in hir. Tetrick's favor on one of the exam questions and that such a ruling would raise his score to either a 79.59 (question deleted) or 79.80 (question graded in hit. Tetrick's favor), we see no reason why the Staff should have further anticipated that the Presiding Officer would then round the revised score upward to the next integer. The version of NUREG 1021 in effect at the time hit. Tetrick took his exam
;(Revision 7i Supp.1 Qune 1994)) did not address rounding directly but did state that a successful applicant must answer correctly "at least 80 percent" of ,
the questions on the written examination.5 We believe that the phrase "at least" on its face suggests strongly that 80% is the minimal acceptable score and that rounding up lower scores is impermissible. Our conclusion is supported by the Osford English Dictionary which defines this two-word phrase as "a qualifying phrase, attached to a quantitative designation to indicate that the amount is the smallest admissible."* See also Webster's Third New International Dictionary (G.&C. hierriam Co.1976) at 1287 ("at least" means "at the lowest estimate").
'the Staff's consistent prior practice confirms our understanding of the "at least 80 percent" standard. The Staff has refused in the past to "round up" almost-passing scores and has considered the 80% cutoff score as the grade below which a candidate will not pass the written exam.8 " Agency practice, of course, is one indicator of how an agency interprets its regulations." l'anker Atomic Electric Co. (Yankee Nuclear Power Station), CLI 96-6,43 NRC 123, I
NURE G-1021 (Revmon 7. supp I, June 1994L Emanuner stand.ards (LS) 401 at 6 of 7 Obrm [S-4niK 4 The Comport Edispi rf the o@nt Farfu4 Dwrnm.4 Vul . Letter L., p 160, col 2 (osford Omv. Press 1979)(enghi askledt othe' p.wtsons of die sanu vernon of NURIG1021 uw the synonym phraw "810 percent . et gremer" see LS-401 at I of 7. t1402 at 5 of 6. ES 301 at 3 of 24 We construe thn quotes phraw to have
- a nwamng identical to "at Ic.nt s0 perant "
8 sce staff's Request for Suy. dated April 11,1997. at 4 and supparung evidence cited dierem 6nclushng three other twem instances in wlui:h the staff refuwd to Ikenw an appheant wuh a wnnen esam score between 79 50 and 800th-
' 31
129 (1996), Given that the Staff itself set the 80% threshold in the first place.' we are disinclined to disturb its consistently held view.' At bottom, the decision whethu to round up near-passhg scores requies a policy choice, Either option is plausible. Here, itthe adjudicatory setting, we decline to set aside the NRC Staff's policy judgment, supported by the language of NUREG 1021, to draw the pass. fait line at 80% minimum, without niunding up. Cf RockwellInternational Cctp (Rocketdyne Division). ALAB-925,30 NRC 709,722 n.15 (1989), aff'd, CL1-90-5,31 NRC 337 (1990). In our view, when the Presiding Officer ordered rounding up on the ground that - the SRO written examinations "are not so precise that tenths of a percent have any meaning"(LilP-97 2,45 NRC at 60) and essentially reduced the passing score from 80% to 79.5%, he stepped into a Staff wea of responsibility,' B. Question 63 Mr. Tetrick raises with the Commission the issue whether he correctly answered Question 63 of his written SRO examination. ht question read as follows: Plant condiuons:
- Preparatons are being snade for refuchng operauons
- The refuehng cavtry is filled with the transfer tube gate ulve open.
- Alarm annunciators Rll 5FP LO LEVEL and G45. CNTMT SUMP fil LEVEL are in alarnt Which ONE of the following ts the required tMMEDIATE ACTION in response to these conditions?
'Ste Menuwandum to'All power Reactor Appheants and Licenwes from Hsold R Denton, Director NRC's ofhce of Nuclear Reacttw Regulation, d.ited Msch 28.1980, apprn&d as Anahment I to Staffs Monon for Reconna&ranon. dated Msch 10.1997
'The NRC tecently revind NURLo-1021 to replae the nurumum paumg grade of "80 percent" muh ~8000 percent." See NURI'G 1021 (Intenm Rev. 8), 8%40! at 39 ef 39 (hwm 1.5 4017)and Appendn E at i of 5 (January 1997). Ilus tius reviuon does not suppon an empheauon that the fornrr term pernutted rounding and therefoni needed cimenon Rather, the reviuon was akin both to the clanf)ing regulatory amem!ments that thas Compusuon ami other agoracies regulaly pomulgate and to the clanfying legislatma that Congren segulaly enacts See, a s. Final Rule,"preparanon. Transfer for Conunercial Dnnibution, and Use of Dyprodut Matenal for Medical Use,* 59 fvd. Reg 61,767. 64.'r76 (Dec. 2,1994i hag Yems $=ng v AlcGrat4. 339 U.5 3147, swedrAed 339 U 5 008 (1930) 8 our rewarch has l&nuhed wseral cases from around the country mhere the judicury dechned to thsturb tesung authannes' refusal to "round up" ainuut pasung scores See Aritty v, tenrr 1988 Wt,49187 at *4 (5 D N Y.
May 6.1988L Afdshn4 v Nor,mr4 of AfanAntsua Owmuerry College. 78 A D 2d 839. 413 N Y S 2d 446,447 (1st Dept 19401. strd 55 N Y.2J 913. 913. 431 N E 2J 1274.1275. 449 N Y 5 2d 26, 27 (1982K Afarquer v. Umwrurv of %Aragna 32 wah. App 302,3tN e48 p 2J 94,98 (1982), Smulsty, another deciuon deferred to the lesang authonty's &ternunmion to follow a " rounding up" pohey. See AvA v. Pnisce Comarttioner of Boo,n 11 Mass App 630. 65L 418 N E.2d 621624 (1981) Tlus bne of cases suppwis our view that the deucon in *round up* or not is for the tesung withonues, not ttw airadscaton. to make 32
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. All' parties, including Mr. Tetrick[ recognize that answer "b" is correct.- .(
i'therefore, the only issue before us on. appeal regarding Question 63 is wlether Mr. Tetrick's answer of"a" is also correct. Tur the reasons set forth in both LBP4
~
; 97-2 and LBP 97 ll, we conclude that answer "a" is incorrect.' We therefore ;
l- . cannot use Mr. Tetrick's' answer to Question 63 as a ground to affirm'the-Presiding Officer's result in this case. j CONCLUSION i;
' . We grant the Staff's petition for review and reverse the Presiding Officer's.
rulings in bodt LBP-97 2 and LBP 97 6 regarding' the." rounding" of Mr. 4 l'Ibtrick's written examination score. Commissic,act Diaz disapproved this order. = 6 IT IS SO ORDERED.- I For the Commission
! JOHN C. H9YLE . f Secretary of the Commission ,
F . Dated at Rockville, Maryland. t this 7th day of August 1997.- S l 4 g- 'la sernameng this issue to the Presiding o(ticer. me rehed in large pwt on Mr. Ikwey's May Ise leiner arsmns that IK'de 'a" and "b" are adequase avsponses to Quesuon 61 We therefore behew that a brief esplanauon it appenpriase as to why og resolw du "Quespoa 63^ snue d:rfere'nly trom Mr. Ihwey. In our view, we agrra with -
,' an theIrnmesase Presadseg Officer that Mr. thwey tuies his conchmon on the erroneous anampnon that Question 63 asked for acuen in sosponse to "ma" annunciator alum The quesnon instead asked for the munediase accon ;
in gesponw to two annunci.uor alarms asider #As spacek piper oundreat speqPed in iAt enesirva U 3 33 4 t i ; i
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t Cite as 46 NRC 35 (1997) DD-9718 UNITED STATES OF AMERICA
- NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Samuel J. Collins, Director in the Matter of Docket Nos. 50 282 50-306 72 10 '
NORTHERN STATES POWER COMPANY (Praltle Island Nuclear Generating Plant; Prairie Island Independent Spent Fuel Storage Installt.tlon) August 29,1997
%e Director of the Office of Nuclear Reactor Regulation denies a petition filed by the Prairie Island Indian Community pursuant to 10 C,F.R. I2.206.
The petition asked that the NRC: (1) find that the Licensee violated NRC reputations by using an Independent Spent Fuel Storage Installation before establishing conditions for safety unloading TN-40 dry storage containers, (2) suspend the license until all significant issues concerning the unloading pn, cess have been resolved. (3) provide the Petitioners with an opportunity to participate
' fully in reviewing the unloading procedures for the casks, and (4) update the relevant technical specifications to incorporate mandatory unloading procedure requirements for the TN 40 dry stora;.e containers.
DIRECTOR'S DECISION UNDER 10 C.F.R. 6 2.206
- 1. INTRODUCTION On May. 28, '997, the Prairie Island Indian Community filed a petition pursuant to section 2.206 of Titic .10 of the Code of Federal Replationt (10 35
C.F.R. 6 2.206) requesting that the U.S. Nuclear Regulatory Commission (NRC) take action to accomplish the following: 1, Deternune that Northern States Power (NSP) notated Ltw requirenwnts of 10 C F.R. 172122(I) by using its M.derials L6 cense No SNM.2Sn6 for an independent Spent Fuci Storage Installaton (ISI'SI) prmt to cuablishing condiuons for safely unlomhng the TN40 dry storage containers. 2 Suspend Matenals License No, SNM-23n6 for cause urnier 10 C F R l 50100 until such tinw as #11 sigmheunt issues in the unloading process, as described herein Ithe petiuun), have been rewived, the unloading process has tren demonstrated, and unut an ir* pendent third party revnew of the TN40 unloahng procedure has been conducted. 1 Provide Ittitwners an opportunity to participate fully in the reviewing of the unbeding procedure for the TN 40 cask, hold heanngs and allow lYtnioners to participate fully in these and any other procedures initiated in response to this pention, and
- 4. Up late the Technical Specihcanons (TS) for itw Prairie Island ISFSI to incorporate nundatory unloading procedure requirements
'the petition has been referred to me pursuant to section 2.206. The NRC letter dated June 27, 1997, to Byron White, on behalf of the Petitioners, acknowledged re(cipt of the petition and provided the NRC Staff's determination that the petition did not require immediate action by the NRC. A notice of receipt was published in the Federal Register on July 3, 1997 (62 fid. Reg. 36,085).
On the basis of the NRC Staff's evaluation of the issues and for the reasons given below, the Petitioners' requests are denied. II, llACKGROUND On October 19,1993, the NRC issued hlaterials License N. t Nht 2506 to NSP (the Licensee) to allow storage of spent nuclear fuel in TN n *v storage rasks, designed by Transnuclear Incorporated, at the ISFSI located . 'airie Islend Nuclear Plant. No spent nuclear fuel was allowed to be hiaded into a storage cask at Prairie Island until several preoperational license conditions were satisfied. Among the preoperational heense conditions were a required training exercise (dry-run) of the loading, handling, and unloading activities for the *1N40 casks and the implementation of wrinen procedures describing the actions to be taken during operation, off-normal, and emergency conditions asaociated with the Prairie Island ISFSI, De NRC issued TS defining operating limits, surveillance requirements, design features, and administratise controls as Appendix A to hiaterials License No. SNht 2506. 36
A report dated April 20,1995, submitted by the Licensee to the NRC pursuant to 10 C.F.R. 6 72.82(e), provided the results of the preoperational tests that were required to be performed by the Licensee before loading of spent fuel into a TN-40 cask. On hiny 11,1995, the NRC granted a schedular exemption to the provision of section 72.82(c) that requires licensees to submit the preoperational test results at least 30 days before receipt of spent fuel into the ISFSt. The basis for the exemption was the fact that the NRC Staff had reviewed cask fabrication records, observed portions of the preoperational test activities as they occurred, and completed its review of the report submitted on April 20,1995. On hiny 12,1995, the Licensee began loading spent fuel assemblies into a TN-40 cask. The Licensee subsequently placed the cask, and casks loaded since that time, onto the storage pad within the Prairie Island ISFSI. i NRC regulations include a requirement that an ISFSI be designed to provide for the ready retrieval of spent fuel or high-level radioactive waste for further processing or disposal. This regulation,10 C.F.R. 5 72.122(I), provides as follows: RetnevaNIny. Storage systems must be designed to allow ready remeral of spent fuet or higfrievci radioacuve waste for further processmg or Aposal. Certain events or conditions could warrant removing a TN-40 cask from the - Prairic Island ISFSI and returning it to the spent fuel pool and unloading the stored fuel assemblies. In addition to the regulatory requirements in section 72.122(I) pertaining to retrieval of the fuel assemblies for further processing or disposal, the TS for the Prairie _ Island ISFSI require the Licensee to take corrective actions in response to those design-basis events or conditions that may challenge the integrity of the storage cask or the chdding of the spent fuel assemblies. For example, section 2.3, "hfaximum Cask Lifting Height" section 3/4.3, "htaximum IIelium Leak Rate," and section 3/4.5, "htaximum Cask Surface Temperature," of the TS include provisions for unh>ading of a
'IN 40 storage cask in response to the specified events or conditions.
NRC regulations in i0 C.F R. Part 72 require that the design of the storage system and the pmcedures implemented by specific licensees .upport the un-loading activity, whether it is being performed to allow further processing or disposal of the spent fuel or it is required as part of the response to an unplanned event or condition, while preventing gross ruptures of the fuel cladding in or-der to prevent operational safety problems. Unhuding procedures should also include contingencies in case fuel cladding has degraded during storage such that additional measures are necessary to address increased radiological hazards during the unloading process. NRC regulations, facility licenses, and NRC appmved quality assurance pmgrams require licensees to establish and maintain a formal process for the 37
preparation and issuance of procedures and changes thereto. NRC assessments of licensee procedures are generally conducted as part of the NRC's inspection program. _in this instance, the major procedures pertaining to dry cask storage activities at Prairie Island, including the procedure for unloading a cask, were reviewed by the NRC Staff during a special inspection conducted from January 24 through May 11,1995, in addition to the review of the Licensee's facihty and procedures, the NRC inspectors observed preoperational testing that the Licensee was required to perform before loading casks with spent fuel assemblies. The inspection findings are documented in NRC Inspection Report 50-282/95002, 50-30W5002,72-10/95002(DRP), dated June 30,1995.
'Ihe NRC inspectors identified several instances in which the procedures for dry cask storage activities that the Licensee had in place at the beginning of the inspection, including the procedures for loading and unloading of the TN 40 casks did not ensure compliance with the requirements of the license Although the inspectors were able to verify that the Licensee corrected the identified piocedural deficiencies during the course of the inspection, a Notice of Violation was issued to the Licensee for failing to satisfy Criterion V of Appendix B to 10 C.F.R. Part 50, which, for activities affecting quakty, requires the preparation and adherence to procedures appropriate to the circumstances. In addition, the inspectors identified weaknesses in the Licensee's initial performance in overseeing the activities of the cask vendor and in overall planning for dry cask storage activities, lloweser, on the basis of the licensing reviews and inspection findings, documented in Inspection Report 50-282/95002, 50-306/95002, 72 lW95002(DRP), the NRC Staff concluded that as of May 1995, the Licensee had corrected the identified deliciencies and was ready to safely load and, if necessary, unload spent nuclear fuel m TN-40 casks.
In July 1995, the NRC Staff issued an action plan for dry cask storage to manage the resolution of a variety of technical and process issues that were identified during the licensing reviews and inspections completed for the first several ISFSI facihties. An item related to the loading and unloading of dry storage casks was added to the action plan, in part to ensure that the importance of the unloaling procedures was emphasized to licensees and technical issues . related to unloading problems were resolved. Addition of an item pertaining to unloading was deemed prudent because the Staff observed that some unloading procedures implemented by licensees neglected to consider contingencies and assumptions related to possible fuel degradation, gas sampling techniques, cask design issues, radiation protection requirements, and the thermal-hydraulic behavior of a cask during the process of cooling and filling it with water from the spent fuel pool. To implement the action plan, the NRC Staff formed a working group to identify issues auociated with loading and unloading processes for dry storage casks and to propose means of informing the industry and the NRC Staff of those 38
issues. He working group coru.idered industry experiences, concerns identified during reviews and inspections, and other issues related to loading and unloading procedures. The working group completed its reviews in = April 1996. De concerns related to unloading procedures reviewcJ by the working group were found to involve either (1) isolated occurrences that had been adequately resolved - Ay site-specific corrective actions or (2) generic issues that were addressed
. by incorporating remedial measures into ongoing Staff activities, such as the preparation of revised inspection procedures or other guidance documents.
To fulfill some of the goals hcluded in its dry cask storage action plan, the
- NRC Staff has emphasired the importance of unloading prwedures and shared obwrvat;ons with licensees using or considering dry cask storage during oppor-tunities such as the Spent thel Storage and Transportation Workshop held in May 1996 and meetings with individual licensees. He Staff revised inspection procedures to specifically instruct NRC inspectors to review unloading proce-dures developed by licensees and to identify those issues that warrant particular attention. Guidance included in NRC Inspection Procedure 60855, " Operation of an ISFSI," issued February 1,1996, states:
hw unhwhng activities, anennon should be paid to how the heensee has prepared to deal wuh the potential hararda nuociated with that task. Sona potennal inues may include: the rachation exposure associated with drawmg and analyziag a sample or the canister's potentially nuhoactive asnunplere; steam R.whmg and pressure control as water is added to the hot camater, and rittenng or scrubbmg the but steam / gas nuxture vented from tie vanidet, as it is Glied with water. Similar guidance was included in NUREG-1536, " Standard Review Plan for Dry Cask Storage Systems," issued in January 1997. Application of the revised guidance ensures that recent and future reviews will address the adequacy of unloading procedures developed by licensees. De Staff also issued NRC Information Notice 97 51," Problems Experienced with Loading and Unloading Spent Nuclear Fuel Storage and Transportation Casks," datet' July 11,1997, to inform licensees of operating experiences and problems ent.ountered with the loading and unloading of storage and transportation casks for spent nuclear fuel. To address those ISFSis that began operation before the improvements in the NRC's review and inspection guidance, the Staff performed audits or inspections of those Licensee programs for which the inspection record did not document whether the unloading pmeedures adequately addressed the major issues included in the action plan. Regarding Prairic Island, the Staff reviewed the available information and determined that additional reviews or inspections were not necessary because the assessment of the unloading prc cedure performed as part of the inspection documented in NRC Inspection Report 50-282/95002, 50 30N9500L 72-10/95002(DRp) adequately addressed the concerns included in the NRC action plan. 39 h- I
III.- DISCUSSION De petition requests fout actions by the NRC on the basis of the contention that the unloading procedure irnplemented by the Licensee was inadequate and, therefore, the Licensee violated the NRC regulation requiring it to have the ability to readily retrieve spent fuel or high-level radioactive waste for further processing or disposat item 1: 1)etermine nat the IJctnser Violated Section 72.122(l) in support of the petition's contention that the Licensee violated NRC requirements, the Petitioners claim that the procedure to unload a TN-40 cask at Prairie Island has not been adequately evaluated or tested because neither the NRC nor NSP has completely demonstrated that a TN-40 dry cask can be unloaded after it has remained on the storage pad for a number of years. The Petitioners state that their request is supported by the fact that the preoperational test results for the Prairie Island ISFSI were submitted to the NRC on the day before the unloading procedure was approved by the Licensee's Operations Committee. De Petitioners also express concern that only portions of the Licensee's unloading procedure were tested during the required preoperational tests and, therefore the tests did not provide assurance that an unloading can be done safely, in addition, the Petitioners state that procedures for unloading a cask should address specific concerns regarding failed fuel recovery and possible contamination of the spent fuel pool, venting of radioactive gases, functional checks of radiation monitoring and ventilation systems, and the buildup of steam when water is pumped into the cask as part of the unloading pmcess. As previously mentioned, cask designs and associatea procedures are required to support the unloading of the spent fuel assembhes either to support further processing or disposal or in response to an unplanned event or condition that may challenge the integrity of the storage cask or the cladding of the spent fuel assemblies. Although the NRC Staff agrees with the Petitioners' premise that actually unloading a storage cask would likely result in licensees !carning lessons that could result in additional enhancements to unloading procedures, the Staff does not agree that an actual demonstration of the unloading procedure at Prairie Island is warranted, in addition to the Staff's review of the procedure for unloadmg a 1N-40 cask at Prairie Island, reasonable assurance that the TN-40 casks can be safely unloaded is provided by a varkty of experiences related to the use and storage of radioactive materials. These experiences include the dry run exercises thrit were performed to verify key aspects of unkiading procedures for the TN 40 cask; related rescarch sponsored by the commercial nuclear industry, the U.S. Department of Energy, and the NRC; actual loading and unloading of transportation casks; kuding of storage casks; handling of spent fuel assemblies 40
~
under various conditions; and performing relevant maiatenance and engineering activities associated with reactor facilities. Regarding the Petitioners' concerns pertaining to the dates of the submittal of preoperational tests and the approval of the Licensee's unlaading procedure, the NRC Staff identified this discrepancy in inspection Report 50-282/95002, 50006/95002,7210/95002(DRP). De administrative controls included in the TS for the Prairie Island ISFSI require that the Operations Committee review arx! approve procedures and changes thereto. The approval of the Operations Committee is usually the last step in the process for preparing or revising a procedure. The fact that the Operations Committee approved the procedure shortly after submittal of the preoperational test results and before fuel loading satisfied the preoperational license condition to implement written procedures before loading spent nuclear fuel into a 'IN-40 cask. nis matter does not, therefore, represent a violation of NRC requirements or introduce concerns pertaining to the technical adequacy of the unloading procedure. He Petitioners identified several concerns pertaining to the lack of specific guidance in the unloading procedure to address a scenario in which signifi-cant fuel degradation occurs during storage. He NRC Staff agrees with the Petitioners that such a scenario would complicate the unloading process by re-quiring additional measures and precautions to limit the release of radioactive materials from the cask into parts of the reactor facility and nearby environs. The Licensee's unloading procedure includes a step to sample the atmosphere within the cask cavity to test for radioactive and flammable gases before vent-ng the ca k cavity and loosening the bolts securing the cask lid. Ibliowing , the attalysis of the gas sample, the Licensee's unloading procedure incledes a hold point to allow personnel to determine whether additional steps or precau-tions are warranted. While acknowledging many of the Petitioners' legitimate concerns regarding the potential difficulties in retrieving failed fuel from dry storage casks, the NRC Staff has concluded that licensees need not be required to incorporate specific guidance into the normal unloading procedure to ad-dress this unkkely situation. The Staff's conclusion is based, in part, on the fact that the required compensatory actions and precautions needed to address such situations may vary significantly, depending on the actual results from the analysis of the gas sample. Requiring the Licensee to include contingencies or steps in the unloading procedure to address the unlikely event of failed fuel may unnecessarily complicate and delay the unh>ading of fuel assemblies that have remained intact during storage. On the basis of licensees' experiences in developing and implementing plans to address the problem of fuel assemblies damaged during reactor operations, in handling radioactive wastes of various forms, and in resolving other comparable problems, the NRC Staff has con-tidence that licensees could, if necessary, develop a plan to retrieve damaged fuel from a storage cask while minimizing the radiological consequences to 41
. plant workers and the general public. In addition to the general confidenc: of the NRC Staff that the technical problems associated with retrieving failed fuel could be overcome, requirements for planning and executing such an activity ,
are contained in the licenses issued for the Prairic Island ISFSI and t! e Prairie Island Nuclear Generating Plant, and NRC regulations in 10 C.F.R. Parts 20, 4 50, and 72. De NRC Staff has, therefore, accepted gas sampling and defined hold or decision points before breaching the cask confinement boundary as an adequate means to address concerns pertaininF to the unlikely degradation of fuel aswmblics during stsrage. De specific issues raised by the Petitioners to support their claim that the Licensee's unkmding procedure is deficient are addressed below.
- a. Failed Fuel Considerations As previously discussed, the NRC Staff has accepted that procedures de-veloped by licensees to support unloading of dry storage casks do not need to address the retrieval of failed fuel, provided that measures to detect possible fuel degradation and a defined hold point for determination of possible compensatory
. actions are appropriately placed within the subject procedures. As documented in NRC Inspection Report 50-282/95(X12,50-30N95002,72 10/95002(DRPJ, the Licensee had originally failed to incorporate a step in the unloading procedure ,
for taking a gaseous sample frorn the cask in order to enaure that fuel degrada-tion had not occurred during storage.' ilowever,in response to the findings of the NRC inspectors, the Licensee incorporated sampling of the cask atmosphere and a hold point for deliberation into the unloading procedure and the revised proce-dure was in place before spent nuclear fuel was loaded into a TN-40 cask. The NRC Staff has found that this action, in combination with the requirement that spent fuel assemblies loaded into TN 40 casks be free of gross cladding defects, provides reasonable assurance that the Licensee will no' unknowingly breach the confinement boundary of a cask containing failed fuel, in the unlikely event that the gaseous sample indicates that spent fuel assemblics have degraded during storage, the unloading procedure instructs the Licensee's Operations Committee to add steps or precautions to the procedure in order to minimize the radio-logical consequences of retrieving the failed fuel. The NRC Staff has found this approach to be acceptable and does not require the Licensee's normal un-loading procedure to include contingency actions to address the possible release of radioactive materials to parts of the reactor facility, including the spent fuel pool, that may occur if fuel assemblies degrade during storage. The NRC Staff believes, however, that the petitioners have identified valid concems regarding the potential recovery of fuel assemblies that have unexpectedly degraded dur-ing storage. As previously mentioned, the Staff believes that the regulations - and licenses issued by the NRC require the Licensee to address these and other 42
problems that may occur in the unlikely event that fuel assemblies that have degraded during storage need to be unloaded from dry storaFe casks. bl ' Venring of Radioactive Gases The possible need to vent radioactive gases from a cask is among the issues that the Licensee would need to address if the required sampling of the atmosphere within a cask indicates that the spent fuel assemblies have experienced unanticipated degradation during storageJ - As with the concern regarding the contamination of the spent fuel pool, the need to vent the cask - while minimizing the radiological consequences of unloading a cask containing . failed fuel is an issue that the Licensee would need to address before revising the procedure and proceeding with the unloading process. In addition to ensuring that the unloading activity results in occupational doses and doses to members of the p ublic that are as low as is reasonably achievable (see 10 C.F.R. 5 20.1101), the Lice..see would need to perform the venting of a cask containing fsited fuel
'in accordance with the Prairie Island Nuclear Generating Plant Pacility Operating
= Licensesiassociated TS, and applicable regulations.
- c. Radiation Monitors ne Petitioners contend that the unloading procedure must include a "stop-check" to verify that ventilation systems and radiation monitors are functioning ,
before the venting of a cask is performed. Although agreeing with the Pe-
' tttioners' general premise that prerequisites to preforming procedures should include establishing confidence in the tools and equipment being used, the NRC Staff notes that during the anticipated unkwding of spent nuclear fuel that has not degraded during storage, spesial ventilation or radiation monitoring equip-ment beyond that specified in the Licensce's unloading procedure and radiation protection program is not required. The unloading procedure requires the in-volvement of radiation protection personnel and the activity must be controlled
- in accordance with the Licensee's radiation protection program, which includes provisions for the maintenance and calibration of radiation detectors. Although the venting process is not expected to need ventilation systems equipped with fil-ters and radiation monitors, the spent fuel pool special ventilation system could be used if necessary. The spent fuel pool special ventilation system is required to be operable during subsequent steps in the procedure if spent fuel assemblies are being moved and the system must be tested and maintained in accordance
- with the TS for the Prairie Island Nuclear Generating Plant. In the unlikely event
- that the Licensee needs to unload a cask containing degraded fuel assemblies, confirming the operability of those ventilation systems and additional radiation 43
.I
monitc;ing equipment being used to minimize the release of radioactive mate-rials is an activity that the Licensee would need to address befire revising the procedure and proceeding with the unloading process.
- d. Steam Buildup The Petitioners expressed concerns regarding the reaction of the cask and stored fuel assemblics to the intniduction of spent fuel pool water during the esecution of the unloading procedure. The unloading procedure includes the partial immersion of the 7%40 cask into the spent fuel pool, connection of hoses to the vent and drain connections, and the slow introduction of spent fuel pocl water to the cask cavity and stored fuel assemblies. The procedure instructs personnel to continuously monitor the temperature and pressure instrumentation installed on the vent connectiori and to stop pumping water if the pressure exceeds 10 psig or the temperature exceeds 240 F. In the Staff's judgment, the cooling process imposed by these limitations on temperatures and pressures at the vent port of the cask will adequately ensure that the cooling of the cask and spent fuel is gradual and, thereby, prevent safety problems that could hypothetically result from damage to the cask or the fuel assemblics because of stresses induced by a poorly controlled addition of cooling water from the spent fuel pool.
The Petitioners expressed concerns pertair.ing to the range of the instrumen-tation used during the venting of a TN-40 cask an<l stated that higher ranges for temperature and pressure are necessary. The instrumentation ranges specified in the unloading procedure's drawing of the cask vent port adapter are 50-30TF for temperature and 0-30 psig for pressure, While not judging if these are the optimum ranges for the instrumentation, the NRC StaiT finds that the ranges are adequate to support the administrative limits of 240 F and 10 psig established in the procedure and the related response action of stcpping the addition of water to the cask if these administrative limits are exceeded. Regarding the Petition-ers' concern regarding the need to post hazard warnings during the refilling of a cask, the unloading procedure does include several notes and precautions to remind personnel that the fluid exiting the vent port may present radiological and thermal hatards. In summary, many of the Petitioners' concerns pertain to potential problems with unloading spent fuel from a TN 40 cask if the fuel cladding has degraded during storage. While acknowledging that such concerns regarding the potential difficulties in retrieving failed fuel from dry storage casks are legitimate, the .NRC Staff has concluded that licensees need not be required to incorporate specific guidance into the normal unloading procedure to address this unlikely situation. On the basis of its review of the information provided by the Petitioners and its reviews of the Licensee's procedure for unloading TN-40 44 i
casks at Prairic Island, the NRC Staff has not idenafted imlahrnu ohecunt,
- 72.122(l) or other regulatory requirements pertainig w the rentent <n 21uabty of the Licensee's unloading procedure.
Item 2: . Suspersd Afaterials License No. SNSI 2$06 On the basis of the contention that the Licensee's unloading procedure was inadequate, the Petitioners requested that Materials License No. SNM-2506 be susper ded until such time as the significant issues in the unloading process have been resolved, the unloading process has been demonstrated, and an independent third party review of the TN 40 unloading procedure has been conducted.'
' As previously stated, the NRC Staff has performed a review of the procedure for unloading a TN 40 cask at Prairie Island. The review, including verification that the Licensee's' unloading procedure was revised to address deficiencies identified by the NRC inspectors, is documented in NRC Inspection Report 50-282/95002, 50406/95002, 7210/95002(DRP). The review performed during the NRC inspection, subsequent evaluations perforrrad by the NRC Staff as part of the activities associated with the dry cask storage action plan smd the review of this petition, and the requireJ contml of the procedure in accordance with Licensee programs developed in accordance with NRC regulations, facility beenses, and NRC-approved quality assurance programs provide reasonable confidence that the Licensee could, if necessary, safely unload a TN 40 cask.
Regarding a third p arty review, the NRC Staff's concern about the quality of licensees' unloading procedures led it to include the issue in the dry cask storage action plan. The action plan provided a framework for the identification and resolution of various technical and administrative issues related to the use of dry stoiage casks. The previously mentioned actions taken by the NRC Staff and licensees adequately resolved the identified issues pertaining to cask unloading pmcedures. In the specific case of the unloading pmcedure at Prairie Island, the Licensee revised the procedure to address the preblems identified by the Staff during its inspection. On the basis of the actions it has already taken, the . NRC Staff does not believe that the situation warrants additional review of the Licensee's unloadmg pmcedere by an independent third party. 8 The Pentioners request that Masenals tJeense No sNM-2506 tie impended for cauw in accordance with to CF R 4 Sn100 Prmsuons for the nudncation, revocanon, or suspenuon of the hcenses fur IsiSI racihues are contained in 10 C F R.1724n The rmsible reasons for impendens hcenses for ISrste in accontance unh nernou 72 60 are sirnilar to the conesponding reausi fur susperr%ng hcenws for productmo and unlaanon facihues in accordance with secuun 30.10t1 45
.advL--"'4 4 4 -J. 4-. +m_. 4
- item 3: lAl!cw ?ctitlaners to Review Procedure, andfor NRC to lloid - -
flearings and Allow Pet ltioners to Plarticyate in the Proceedings , The Licensee has provided the NRC with the unloading procedure, including
; Revision 2, dated November 8,1996, for placement into the public record, and "
the Petitioners have been supplied with or have obtained copies of the procedure from the NRC's document control system. Accordingly. Petitioners have had the opportunity to review a recent revision of the unloading procedure, For the -- reasons previously discussed in this Decision, the NRC Staff sees no reastm to undertake additional reviews of ihe procedure or to initiate a formal proceeding in which the Petitioners could participate. Although the NRC has <tecided not to initiate a bearing in response to this petition, the Petitioners are encouraged to g continue their interactions with the NRC Staff regarding concerns or questions - ; about the operation of the Prairie Island Nuclear Generating Plant or the Prairie Island ISFSI, l leva 4:. Update the Technical Specifications for the Prairie Island ISFSI
;to incorporate Mandatory Un'oading Procedure Repirements
'the TS for ISFSis are required by 10 C.F.R. 672.44 to include requirements in the following categories:
(1) Tbnctional and' operating limits and monnonng instruments and limiting control settings; (2) Limiting condiuans; --
- 0) Surveillance requirements; (4) . Design featurva; and ' '
($) ' Adminhtradve contmit. Although the TS for the Prairie Island ISFSI requires that TN-40 casks be ~ unloaded if certain events or conditions defmed in the TS are satisfied, the TS do not include specific requirements for the unloading process, The content , - of the TS for the Prairie Island ISFSI is typical in this respect since neither a:ction 72,44 nor the anociated regul.itory guidance documents specify that technical specifications should include special requirements for the unloading procedure.8 Instead, the functional and operating limits,- limiting conditions, administrative controls, and other requirements included in the TS for the Prairie 3 Recent NRC stalt guiskukv prwmng to the uppmpnate content or techmcal specthcanons is provided in
,NUREG.15%. "Standad Reue, s%n for D y Cad Storage syvemv puNahed in Lumery 1997. Similar swd.ance h prmeded by NRC Reg Anuwy ouide 3 61, "St.smiiud rirma and Coment for a Torucal Safety Analysis .
Repcst rur a $ pent luel Dry Skag Cask /'suued in february 1989, amt NRC Regukunty Guide 3 48," Standard rivmat and Contene fat the Safety Analyus Report tot an ladepeneet Spent het SKwage Inst.dlanen (Dry
%ttaagel" kwsed in oder 198L 46 e -
- T --
1 - _ ]
lsland ISFSI are intended to maintain the cask and stored spent fuel assemblies within the limits established for safe operation during storage within the ISFSI and activities such as loading and unloading of the casks. For example TS 2.3
- limits the allowable lifting heights Juring movement of the cask from the ISFSI and TS 3/4.2 requires a measurement of the boron concentration of tim water in the spent fuel pool before water is introduced to the cask during the .mloading process.
The absence of specific requirements in the TS to control the unloading pro-cess does not diminish the importance that the NRC Staff places on this activity or the validity of the Petitioners' concerns. 'Ihe NRC Staff believes that other. regulatory requirements provide an equivalent level of protection to the Petition-ers' request to include specific requirements in the TS to control the unloading of a TN40 cask. The administrative controls in the TS for the Prairie Island ISFSI require that the associated precedures, including the unloading procedure, be prepared, reviewed, ,ad maintained in accordance with the requirements of the Prairie Island Nuclear Generating Plant Facility Operating Licenses and as-sociated TS. In addition, under existing NRC requirements, the Licensee must adequately implement pt,cedures to control loading, maintaining, and unload. ing of dry storage casks pee 10 C.F.R. 5672.122, 72.150, and 72.152). Ibr sample, the NRC inspection documented in inspection Report 50-282/95002, 50-306/95002, 72-10/95002(DRP) resulted in a Notice of Violation issued to the Licensee because the Licensee failed to satisfy the NRC's requirements in Criterion V of Appendix 11 to 10 C.F.R. Part 50 by not having incorporated ap-propriate steps and precautions into the origina procedure developed to control unloading of a TN 40 cask. As demonstrated by ti e example, no changes to the TS or the Safety Analysis Report (SAR) are needed to ensure that enforceable operating controls and limits are in place to address the unloading of a cask. In regard to another concern raised by the Ittitioners, the Prairie Island r ISFSI SAR and other docketed correspondence do state that unloading a TN40 cask would be performed using a procedure that is basically the reverse of the procedure used to load the cask. Although this statement, in a general sense, is true, the NRC Staff agrees with the Petitioners that such statements may be misleading in that they oversimplify the description of the unloading activity. Ihr this reason, the NRC Staff included an item related to unloading procedures in its dry cask storage action plan.to ensure that actur.1 unloading procedures did not reflect such an oversimplified representation. The unloading procedure for the dry storage casks at Prairie Island was inspected by the NRC Staff and, as previously discussed, was ultimately found to provide adequate guidance to control the unk>ading process. 47 k
., - n
l I IV. CONCLUSION l lbr the reasons described bove, the NRC has determined that no adequate basis exists for granting the Pentionere' request for suspension of Northern l States Power Company's license for dry cask storage of spent nuclear fuel at Praitic Island or for taking the other actions requested by the Petitioners. While i w,Lnowledging that the Petitioners' concerns regarding the potential difficulties in retrieving failed fuel frorn dry storage casks are legitimate, the NRC Staff has - concluded that licensees need not be required to incorporate specific guidance into the normal unloading procedure to address this unlikely situation. A copy of this Decision will be filed with the Secretary of the Commission
- for the Commission to review in accordance wi;h 10 C.F.R. I 2.20Nc).
As provided by this regulation, this Decision will constitute the final action of the Comrnission 25 days after issuance, unless the Commission, on its orm motion, institutes a review of the Decision within that time. ; IUR Till! NUCLIIAR 1 R13GULA1 DRY COMMISSION t Samuel J. Collins, Director Office of Nuclear Reactor Regulation Dated at Hoskville. Maryland, this 29'.h day of Aupst 1997. 1 48 c 1 l .. l: [-}}